NLS2014015, Supplement to 60-Day Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805

Supplement to 60-Day Response to Request for Additional Information Regarding License Amendment Request to Adopt National Fire Protection Association Standard 805

ML14056A240
Person / Time
Site:	Cooper
Issue date:	02/18/2014
From:	Limpias O Nebraska Public Power District (NPPD)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NLS2014015
Download: ML14056A240 (39)

Text

H Nebraska Public Power District Always there when you need us 50.90 NLS2014015 February 18, 2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

Subject:

Supplement to 60-Day Response to Request For Additional Information Regarding License Amendment Request To Adopt National Fire Protection Association Standard 805 Cooper Nuclear Station, Docket No. 50-298, DPR-46

Reference:

1. Letter from Oscar A. Limpias, Nebraska Public Power District, to U.S.

Nuclear Regulatory Commission, dated January 17, 2014, "60-Day Response to Request For Additional Information Regarding License Amendment Request To Adopt National Fire Protection Association Standard 805" (NLS2014001)

2. Letter from Brian J. O'Grady, Nebraska Public Power District, to U.S. Nuclear Regulatory Commission, dated April 24, 2012, "License Amendment Request to Revise the Fire Protection Licensing Basis to NFPA 805 Per 10 CFR 50.48(c)" (NLS2012006)

3.

Letter from Joseph M. Sebrosky, U.S. Nuclear Regulatory Commission, dated February 6, 2014, "Summary of Cooper Nuclear Station January 28 and 29, 2014, Audit Associated With License Amendment Request to Transition to National Fire Protection Association 805 Standards (TAC No. ME8551)"

Dear Sir or Madam:

The purpose of this letter is for the Nebraska Public Power District to supplement the information provided in the 60-day response to a Nuclear Regulatory Commission (NRC)

Request for Additional Information (Reference 1) related to the Cooper Nuclear Station (CNS)

License Amendment Request (LAR) to adopt National Fire Protection Association (NFPA)

Standard 805 as the CNS Fire Protection licensing basis per 10 CFR 50.48(c) (Reference 2). On January 28 and 29, 2014, the NRC conducted an onsite audit of the referenced response and determined that supplemental information was necessary for the Staff to complete the review of the application (Reference 3). This information is provided in Attachment 1. Additionally, certain requested changes to the LAR are provided in Attachment 2. No new commitments have COOPER NUCLEAR STATION P.O. Box 98 / Brownville, NE 68321-0098 Telephone: (402) 825-3811 / Fax: (402) 825-5211 www.nppd.com kA&-

NLS2014015 Page 2 of 2 been made in this submittal. However, commitments NLS2012006-01 and NLS2013104-01 are rescinded, as they have been escalated to regulatory obligations.

Should you have any questions concerning this matter, please contact Troy Barker, Engineering Programs and Components Manager, at (402) 825-5027.

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

Executed on 7-/ 13 b4 (Date)

Sincer 0OiA.Limp Vice President -1Nuclear and Chief Nuclear Officer OAL/wv Attachments:

1. Supplemental Information to the 60-Day Response to the Request For Additional Information Regarding License Amendment Request To Adopt National Fire Protection Association Standard 805

2. Revisions to the Cooper Nuclear Station License Amendment Request To Adopt National Fire Protection Association Standard 805 Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants cc:

Regional Administrator w/ Attachments USNRC - Region IV Cooper Project Manager w/ Attachments USNRC - NRR Project Directorate IV-1 Senior Resident Inspector w/ Attachments USNRC - CNS Nebraska Health and Human Services w/ Attachments Department of Regulation and Licensure NPG Distribution w/o Attachments CNS Records w/ Attachments

NLS2014015 Page 1 of 25 Attachment I Supplemental Information to the 60-Day Response to the Request For Additional Information Regarding License Amendment Request To Adopt National Fire Protection Association Standard 805 During an audit conducted on January 28 and 29, 2014, at Cooper Nuclear Station (CNS), the Nuclear Regulatory Commission (NRC) audit team requested additional information to supplement several of the Nebraska Public Power District (NPPD) January 17, 2014, responses to the 60-day probabilistic risk assessment (PRA) Request for Additional Information (RAI) regarding the National Fire Protection Association (NFPA) Standard 805 Transition License Amendment Request (LAR). Revisions to these responses are shown by revision bars in the right hand margin.

PRA RAI 02.h.01 RAI-02h involved an F&O against QU-E3. The licensee's response to PRA RAI-02.h dated January 14, 2013 (ADAMS Accession No. ML13018AO06) indicates that statistical propagation ofparametric uncertainty has not been performed and the response's qualitative factor of 5 to 10 overestimation of the estimated results appears to be some measure ofperceived conservatism in the analyses. The Capability Category II (CC-II) for QU-E3 supporting requirement (SR) addresses the uncertainty interval around the estimated value taking into account the state-of-knowledge correlations. Describe how the effect ofpropagating parametric uncertainty on the change in risk estimate was evaluated. In addition, clarify ifstatistical propagation of parametric uncertainty would cause the risk estimates to increase beyond the Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, "Rev. 2, acceptance guidelines; and if not, provide an explanation.

NPPD Response As noted in the PRA Standard for Capability Category II for supporting requirement QU-E3 (ASME/ANS Ra-Sa-2009) propagation of parameter uncertainty is not required; state-of-knowledge correlation needs to be taken into account in estimating the uncertainty interval. As discussed in the response to PRA RAI 02 h, correlation between probabilities is not significant in the fire PRA (FPRA). This conclusion is based on a review of the cut-sets contributing to core damage frequency (CDF) and large early release frequency (LERF). Also, as noted in NUREG-1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making" on page 44, section 4.1.2, paragraph two: "However, for CC II, this correlation may be ignored in calculating the probability distribution provided that it is shown not to be significant for the particular case under assessment."

The distribution is not used in the analysis, as the decision guidance is based on mean values, which have been used in this analysis.

NLS2014015 Attachment I Page 2 of 25 Propagation of parameter uncertainty would not cause the risk estimates to increase beyond the RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Rev. 2, acceptance guidelines due to the insignificance of correlation on the calculated results.

Per the response to PRA RAI 19.01 a. Implementation Item S-3.30 addresses potential changes as follows:

Upon completion of all Fire PRA credited implementation items in Transition report Tables S-2 and S-3, verify the validity of the change-in-risk (total modifications) provided in Attachment W. If this verification determines that the risk metrics have changed such that the risk metrics from LAR Attachment W are exceeded, additional analytical efforts, and/or procedure changes, and/or plant modifications will be implemented to assure the RG 1.205 acceptance criteria are met.

Correlation will be formally included in the model as the FPRA is reconciled to the "total modifications."

As stated in the response to PRA RAI 19.01 b., the acceptance guidelines will not differ from those applied in the LAR. This includes application of RG 1.174 criteria as well as RG 1.205 Position 2.2.4.2.

Supplement to PRA RAI 02.h.01 Response The PRA RAI 40 response has been updated to include the conduct of the uncertainty analyses including state-of-knowledge before using the Fire PRA to support self-approval. Further, a new implementation item S-3.19 will update the Fire PRA model as described in PRA RAI 40 Supplement 3.

PRA RAI 03.01 Section 2.4.3.3 of NFPA 805 states that the probabilistic risk assessment (PRA) approach, methods, and data shall be acceptable to the NRC. RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, " identifies NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, " Volumes 1 and 2, and Supplement 1, "as documenting a methodology for conducting afire PRA (FPRA) and endorses, with exceptions and clarifications, NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), " Rev. 2, as providing methods acceptable to the stafffor adopting afire protection program consistent with NFPA-805. Additional information is requested on the main control room (MCR) risk analysis.

LAR Attachment W indicates that a conditional core damage probability/conditional large early release probability (CCDP/CLERP) equivalent to shutting down from the alternate shutdown (ASD) panel is assumed for those sequences involving failure of the incipient detection system to

NLS2014015 Attachment I Page 3 of 25 notify operators, failure of operators to respond to alert, or failure of operators to respond from the MCR.

a. For each of these failure paths, justijy the use of a.CCDP/CLERP associated with MCR abandonment.

b. For the scenario in which operators fail to respond from the MCR, provide justification that failure of associated MCR and local recovery actions (RAs) may be further mitigated by use of the ASD path.

c. For the scenario in which the incipient detection system fails to notify operators, clarify how operators are made aware of afire in Relay Panels 9-32 and 9-33.

NPPD Response

a.

As stated in the responses to parts b and c of this RAI (provided in 30-day response),

should the incipient detection or Very Early Warning Fire Detection System (VEWFDS) fail or operators fail to respond to a VEWFDS alert, other fire detection would alert operators to a fire in the Auxiliary Relay Room (Fire Zone 8A of Fire Area CB-D).

Given the loss of trains/systems from a fire in Panels 9-32 or 9-33 along with a failure of VEWFDS, a failure to respond to VEWFDS or failure of operators to respond from the MCR, the operators would perceive a significant loss of command and control in the MCR and enter Procedure 5.4FIRE-SD. This procedure provides for MCR abandonment and reaching a "safe and stable" condition from the ASD panel for fires in Fire Area CB-D. Therefore, using MCR abandonment CCDP/CLERP for the fire scenarios where VEWFDS is failed or operators do not respond to the early fire alert properly or do not respond from the MCR is justified.

b.

PRA RAI 03.01b was addressed in the 30-day response.

c.

PRA RAI 03.01c was addressed in the 30-day response.

Supplement to PRA RAI 03.01a Response The refined value for CCDP/CLERP provided in the response to PRA RAI 14.01 is used in the analysis. The updated response to PRA RAI 40 discusses the Fire PRA model update using the refined analysis model. In summary, the analysis in PRA RAI 14.01 is appropriate for scenarios in which the incipient detection system is assumed to fail. This value is applicable because failure of the incipient detection system would result in the exact same Operator response as if the incipient detection system was not installed. Without an alert from the incipient detection system, Operators would respond consistent with any identified fire requiring the use of the alternate shutdown procedure.

NLS2014015 Attachment I Page 4 of 25 PRA RAI 14.01 In a letter dated January 14, 2013 (ADAMS Accession No. ML13018AO06) the licensee responded to PRA RAI-14 and providedjustification of a value of 0. 1 to represent the failure to reach safe shutdown using alternate means. The justification largely consists of a qualitative argument that the feasibility assessment and the seven considerations identified in NUREG-1921 are addressed for alternate shutdown. A quantitative assessment of the failure of alternate shutdown is not presented. It appears from the response that this single HEP value of 0. 1 is used for every MCR abandonment scenario.

a. Describe whether there are any values other than 0. 1 used to characterize the HEP following MCR abandonment and whether the 0. 1 also represents the CCDP. In addition, describe the CLERP.

b. If any values other than 0. 1 are used, (e.g., 1.0), provide the other values, a characterization of the scenarios where these values are used, and a summary of how each value is developed. This information should include explanations of how the following scenarios are addressed.

i. Scenarios where fire induced spurious actuations of equipment can affect the preferred shutdown path.

ii. Scenarios where the fire could cause recoverable failures to the preferred shutdown path. If every such.failure, absent recovery, is assumed to immediately fail the success path, provide confirmation.

iii. Scenarios with fire induced failures unrelated to the preferred success path. Such failures can complicate the efforts to shut down the plant through, for example, spurious operations of unrelated equipment to the success path. If every such failure is assumed [to] immediately fail the success path, please confirm this.

c. If no values other than 0. 1 are used, explain how scenarios characterized under b.i, b.ii, and b.iii (above) are included in the MCR abandonment evaluations.

d. The fire risk evaluations (FREs) should be performed consistent with FAQ 0 7-0030, "Establishing Recovery Actions" (ADAMS Accession No. MLl 10070485), and FAQ 08-0054, "Demonstrating Compliance with Chapter 4 of National Fire Protection Association 805" (ADAMS Accession No. ML110140183) guidance. Note that FAQ 08-0054 provides guidance on the additional risk of RAs for alternative or dedicated shutdown. Discuss the FRE method followed from these FAQsfor the MCR FREs, and explain how the compliant case is defined.

NLS2014015 Page 5 of 25 NPPD Response

a.

In the analysis for the LAR submittal, only the value of 0.1 was used for MCR abandonment fire scenarios. The MCR abandonment CCDP/CLERP of 0.1 was characterized by incorporating both the HEP for shutdown from outside the MCR and system/train failures into the CCDP/CLERP. The MCR abandonment value for CLERP was taken to be equal to the MCR abandonment value for CCDP.

b.

In response to this RAI, the MCR abandonment analysis was revisited. The re-analysis for MCR abandonment addresses human and equipment performance in lieu of the screening approach presented in the LAR. The same procedure is used for all fires leading to MCR abandonment; whether abandonment was due to loss of command and control or loss of MCR habitability. The re-analysis identified actions at the primary control station (PCS), local actions (actions away from the PCS), and equipment reliability/availability.

The ASD procedure uses Reactor Pressure Vessel (RPV) injection with High Pressure Coolant Injection (HPCI) as the preferred safe shutdown path with Residual Heat Removal (RHR) pump D in Suppression Pool Cooling (SPC). Should HPCI injection fail, RHR pump D is available for Low Pressure Coolant Injection (LPCI) injection providing the RPV is depressurized. A MCR abandonment tree was developed for ASD as shown in Figure 1 below.

The ASD HEPs for all actions, both at the PCS and away from the PCS, were developed using the Human Reliability Analysis (HRA) Calculator and the methods in NUREG-1921. HEPs for MCR abandonment scenario actions are provided in Table I below.

Equipment reliability/availability was determined for ASD systems and is provided in Table 2 below. The incorporation of ASD human actions and equipment performance into the model resulted in a MCR abandonment CCDP of approximately 0.108 as shown in the MCR abandonment tree (Figure 1). MCR abandonment CLERP was again taken to be the same as CCDP.

i. For some MCR abandonment scenarios, fire-induced spurious actuations can affect the preferred safe shutdown path. MCR abandonment procedures are in place to mitigate the impact of these potential spurious operations; for example, by removing power from Safety Relief Valves (SRVs) to prevent spurious opening leading to depressurization and the loss of HPCI, the preferred high pressure injection source. In addition, the conditional probability of spurious actuations is small compared to the overall system failure probabilities.

ii. Fire-induced recoverable failures can affect the preferred safe shutdown path for some MCR abandonment scenarios. MCR abandonment procedures provide recovery actions to restore the preferred safe shutdown path. These actions are included in the MCR abandonment model. Recovery Actions, such as local operation of HPCI valves or diesel generator alignment, are modeled independently from actions taken at the

NLS2014015 Page 6 of 25 PCS. The MCR abandonment model was constructed so that RAs must be successful for the system to perform its function. If the RAs fail, the system is immediately considered failed.

iii. Fire-induced spurious operations of equipment unrelated to the preferred safe shutdown path, but that might have a detrimental impact on the safe shutdown path, are addressed by the MCR abandonment procedure and are removed from service as necessary. Should operators fail to perform actions to mitigate detrimental impacts, and remote shutdown is not successful, the fire scenario results in core damage.

c.

In the initial LAR submittal, no values other than 0.1 were used for MCR abandonment scenarios. In the revised MCR abandonment model, other values for the MCR abandonment HEPs were used. Please refer to the discussion in b.(i), (ii), and (iii) above.

d.

Consistent with FAQ 07-0030 and FAQ 08-0054, in the MCR abandonment re-analysis, the base case identified actions performed at the PCS and actions performed locally.

HEPs were developed for each of the actions. Those actions performed away from the PCS were considered RAs. For the compliant case, human factors evaluations (HFEs) performed at the PCS retained the same HEPs as in the base case. However, the HEPs for RAs performed away from the PCS were set to success, i.e., failure probability of 0.

This is addressed in the response to PRA RAI 40.

Table 1: Main Control Room Abandonment Human Failure Events Recovery HFE Name HFE Description HEP Error Actions Factor ADS-XHE-FI-SORV-ASD OPERATOR FAILURE TO DEPRESSURIZE 3.30E-02 5

WITH SRVs FROM ASD OPERATOR FAILS TO START DG2 FROM RA EAC-XHE-FI-DG2BUS1G-HPCI DG2 ROOM & ALIGN TO BUS 1G FOR 3.20E-03 5

HPCI OPERATOR FAILS TO START DG2 FROM RA EAC-XHE-FI-DG2BUS1G-RHR DG2 ROOM & ALIGN TO BUS 1G FOR 6.80E-02 5

LPCI

°..................................

ECS-XHE-FI-TRANS-ASD OP FAILS TO MANUALLY INITIATE LPCI 1.30E-02 5

FROM ASD ECS-XHE-FI-TRANS-ASD-OPERATOR SUPPORTS LPCI INITIATION SUPPORT AWAY FROM ASD OPERATOR FAILS TO START AND HCI-XHE-FI-HPCIASD OPERATE HPCI FROM ASD AFTER MCR 5.10E-02 5

ABANDONMENT RA HCI-XHE-FI-HPCIASDSUPPORT OPERATOR FAILS TO SUPPORT ACTIONS 2.20E-02 5

FOR HPCI AWAY FROM ASD

NLS2014015 Attachment I Page 7 of 25 Recovery Error Actions HFE Name HFE Description HEP Factor OPERATOR FAILS TO INITIATE RHR-XHE-FI-RHRASD SUPPRESSION POOL COOLING AFTER 9.30E-04 10 MCR ABANDONMENT OPERATOR FAILS TO SUPPORT RA RHR-XHE-FI-RHRASDSUPPORT SUPPRESSION POOL COOLING 1.SOE-03 5

INITIATION AWAY FROM ASD Table 2: ASD Equipment Failure Probability Results RHR in LPCI RHR in SPC HPCI Failure SRV Failure Bus 1G Failure Mod il Mode Failure Mode Failure Mode Failure Probability Probability Probability Probability Probability 5.02E-02 7.87E-04 4.26E-02 2.20E-02 4.89E-02

NLS2014015 Attachment I Page 8 of 25 Figure 1: Main Control Room Abandonment Tree HPCI RPV Depressurization Bus 16 Powered by EDG 2 RHR Pump D in LPCI Mode RHR Pump D In SPC Mode Fee cP ERNV RNV Bus 16 W/

Bus 1G RHRDLPCI LPCI HFE L

E

)

Failure SPCS)

HPCI Failure HPCI HFE SRV Failure Dre Dpre EQ Failur W/V Bus 16w w/Dw2 WHE (RA)

Failure P LFaLPI HFE (RA)

SPC HFE (RA)

Fire Event Probability (PCs)

HPCI HFE (RA)

Probability (PCs)

(RA)

ProbabiRlty Probability s

wF(PCs)

Problit (CAPr Leading to KPs)

CLASS Prob MCR EAC-XNE-F4-EAC-X4-HE-FI-Abandonment HaXHE-FI-HC-XHE-FII-ADS-NHE-DG2BUS1G-DG2BUS1G-E6S-EAS-HEF-SPC-TRNB-RHR-XiiE-RHR-XHE-FI-Top-Ul HPASD HPASDSUPPORT Top-SO NA EAC-G-O NA HPCI RHR AS-T R

T 013 FI-RHRAST RHRASDSUPPORT A__ _ _

(4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />)

(35 minutes)

ASO SUPPORT 5.02E-02 5.1OE-02 2.20E-02 7.87E-04 3.30E-02 NA 4.26E-02 NA 3.20E-03 6.80E-02 2.20E-02 1.30E-02 4.60E-02 4.89E-02 9.30E-04 1.S0E-03 OK CD CD 4.31E-02 4.02E-02 OK CD 8.08E-03 CD 1.23E-02 CD 4.OOE-03 CCDP 1.08E-01 0.0338

NLS2014015 Attachment I Page 9 of 25 Supplement to PRA RAI 14.01c Response To provide additional information on the appropriateness of using a single value for the MCR abandonment the following approach was taken:

1. To illustrate that the exact same operator actions will be taken independent of the cause of MCR abandonment, a summary of the use of the CNS fire response procedures is provided below.

2. An additional, detailed review of the sequences leading to MCR abandonment was conducted.

3. The three major considerations in determining the CCDP/LERP are addressed.

Then, a final summary and conclusion are provided.

CNS Fire Response Procedures After a Fire Red Tile alarm (highest category of fire alarm) is received, Operators enter the 5.1 INCIDENT procedure. For ASD areas, this requires an immediate fire brigade response.

Then operators enter the 5.4POST-FIRE procedure and identify the specific fire area. If an ASD area is identified, operators transition to the 5.4FIRE-S/D procedure. At this point operators will continue to monitor plant parameters in the MCR. The Shift Manager determines if MCR abandonment is required. This determination is based on the occurrence of spurious operations and/or control room habitability. If MCR abandonment is ordered, operators then trip the plant, close main steam isolation valves, and inhibit the automatic depressurization system in the MCR.

Then the operations team initiates ASD actions in the 5.4FIRE-S/D procedure. This includes preventative measures for potential worse case spurious operations (SOs). Timing evaluations using worst case scenarios have been performed and validated. These actions will be taken independent of the reason for MCR abandonment and even if an SO(s) has not occurred. The procedural direction takes the operators directly to 5.4FIRE-S/D and its safe shutdown strategy regardless of the amount of spurious alarms or failed equipment contributions that occur.

In summary, there is a single ASD procedure, and it will be used for all MCR abandonment scenarios. The same actions will be taken independent of the scenario.

Detailed review of Scenarios leading to MCR Abandonment Before deciding on the use of a screening value of 0.1 in the Fire PRA used to support the LAR, scenarios were reviewed to establish that an overall value of 0.1 was appropriate. A summary is provided below.

The multiple spurious operations that appeared in the MCR abandonment cutsets include the following types of sequences: 1) multiple SRVs spuriously opening, 2) two main steam line (MSL) drain valves spuriously opening, 3) diesel generator start and load, and 4) Reactor Water Cleanup Boundary valves coupled with radwaste containment isolation valves. In the first case, even though multiple SRVs could be opening, an operator would treat this as a single RPV

NLS2014015 Page 10 of 25 depressurization event that needs to be addressed per procedure, as appropriate; and is included in the analysis as the actions affect timing for establishing the alternate shutdown path. In addition, the frequency is very low. For the case of two MSL drain valves both spuriously opening, the operators would again treat this as a single containment integrity event which is addressed per procedure; and is included in the analysis as the actions affect timing for establishing the alternate shutdown path. The frequency is also very low for this event. In the third case, only a single plant parameter is affected, diesel power to the critical bus. The actions were explicitly included in the HEP developed for the MCR abandonment event tree and are part of the model for the CCDP/CLERP determination of 0.108. In the fourth case, there are spurious operations on different systems. The fourth case was treated as a complex event. The complex event, however, had an insignificant impact on risk (less than 1%).

All other cutsets examined had at most a single spurious operation. These single spurious operations generally had a contribution to the scenario CCDP several orders of magnitude lower than the cutsets dominating any specific scenario.

In conclusion, the contribution of multiple spurious operations is very low and for specific, insignificant scenarios a higher CCDP/CLERP could have been used. The results and insights, however, would not change. This conclusion is further supported through quantitative evaluation that is summarized in the following discussion. The loss of command and control MCR abandonment fire scenarios were developed through a detailed analysis of potential fires and propagation paths (including both panel and transient fires). The level of damage was assessed using detailed circuit analysis and Nuclear Energy Institute (NEI) guidance on spurious and multiple spurious operations (NEI 00-01). After quantification, a detailed review of the cutsets for the potential fire scenarios involving MCR abandonment was conducted to identify potentially complex fire damage states. This cutset review allowed performance of a quantitative study that assumed a CCDP/CLERP of 1.0 for complex MSO fire damage states. The resulting increase in risk was found to be less than 1 percent of CDF/LERF.

Three Fire Scenario Categories Each category is addressed below.

Category 1: Scenarios where fire induced spurious actuations of equipment can affect the preferred shutdown path.

Response: For some scenarios there is the potential of fire-induced spurious operations of equipment unrelated to the preferred safe shutdown path, but that might have a detrimental impact on the safe shutdown path. These spurious operations are addressed by the MCR abandonment procedure and spuriously operating components are removed from service as necessary; for example, by removing power from SRVs to prevent spurious opening leading to RPV depressurization and the loss of HPCI, the preferred high pressure injection source. In addition, the conditional probability of spurious actuations is small compared to the overall system failure probabilities. See previous section above on insignificance of any such scenarios.

NLS2014015 Page 11 of 25 Category 2: Scenarios where the fire could cause recoverable failures to the preferred shutdown path. If every such failure, absent recovery, is assumed to immediately fail the success path, provide confirmation.

Response: These actions are included in the MCR abandonment model. RAs, such as local operation of HPCI valves or diesel generator alignment, are modeled independently from actions taken at the PCS. The MCR abandonment model was constructed so that RAs must be successful for the system to perform its function. If the RAs fail, the system is considered failed.

Category 3: Scenarios with fire induced failures unrelated to the preferred success path. Such failures can complicate the efforts to shut down the plant through, for example, spurious operations of unrelated equipment to the success path. If every such failure is assumed [to]

immediately fail the success path, please confirm this.

Response: Fire-induced SOs of equipment unrelated to the preferred safe shutdown path, but that might have a detrimental impact on the safe shutdown path, are addressed by the MCR abandonment procedure, and are removed from service as necessary. Should operators fail to perform actions to mitigate detrimental impacts, and remote shutdown is not successful, the fire scenario results in core damage. See previous section above on insignificance of such scenarios.

Summary and Conclusion The abandonment procedure is written to mitigate the worst fire possible including SOs; however, the operators do not need to diagnose the status of any required equipment. The number of spurious alarms and indications would have no impact on the operators once they leave the MCR. All procedure steps will be performed by the operators, and the operators will simply follow the steps in the procedure. Once outside the MCR, the status of the MCR is irrelevant to the success of the required actions.

The 0.108 value is appropriate as potential sequences where slightly higher values might be used have insignificant frequencies. This applies to loss of control and habitability scenarios equally.

In addition, fires related to the panels for which incipient detection will be provided are the same as loss of control scenarios if the incipient detection system were to fail to operate.

In addition, the following conservatisms in the analysis should be considered. First, the CLERP given core damage is assumed to be 1.0. Second, the refined analysis used above assumed alternating current (AC) power would be provided by an emergency diesel generator (EDG).

The preferred AC source is offsite power via the emergency transformer to the G emergency bus.

Thus, the use of an EDG and RAs to operate the EDG is conservative.

PRA RAI 16.02 FAQ 08-0054 (ADAMS Accession No. ML]10140183), discusses evaluating the additional risk of RAs and includes options to evaluate the change in risk associated with a VFDR and associated RAs. As discussed in the LAR, for Fire Area RB-A, cable damage to M923 represents

NLS2014015 Attachment I Page 12 of 25 two separate VFDRs: RBA-02 and RBA-03. Cable damage to M923 would result in failure of more than one PRA target. However, neither RBA-02 or RBA-03 included all failed equipment due to loss of the cable, and therefore their FREs do not appear to correspond to a physical event. The FPRA must be of sufficient technical adequacy to evaluate fire scenarios for use in performing FREs. Therefore, address the following:

a. The FRE for failure of this cable should be re-evaluated unless justification can be provided such that the cause and affect of the cable failure is accounted for in the analysis.

b. Consider if there are other VFDRs for which the FREs do not accurately model the physical impact of the fire, and re-evaluate or provide justification for those also.

c. If any LAR results change as a result ofparts a or b, discuss their impact.

d. Confirm that FAQ 08-0054 (ADAMS Accession No. MLl 10140183), and FAQ 07-0030 (ADAMS Accession No. ML110070485), guidance related to calculating both change in risk and the additional risk of RAs was followed. If not, provide justification for any differences and provide an assessment of the impact on the reported risk results of following the FAQ guidance.

NPPD Response

a.

The VFDRs RBA-02 and RBA-03 do consider cable damage to M923 separately, however the FRE for fire area RB-A also includes a VFDR RBA-ALL case. The ALL case combines all the fire area RB-A VFDRs into a single composite VFDR to address any interdependencies and synergisms. Therefore, composite VFDR RBA-ALL accounts for the cable failure in both RBA-02 (air-operated valve securing flow to Service Water through the credited Reactor Equipment Cooling system heat exchanger) and RBA-03 (motor-operated valve securing flow to Service Water through the credited RHR heat exchanger).

b.

In addition to fire area RB-A, fire areas RB-FN, RB-CF, and RB-M each include instances where multiple VFDRs, within each respective fire area, are impacted by the same cable failure but not all equipment is included in each VFDR. However, similar to fire area RB-A, each fire area includes a composite VFDR designated as XXX-ALL. The composite ALL case for each fire area consolidates all of the VFDRs for that fire area into a single composite VFDR. The delta risk values reported in LAR Attachment W, Table W-2, are based upon the composite VFDR XXX-ALL case for each fire area.

c.

There are no changes to the LAR results based on the evaluation performed for parts a and b above.

NLS2014015 Page 13 of 25

d.

The method for calculating both change in risk and the additional risk of RAs followed the guidance provided in both FAQ 08-0054 (see LAR, Section 4.2.4) and FAQ 07-0030 (see LAR, Section 4.2.1.3 and Att. G).

Supplement to PRA RAI 16.02 Response Performance of FREs for CNS NFPA 805 Transition To clarify the approach to the FREs, the following is provided:

1. For each fire scenario the equipment damaged is identified using fire modeling and cable/circuit analysis.

2. For each fire area the VFDRs are determined assuming complete fire area damage.

3.

Those components identified in the VFDRs as non-compliant are assessed to determine the approach to modeling the compliant case. There are two options considered for modeling compliance and one is selected:

a. If a recovery action is used, set the action to recover the non-compliant component to 100 percent success; set basic event in Fire PRA model to 0.0.

b. Set the component itself to 100 percent success for fulfilling its function (remove component failure due to fire damage).

4. The compliant case is quantified by setting the fire affected equipment to fail with the exception of the non-compliant components, which are either assumed to be protected or the RA success probability is 100 percent.

5. FRE delta risk was determined by subtracting the compliant case from the NFPA 805 case.

In summary all "attached" components are assumed to be failed in the NFPA 805 case; and the non-compliant components assumed to be protected or RA success probability is set to 100 percent. Also as has been previously noted for each fire area, an analysis is performed assuming all VFDRs are addressed. This addresses the potential for any synergies among the potentially affected components.

PRA RAI 35 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies that recovery actions must be addressed. The NRC staffis sample review of FREs noted that some HRA basic events appear to be credited; however do not appear in the LAR, Attachment G as credited recovery actions.

a. A review of the FREfor RB-M suggests that the recovery of SW-MOV-MO89B may be a credited RA for VFDR RBM-07; however, neither LAR Table G-1 (including revisions

NLS2014015 Page 14 of 25 made in response to SSD RAI-09) nor LAR Table B-3 credit the recovery ofSW-MOV-M089B as a RAfor VFDR RBM-07. Confirm whether or not the recovery ofSW-MOV-M089B is a RA that should be in Table G-J. If it is a RA, clarify if it is reflected in LAR Attachment W, or provide updated results ifnecessary.

b. A review of FREs for RBK-04 and TBA-05 appear to credit EAC-XHE-FI-SWEDG1 and SWS-XHE-FI-SWPA COMA; however, according to LAR Table G-1 and LAR Table B-3, no RAs are credited to meet RG 1.174 risk acceptance guidelines or defense-in-depth (DID) criteria for VFDRs RBK-04 and TBA-05. Clarify if these actions are credited in the LAR Attachment W results, and provide updated results if necessary.

c. Explain the cause of these apparent discrepancies and confirm that LAR Table G-J, as supplemented by the response for safe shutdown analysis (SSD) RAI-09, is a complete list of RAs for the LAR, (i.e., it includes RAs which may or may not be credited in the FPRA in performing the FREs.)

NPPD Response

a.

The recovery of SW-MOV-MO89B was included in the pre-NFPA 805 plant FPRA model as part of the VFDR RBM-07. The original risk was of the magnitude of IE-06 and was based on combining 4 VFDRs (RBM-2678). Sensitivities were performed for fire area RB-M, focusing specifically on VFDRs RBM-6 and RBM-7 which include SW-MOV-MO89B. These sensitivities were developed for the post-NFPA 805 plant configuration after implementation of the proposed plant modifications and procedural changes. As documented in the Delta Risk Calculation (NEDC 11-108 Attachment C, Section C.2), the sensitivity was performed by setting all of the RBM-6 and RBM-7 HFEs to guaranteed failure (1.0 failure probability). The Delta CDF and Delta LERF for this sensitivity were 2.16E-09 and 1.25E-09, respectively. Due to the low risk, it was determined that the RA SW-MOV-MO89B was no longer required in Area RB-M. The RA for SW-MOV-MO89B was not credited in Area RB-M in the post-NFPA 805 FPRA model and does not need to be included in Table G-1.

b.

VFDRs RBK-04 and TBA-05 are similar to RBM-07 from part a, except that modifications will be made as documented in LAR Attachment S, Table S-2 (S-2.1 and S-2.2). These modifications result in EAC-XHE-FI-SWEDG1 and SWS-XHE-FI-SWPACOMA no longer being RAs. Therefore, the RAs were not included in the LAR Attachment G, Table G-1.

c.

The LAR, Attachment G, Table G-l, as supplemented by the response for SSD RAI-09, was the complete listing of RAs at the time of submittal. However, there have been changes to this attachment due to modifications and proposed procedural changes in the plant that have created potential new RAs. The risk quantifications resulting from the response to PRA RAI 40 that will determine the risk of these actions has not yet been completed. Accordingly, an updated Table G-I will be provided in a future submittal, in a time frame mutually agreed to by NPPD and the NRC Staff.

NLS2014015 Attachment I Page 15 of 25 Supplement to RAI PRA-35.c Response As agreed upon between the NRC team and CNS during the follow-on audit, there are no additional RAs at this time, and accordingly, there is no need to provide an updated LAR Attachment G, Table G-1. During the implementation process, if a new VFDR is identified, it will be quantified using the FRE process. The risk quantifications for that VFDR will determine if it becomes an RA, or if there is no risk significance associated with the VFDR.

PRA RAI 40 The NRC staff identified several methods and weaknesses used in the FPRA that have not been accepted by the staff RAIs were provided about these methods and weaknesses and the responses have been reviewed. The staff has concluded that some of these methods and weaknesses are unacceptable in that justification does not seem to be technically available (e.g.,

credit for control power transformers is not supported by experiments).

Unacceptable methods and weaknesses:

Transient fire influence factors (LAR Supplement dated July 12, 2012)

" Treatment of kerite cables (PRA RAI-02b)

" Addition of Fire Area D W (PRA RAI-32)

Corrections associated with Fire Area TB-A (PRA RAI-36).

" Corrections associated with Fire Area RB-FN (PRA RAI-] 6e)

" Credit for controlpower transformers (PRA RAI-15)

" Changes to the recovery actions (PRA RAI-34)

The following methods and weaknesses have been identified, but the NRC Staff review is continuing with additional RAIs and further supporting information being requested.

Alternatively, any of these methods and weaknesses may be replaced with a method or model previously accepted by the NRC by modifying the FPRA.

Methods and weaknesses still under review:

Use ofprobabilities less than 1E-5 as thefloor HEP screening value (PRA-02-01)

" Credit for minim instrumentation after loss of RPV level instrumentation (PRA RAI 02c-01)

" Estimate of CCDPs including vented cable run atop the MCBs (PRA RAI 02f(i)-01)

Use of less than 317KWfor transient fires (PRA RAI 04-01)

Estimate of HEP/CCDPs following MCR abandonment (PRA RAI 11-01, 14-01) a) Please provide the results of a composite analysis that shows the integrated impact on the fire risk CDFLERF, ACDF, ALERF) after replacing all the unacceptable methods and weaknesses with acceptable ones. As the review process is concluded, additional changes to

NLS2014015 Page 16 of 25 replace any method or weakness still under review that are determined to be unacceptable may be required. In this composite analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed. For those cases where no synergy exists, a one-at-a-time analysis may be done. In the response, explain how the RG 1.205 risk acceptance guidelines are satisfied for the composite analysis and, if applicable, a description of any new modifications or operator actions being credited to reduce delta risk and the associated impacts to the fire protection program.

b) If any of the unacceptable methods or weaknesses will be retained in the PRA that will be used to estimate the change in risk ofpost transition changes to support self-approval, please explain how the quantitative results for each future change will account for the use of the unacceptable method or weakness.

NPPD Response

a. To demonstrate the integrated impact to CDF, LERF, ACDF, and ALERF, a composite fire PRA model was quantified (See Supplement 1 below). This analysis (a one-at-a-time analysis) was conducted and the potential for any synergistic impacts addressed conservatively. This model incorporated the items shown in Table 1 below.

The FPRA composite model CDF is 6.07E-5 per year and LERF is 1.26E-5 per year. ACDF and ALERF are -6.94E-6/year and -1.25E-5/year respectively. Thus the conclusions reached in the LAR have not changed. As noted in Supplement 2, considerable conservatism remains in these values.

The RG 1.205 risk acceptance guidelines were satisfied for the composite model. The total change in risk is consistent with the acceptance guidelines in RG 1.174. Note that the total change in risk for the composite model was negative. No new modifications or operator actions were credited in the development or quantification of the composite model, other than in TB-A (see response to PRA RAI 36).

b. No unacceptable methods will be retained in the reconciled model if there is any significance to results, conclusions and use. Each of the bullets listed under "Unacceptable methods and weaknesses" has been addressed (Please see Table 1.)

Table 1: Elements Incorporated into Composite Fire PRA Model RAI 1st Round Included PRA Parameter in PRA RAI 40 Sensitivity

RAI 40

PRA RAI 40 Notes Performed Composite Bullets Model Transient fire influence factors (LAR Y

1st set, 1st bullet LIC-109 Response Comment 3 Supplement dated July 12, 2012) 1 1

NLS2014015 Page 17 of 25 RAI 1st Round Included PRA

RAI 40

PRA RAI 40 Parameter in PRA RAI 40 Sensitivity C

pI B l Notes Prom d Composite Bullets PerformedMoe Model Treatment of kerite cables (PRA RAI-1st set, 2nd 02b) bullet Fire Area DW (Drywell) has always been included in the fire PRA model. DW contains no 1st set, 3rd VFDRs. DW has no impact on Addition of Fire Area DW (PRA RAI-32)

See note Ybulet det rs.

PR h as a dded bullet delta risk. PRA RAI-32 added entry for DW to Table W-2. No computations or changes to model were made.

Corrections associated with Fire Area 1st set, 4th TB-A (PRA RAI-36) bullet See response to RAI-i5 Corrections associated with Fire Area 1st set, 5th RB-FN (PRA RAI-16e) bullet Credit for control power transformers in determining Credit for control power transformers 1st set, 6th likeliood of spu rious (PRA RAI-15)

Y Y

bullet likelihood of spurious operations removed from model.

Changes to the recovery actions (PRA See note N

1st set, 7th Not necessary. No impact RAI-34) bullet Use of probabilities less than 1E-5 as 2nd set, 1st Not necessary. See response to the floor HEP screening value (PRA N

2nd let 1st Not n s

.S rp e

01) bullet PRA RAI 02.o.01 Credit for minimum instrumentation 2nd set, 2nd Not necessary. See response to after loss of RPV level instrumentation N

bullet PRA RAI 02.c.01 (PRA RAI 02c-01)

Insignificant to CDF/LERF and Estimate of CCDPs including vented 2nd set, 3rd bounded by Supplement 2 cable run atop the MCBs (PRA RAI Partially bullet approach to assessing delta 02f(i)-01)

CDF/LERF. Will be included in final model.

Use of less than 317KW for transient 2nd set, 4th Not necessary. See response to fires (PRA RAI 04-01)

N bullet PRA RAI 04.01 Estimate of HEP/CCDPs following MCR 2nd set, 5th abandonment (PRA RAI 03.01, 11.01, y

2nd let See Supplement 2 14.01) bullet

NLS2014015 Page 18 of 25 RAI 1st Round Included PRA Parameter in PRA RAI 40 Sensitivity

RAI 40

PRA RAI 40 Notes Performed Composite Bullets Model PRA RAI-11 was not explicitly included in the PRA RAI 40 request; however, MCR transients were considered in composite model and are MCR Transients (PRA RAI-11)

YPartially NA insignificant to CDF/LERF and bounded by Supplement 2 approach to assessing delta CDF/LERF. Will be included in final model.

Per the response to PRA RAI 19.01 a., Implementation Item S-3.30 addresses potential changes as follows:

Upon completion of all Fire PRA credited implementation items in Transition report Tables S-2 and S-3, verify the validity of the change-in-risk (total modifications) provided in Attachment W. If this verification determines that the risk metrics have changed such that the risk metrics from LAR Attachment W are exceeded, additional analytical efforts, and/or procedure changes, and/or plant modifications will be implemented to assure the RG 1.205 acceptance criteria are met.

Supplement 1 to Response to PRA RAI 40 (Composite Assessment)

An assessment has been completed. The assessment considered each item as discussed below.

1. CDF/LERF and delta CDF/LERF included in Attachment W of the CNS LAR are as follows:

a. CDF = 5.07E-5/year

b. LERF = 1.05E-5/year

c. Delta CDF = -8.71 E-6/year

d. Delta LERF = -1.29E-5/year

2. Transient Influence factors (based on the sensitivity evaluation previously conducted and provided in the LAR supplement dated July 12, 2012) have the following impact:

a. CDF/LERF increases by 2.54E-6/year and 1.51E-6/year respectively

b. Delta CDF/LERF increases by 6.1OE-8/year and 1.1E-8/year, respectively

3. Kerite (PRA RAI-02b Response)

a. CDF/LERF increases by about 4%/7%, respectively

NLS2014015 Page 19 of 25

b. Delta CDF/LERF not assessed as will affect compliant and post-NFPA cases comparably; for conservatism assumed the total increase is applied to delta risk

c. CDF increase is (5.07E-5/year + 2.54E-6) *.04 = 2.13E-6/year

d. LERF increase is (1.05E-5/year + 1.51E-6/year) *.07 = 8.41E-7/year

e. Delta CDF increase is 2.13E-6/year

f. Delta LERF increase is 8.41E-7/year

4. DW (PRA RAI-32)

a. CDF/LERF is about 1.27E-7/year and 1.27E-7/year, respectively, so the CDF/LERF is increased by these values

b. No VFDRs so no delta risk

5.

TB-A (PRA RAI-36)

a. CDF decreased by 2.8E-6/year (4.8E-5 minus (5.07E-5 + 1.27E-7))

b. LERF decreased by 1.9E-6/year (8.7E-6 minus (1.05E-5 + 1.27E-7))

c.

Delta CDF decreased by 3.3E-6/year (-1.2E-5 minus (-8.71E-6))

d. Delta LERF decreased by 2.1E-6/year (-1.5E-5 minus (-1.29E-5))

6. RB-FN (PRA RAI-16e)

a. CDF/LERF decreased 1E-8/year and I E-10/year, respectively

b. Delta CDF/LERF increased by 2.8E-7/year (1.59E-7 + 1.24E-7) and 5.8E-9/year (5.08E-9 + 7.03E-10), respectively

7. CPTs (PRA RAI-15)

a. CDF and LERF increased by 7E-6/year and 5E-7/year, respectively

b. Delta CDF/LERF increased I E-6/year and 0, respectively

8. Recovery Actions (PRA RAI-34): No changes

9. MCR Abandonment Refined HEPs (RAI 14.01) Please see Supplement 2 which documents the following conservative analysis:

a. Less than a 10% change in CDF and LERF so a net increase of about I E-6/year for each

b. Delta CDF/LERF increase of 1.6E-6/year

10. CDF and LERF are estimated as follows:

a. Total calculated CDF increases by 2.54E-6/year + 2.13E-6/year + 1.27E-7/year -

2.8E-6/year - I E-8/year + 7E-6/year + 1 E-6/year = I.OE-5/year, and thus yielding a calculated CDF of 6.07E-5/year.

NLS2014015 Page 20 of 25

b. Total calculated LERF increases by 1.51 E-6/year + 8.41 E-7/year + 1.27E-7/year -

1.96E-6/year - I E-10/year + 5E-7/year + I E-6/year = 2.08E-6/year, and thus yielding a calculated LERF of 1.26E-5/year.

11. Delta CDF and Delta LERF are estimated as follows:

a. Delta CDF = -8.71E-6/year + 6.1E-8/year + 2.13E-6/year + 0 - 3.3E-6/year +

2.8E-7/year + 1 E-6/year + 1.6E-6/year = -6.94E-6/year, and thus delta CDF is still negative

b. Delta LERF = -1.29E-5/year + 1.1E-8/year + 8.41E-7/year + 0 - 2.1E-6/year +

5.8E-9/year + 0 + 1.6E-6/year = -1.25E-5/year, and thus delta LERF is still negative Supplement 2 to Response to PRA RAI 40 (MCR Abandonment)

The response to PRA RAI 14.01 developed CCDP and CLERP using a detailed analysis including RAs and equipment availability and reliability. As noted in the response to PRA RAI 14.01 to develop the delta risk the recovery actions will be assumed to have an HEP of zero for the compliant case. This bounds the delta risk. The calculation is provided below as follows:

• First, the top event probabilities are recalculated assuming the HEPs for recovery actions are zero

" Second, each core damage sequence is recalculated to develop a CCDP/CLERP

" Third, the CCDPs/CLERPs for each sequence are summed to develop a total CCDP/CLERP

" Finally, the impact on delta risk is developed using a bounding approach Note: several significant digits are used simply to illustrate the math.

I1.

Top Event Probabilities with RA HEPs Set to Zero HPCI Failure probability =.0502 +.9498 *.0510 =.0986 compared to the base case value of

.1185 Success probability =.9014 compared to a base case value of.8815 RPV Depressurization Failure probability =.000787 +.999213 *.033 =.0338 compared to a base case of 0.0338 Success probability =.9662 Bus 1 G Powered from EDG2 (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> available)

Failure probability =.0426 compared to a base case of.0456 Success probability =.9574

NLS2014015 Page 21 of 25 Bus 1 G Powered from EDG2 (35 minutes available)

Failure probability =.0426 compared to a base case of. 1077 Success probability =.9574 RHR Pump D in LPCI Mode Failure probability =.022 +.978 *.013 =.0347 compared to a base case of.0791 Success probability =.9653 RHR Pump D in SPC Mode Failure probability =.0489 +.9511 *.00093 =.0498 compared to a base case of.0512

2.

CCDP/CLERP Sequence Calculations CD Sequence 1 =.9014 *.9574 *.0498 =.0430 CD Sequence 2 =.9014 *.0426 =.0384 CD Sequence 3 =.0986 *.9662 *.9574 *.0347 =.0032 CD Sequence 4 =.0986 *.9662 *.0426 =.0041 CD Sequence 5 =.0986 *.0338 =.0033

3.

Summation The total CCDP/CLERP is.0919. This compares to a base case of. 108. These are effectively the same value as judging the uncertainty in such calculations. RAs are not significant.

4.

Delta Risk Using a bounding approach the delta risk increase is developed as follows:

A bounding MCR abandonment frequency of 1 E-4/year is used The change in CCDP/CLERP =.016. Therefore the delta CDF/LERF is 1.6E-6/year.

There remains considerable conservatism in this estimate as follows:

Conditional LERP given core damage is assumed to be 1.0 Per the response to PRA RAI 03:

In reviewing the latching mechanism for cabinets that VEWFDS is being installed, it was noted that the Relay Panels 9-32 and 9-33 latch at the top and bottom, but not at the center handle and may not meet the criteria of a "robustly secured" cabinet per Frequently Asked Question (FAQ) 08-0042. Based on the risk significance of these panels, NPPD has decided to modify the panel doors to include additional mechanical latching around the perimeter of the panel doors, which will ensure that the doors are fully and mechanically secured, and will not create openings or gaps due to warping during an internal fire, in accordance with FAQ 08-0042. There

NLS2014015 Page 22 of 25 are five (5) additional relay panels 9-30, 9-39, 9-41, 9-42, and 9-45 located in the Auxiliary Relay Room that are of similar latch construction.

Therefore, NPPD is modifying these five relay panels similar to the modification identified for the 9-32 and 9-33 panels (see Attachment 2, Change 2).

Accordingly, per the response to PRA RAI 03, NPPD will robustly secure several cabinets which will reduce the potential for MCR abandonment. These cabinets do not need to be sealed to achieve a compliant configuration. The delta risk reduction associated with this commitment is not included.

Supplement 3 to PRA RAI 40 Fire PRA Self-Approval Model CNS has enhanced the descriptions in Table 1, as noted below, to clarify the contents. In addition, in Table 2 below CNS provides the Fire PRA modeling topics and their status prior to using the Fire PRA to support self-approval.

Table 1 (Revised): Elements Incorporated into Composite Fire PRA Model RAI 1st Round Included PRA Parameter in PRA RAI 40 Sensitivity

RAI 40

PRA RAI 40 Notes Performed Composite Bullets Model Transient fire influence factors (LAR NA Y

1st set, 1st bullet LIC-109 Response Comment 3 Supplement dated July 12, 2012)

Treatment of Kerite cables (PRA RAI 1st set, 2nd Included as described in 02b nY of blet, 2supplement 1. CDF/LERF and 02b)

Ybullet delta CDF/LERF are addressed.

Fire Area DW (Drywell) has always been included in the fire PRA model. DW contains no VFDRs. DW has no impact on Addition of Fire Area DW (PRA RAI 32)

NA (See note)

Y1st set, 3rd delta risk. PRA RAI-32 added bullet entry for DW to Table W-2. No computations or changes to the model were made.

Included as described in Supplement 1.

s.u..... e.e.n.t....:

Corrections made. Included as Corrections associated with Fire Area 1st set, 4th described in Supplement 1.

TB-A (PRA RAI 36)

Y bullet CDF/LERF and delta CDF/LERF are addressed.

NLS2014015 Attachment I Page 23 of 25 RAI 1st Round Included PRA Parameter in PRA RAI 40 Sensitivity

RAI 40

PRA RAI 40 Notes Performed Composite Bullets Model Corrections made. Included as Corrections associated with Fire Area 1st set, 5th described in Supplement 1.

RB-FN (PRA RAI 16e) bullet CDF/LERF and delta CDF/LERF are addressed.

Credit for control power transformers in determining likelihood of spurious Credit for control power transformers y

y 1st set, 6th operations has been removed (PRA RAI 15) bullet from the analysis in Supplement 1 CDF/LERF and delta CDF/LERF are addressed.

Changes to the recovery actions (PRA See note Y

1st set, 7th Not necessary, as there are no See note bullet changes in RAs.

Not necessary as approach is Use of probabilities less than 1E-S as 2nd set, 1st correct. See response to PRA the floor HEP screening value (PRA RAI y

bullet RAI 02.o.01. HRA was 02.o.01) conducted in accordance with NUREG-1921.

Credit for minimum instrumentation 2nd set, 2nd Not necessary as modeling is after loss of RPV level instrumentation

¥ bullet correct. See response to PRA (PRA RAI 02.c.01)RA02c1 Insignificant to CDF/LERF and Estimate of CCDPs including vented 2nd set, 3rd bounded by Supplement 1 cable run atop the MCBs (PRA RAI y

bullet approach to assessing delta 02f(i).01)

CDF/LERF. Will be included in final model as noted in Table 2.

D....................................

Use of less than 317KW for transient Y

2nd set, 4th Approach is justified in fires (PRA RAI 04.01) bullet response to PRA RAI 04.01.

See Supplement 2 for analysis Estimate of HEP/CCDPs following MCR 2nd set, 5th of compliant and variant case abandonment (PRA RAI 03.01, 11.01, y

bullet change in CCDP/CLERP.

14.01)

Supplement 1 includes delta risk.

NLS2014015 Page 24 of 25 RAI 1st Round Included PRA Parameter in PRA RAI 40 Sensitivity

RAI 40

PRA RAI 40 Notes Performed Composite Bullets Model PRA RAI-11 was not explicitly included in the PRA RAI 40 request; however, MCR transients were considered in the composite model and are insignificant to CDF/LERF and MCR Transients (PRA RAI 11) y Y

NA bounded by Supplement 1 approach to assessing delta CDF/LERF. Will be included in final model. Addressed as described in Supplement 1 using bounding approach for delta risk.

Table 2 Fire PRA Self-Approval Model Fire PRA Model Prior to Using Fire PRA Model Changes for Self-Approval Transient fire influence factors (LAR See Note 1 below. No fractional Supplement dated inluenye 1a s

2

)

transient influence factors will be Supplement dated July 12, 2012) included in the Fire PRA model.

Treatment of Kerite cables (PRA RAI Note 2 below documents the treatment 02b) of Kerite cables in the Fire PRA model.

DW fire scenario explicitly included in Addition of Fire Area DW (PRA RAI 32) the Fire PRA model.

Corrections associated with Fire Area Corrections explicitly included in the TB-A (PRA RAI 36)

Fire PRA model.

Corrections associated with Fire Area Corrections explicitly included in Fire RB-FN (PRA RAI 16e)

PRA model No credit will be given for control Credit for control power transformers power transformers in determining hot (PRA RAI 15) short / spurious operation likelihoods in the Fire PRA model.

Changes to the recovery actions (PRA There are no new recovery actions.

RAI 34)

Use of probabilities less than 1E-5 as the Method meets NUREG-1921, and this floor HEP screening value (PRA RAI method will continue to be used.

02.o.01)

Credit for minimum instrumentation Included and no change from previous after loss of RPV level instrumentation model.

(PRA RAI 02.c.01)

Estimate of CCDPs including vented Fire scenarios incorporating the vented cable run atop the MCBs (PRA RAI cable run atop the MCBs will be 02f(i).01) included in the Fire PRA model.

NLS2014015 Page 25 of 25 Fire PRA Model Prior to Using Fire PRA Model Changes for Self-Approval Transient fire heat release rates less Use of less than 317KW for transient than 317 kW are included in the Fire fires (PRA RAI 04.01)

PRA model as identified in PRA RAI 04 and PRA RAI 04.01.

Estimate of HEP/CCDPs following MCR Refined MCR abandonment model abandonment (PRA RAI 03.01, 11.01, (with a present value of 0.108) will be 14.01) included in the Fire PRA model.

Fire scenarios incorporating MCR MCR Transients (PRA RAI 11) transients will be included in the Fire PRA model.

There are no non-suppression probability values less than 1.OOE-03 in Non-suppression probabilities minimum the Fire PRA model. The IVICR value of 1E-3 (PRA RAI 06)

Abandonment non-suppression probability values were revised as identified in PRA RAI 06.

The Fire PRA will explicitly include Uncertainty (PRA RAI 02.h.01) state-of-knowledge correlation.

Note 1 The fractional values less than I utilized for transient fire frequency influencing factors for proposed modifications/enhancements to the combustible and hot work controls for Fire Zone 8A (Auxiliary Relay Room) and Fire Zone 9A (Cable Spreading Room) were replaced with values consistent with the guidance of NUREG/CR-6850. The 0.1 multiplier for crediting enhanced transient controls for the three specific locations within fire zones Fire Zone 3C and 3D, the area located above the TIP Room on the 903'-6" Elevation of the Reactor Building, and Fire Zone 2C, the floor areas immediately located around Instrument Racks 25-5 and 25-6 on the 931 '-6" Elevation of the Reactor Building has been removed in the Fire PRA analysis.

Note 2 The fire scenario frequencies (including severity factors and non-suppression probabilities) and target impacts were revised based on the reduced failure criteria of Kerite FR cables. This revision utilized the final test results of the Kerite cable testing issued in NUREG/CR-7102, "Kerite Analysis in Thermal Environment of FIRE (KATE-Fire)" which recommends that Fire PRA applications assume a nominal fire-induced failure threshold of Kerite FR cables of 247°C. Because NUREG/CR-7102 does not contain information on a damaging heat flux value, the thermoplastic damaging heat flux threshold has conservatively been used in the fire modeling.

NLS2014015 Page 1 of 12 Revisions to the Cooper Nuclear Station License Amendment Request To Adopt National Fire Protection Association Standard 805 Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants This attachment provides changes to the National Fire Protection Association (NFPA) 805 License Amendment Request, as revised, based on the Nuclear Regulatory Commission (NRC) audit conducted at Cooper Nuclear Station (CNS) on January 28 and 29, 2014. The changes are presented in underline/strikeout format.

1.

Section 5.5, "Transition Implementation Schedule," is revised to state:'

The Nebraska Public Power District (NPPD) proposes a twelve-month implementation period for the transition of CNS to the new fire protection licensing basis. Pursuant to this, the following activities are planned:

" NPPD eemmits to will complete the specific actions identified in Implementation Items S-3.1 through S-3.29 of Table S-3 of Attachment S to Enclosure 1, within twelve months after issuance of the NFPA 805 License Amendment.

" NPPD will complete implementation of the required modifications identified in Table S-2 of Attachment S to the Transition Report prior to startup from the first refueling outage greater than 12 months following the issuance of the NFPA 805 License Amendment. Appropriate compensatory measures will be maintained until modifications are complete.

" NPPD will implement Implementation Item S-3.30 of Table S-3 of Attachment S to Enclosure 1 by May 31, 2017.

Basis for Change: During the January 28 and 29, 2014, audit, the NRC requested that the Implementation Items on the S-3 table be included under the "Transition License Conditions" of the revised License Condition 2.C(4). As a result, these actions are no longer commitments, but enforceable regulatory obligations.

2.

Attachment M, "License Condition Changes," Section "Transition License Conditions" is revised to read:

Transition License Conditions

1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2) below, risk-informed changes to NPPD's fire protection program may not be made without prior NRC review and approval unless the change has been 1 Section 5.5 was previously changed in NLS2013104 (ADAMS Accession Number ML13353A073).

NLS2014015 Page 2 of 12 demonstrated to have no more than a minimal risk impact, as described in (2) above.

2) 3)

The licensee shall implement the following modifications to its facility, as described in Table S-2, "Plant Modifications Committed," of NPPD letter NLS2014015, dated February 18, 2014, to complete the transition to full compliance with 10 CFR 50.48(c) a. proI I in Table S 2 of the Cooper Nuclear S

.tation L.icen.e Amendment Request dated April 27, 2012, prior to startup from the first refueling outage greater than 12 months following the issuance of the License Amendment. The licensee shall maintain appropriate compensatory measures in place until completion of these modifications.

Tho licene,-e shall maintain apprpFriate compensatoF;Fmeasures in place until completion of the modifications deloneated above. The licensee shall implement the items S-3.1 through S-3.29 as listed in Table S-3. "Implementation Items," of NPPD letter NLS2014015, dated February 18, 2014, within twelve months after issuance of the License Amendment.

4) The licensee shall implement item S-3.30 as listed in Table S-3, "Implementation Items," of NPPD letter NLS2014015, dated February 18, 2014, no later than May 31,2017.

Basis for Change:

Paragraph 2 - Modifications S-8 and S-9 were added in letters subsequent to issuance of the LAR. Rather than include these letters in the License Condition, it is more straightforward to provide a reference to this letter, which provides a consolidated list of the Table S-2 modifications (see Change 3).

Paragraph 3 - The compensatory measures are more appropriately included with Paragraph 2 since they relate to the Table S-2 modifications. During the onsite audit conducted January 28 and 29, 2014, the NRC requested that the Table S-3 Implementation Items be included in the License Condition, so that they would become enforceable regulatory obligations, rather than NRC commitments. Accordingly, Implementation Items S-3.1 through S-3.29 are added.

Paragraph 4 - During the onsite audit conducted January 28 and 29, 2014, the NRC requested that the Table S-3 Implementation Items be included in the License Condition, so as they would become enforceable regulatory obligations, rather than NRC commitments. Accordingly, Implementation Item S-3.30, which has a different implementation date than S-3.1 through S-3.29 included in Paragraph 3, is established in this paragraph.

NLS2014015 Page 3 of 12

3.

Consolidated Tables S-2 and S-3, as revised, are provided:

Table S-2 Plant Modifications Committed Item Rank Problem Statement Proposed Modification In Comp Risk Informed Characterization FPRA Measure S-2.1 High Control cables associated Cables to be re-routed such Y

Y Risk is reduced as change provides RB-K with 4kV circuit breakers 1 FA that they do not traverse the for AC power availability without an and 1FS and first-level G Critical Switchgear Room.

RA. Defense-in-depth is improved.

undervoltage circuitry are routed from the A Non-Compensatory measure for NFPA Critical Switchgear Room to 805: Appropriate compensatory the F Critical Switchgear measures will be established per CNS Room, but pass through the Procedure 0.23, as required, until the G Critical Switchgear Room.

modification is implemented.

In addition, control and feeder cables associated with Compensatory measure for 4kV circuit breakers 1 FA and 10 CFR 50 Appendix R: Yes.

1 FS are also routed similarly.

Alternate compensatory measures in the form of Operator Manual Actions as documented in Procedure 5.4 Post-Fire, Attachment 18, are in place for this issue. The alternate compensatory measures are implemented per CNS Procedure 0.23.

S-2.2 Hiah Control cables associated Cables to be re-routed such Y

Y Risk without creditina all current RA is TB-A with 4kV circuit breakers 1 FA and 1 FS and first-level undervoltage circuitry are routed along the same path as the 1GB and 1GS through the Non-Critical Switchgear Room.

that they do not traverse any fire scenarios common with the control cables associated with the 1GB and 1 GS breakers in the Non-Critical Switchgear Room.

reduced as change provides for ac power availability without RA for design basis scenarios. However, some existing RA provide for restoring power for beyond design basis scenarios.

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS

NLS2014015 Page 4 of 12 Table S-2 Plant Modifications Committed In Comp Rs nomdCaatrzto Item Rank Problem Statement Proposed Modification FPRA Measu Risk Informed Characterization Procedure 0.23, as required, until the modification is implemented.

S-2.3 CB-A Medium Feeder cables to 1 a and 1 B 125 and 250V DC Battery Chargers are routed along the same path through the Control Building Controlled Corridor resulting in a loss of all battery chargers.

Cables MLX36 and MLX37 to be re-routed through the RPS Rooms to provide a minimum of a channel of battery chargers to support long term battery usage.

Y Compensatory measure for 10 CFR 50 Appendix R: Yes.

Alternate compensatory measures in the form of Operator Manual Actions as documented in Procedure 5.4 Post-Fire, Attachment 24, are in place for this issue. The alternate compensatory measures are implemented per CNS Procedure 0.23.

Y Risk is reduced as DC power availability is improved. Defense-in-depth is improved.

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS Procedure 0.23, as required, until the modification is implemented.

Compensatory measure for 10 CFR 50 Appendix R: Yes.

Alternate compensatory measures in the form of Operator Manual Actions as documented in Procedure 5.4 Post-Fire, Attachment 3, in place for this issue. The alternate compensatory measures are implemented per CNS Procedure 0.23.

NLS2014015 Page 5 of 12 Table S-2 Plant Modifications Committed Item Rank Problem Statement Proposed Modification In Comp Risk Informed Characterization FPRA Measure S-2.4 High Control Room abandonment Install incipient detection in Y

Y Risk is reduced considerably as the CB-D is required along with the Panel 9-32 and 9-33 in the installation reduces the frequency of usage of the alternate Auxiliary Relay Room (Fire Control Room abandonment.

shutdown procedures for a Zone 8A) allows for shutdown Defense-in-depth is improved.

fire in Panel 9-32 or 9-33.

from the Control Room with minimal field actions.

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS Procedure 0.23, as required, until the modification is implemented.

Compensatory measure for 10 CFR 50 Appendix R: None. Fire Area CB-D is deterministically compliant with 10 CFR 50 Appendix R.

S-2.5 Medium Transient fires impacting Install conduit shields to Y

Y Risk is reduced as equipment RB-M multiple trains of conduit outside the Critical Switchgear Room common passageway.

prevent damage to conduit banks from transient fires in Fire Area RB-M/Fire Zone 3C.

available to provide core cooling is protected. Defense-in-depth is improved.

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS Procedure 0.23, as required, until the modification is implemented.

Compensatory measure for 10 CFR 50 Appendix R: Yes.

Alternate compensatory measures in the form of Operator Manual Actions as documented in Procedure 5.4 Post-Fire, Attachment 19, are in place for this issue. The alternate

NLS2014015 Page 6 of 12 Table S-2 Plant Modifications Committed Item Rank Problem Statement Proposed Modification In Comp Risk Informed Characterization FPRA Measure compensatory measures are implemented per CNS Procedure 0.23.

S-2.6 Medium Transient fires impacting Install bottom tray covers to Y

Y Risk is reduced as equipment RB-M vertical cable trays in corner prevent damage to cable tray available to provide core cooling is of RB-M for opposite train, risers from transient fires in protected. Defense-in-depth is Fire Area RB-M/Fire Zone improved.

3C.

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS Procedure 0.23, as required, until the modification is implemented.

Compensatory measure for 10 CFR 50 Appendix R: Yes.

Alternate compensatory measures in the form of Operator Manual Actions as documented in Procedure 5.4 Post-Fire, Attachment 19, are in place for this issue. The alternate compensatory measures are implemented per CNS Procedure 0.23.

S-2.7 Medium LNK 6 fire impacts conduits Installation of board shielding Y

Y Risk is reduced as equipment CB-D and trays in vicinity requiring for cable trays and conduit to available to provide core cooling is Control Room abandonment prevent damage from fires protected. Defense-in-depth is along with usage of the involving panel PMIS-MUX-improved.

alternate shutdown procedure.

LNK6 and PMIS-MUX-LNK7 in the Cable Spreading Room (Fire Zone 9A).

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS Procedure 0.23, as required, until the modification is implemented.

NLS2014015 Page 7 of 12 Table S-2 Plant Modifications Committed In Comp Item Rank Problem Statement Proposed Modification FPRA Measure Risk Informed Characterization Compensatory measure for 10 CFR 50 Appendix R: None. Fire Area CB-D is deterministically compliant with 10 CFR 50 Appendix R.

S-2.82 Low Fire damage to some Motor Install safe and reliable N

Y Some risk benefit would be realized Operator Valves (MOVs) will not allow the valve to be operated from the Control Room. Currently, there are compensatory measures to remove the control power fuse(s) at the MCC or DC starter and pushing the contactor to open or close MOV.

controls for the MOVs at or near their respective MCC or DC starter. The controls will provide position indication, a control switch, an isolation switch and control power fuses to each affected MOV control circuit. This will be implemented with CED 6033461.

because of simplifying the recovery action from removing control fuses and pushing contactors to operate from a remote control panel. The Fire PRA does not quantify the difference in the risk between the different recovery actions.

Compensatory measure for NFPA 805: Appropriate compensatory measures will be established per CNS Procedure 0.23, as required, until the modification is implemented.

Compensatory measure for 10 CFR 50 Appendix R: Yes.

Alternate compensatory measures in the form of Operator Manual Actions as documented in Procedures 5.4FIRE S/D and 5.4POST-FIRE are in place for this issue. The alternate compensatory measures are implemented per CNS Procedure 0.23.

2 Item S-2.8 was added in NLS2013011 (ADAMS Accession Number ML13018A006).

NLS2014015 Page 8 of 12 Table S-2 Plant Modifications Committed Item Rank Problem Statement Proposed Modification In Comp Risk Informed Characterization FPRA Measure S-2.94 High Control Room abandonment Install additional mechanical Y

Y Risk is reduced considerably as the is required along with the latching around the perimeter installation reduces the frequency of usage of the alternate of the panel doors of Relay Control Room abandonment.

shutdown procedures for fires Panels 9-30, 9-32, 9-33, 9-Defense-in-depth is improved.

in Relay Panels 9-30, 9-32, 39, 9-41, 9-42, and 9-45 in 9-33, 9-39, 9-41, 9-42, and 9-the Auxiliary Relay Room Compensatory measure for NFPA

45. These panels latch on the (Fire Zone 8A) to provide 805: Appropriate compensatory top and bottom, but not at the robustly secured cabinets measures will be established per CNS center handle, and may not preventing propagation of fire Procedure 0.23, as required, until the meet the criteria of a outside the panels. This modification is implemented.

"robustly secured" cabinet allows for shutdown from the per FAQ 08-0042 Control Room with minimal Compensatory measure for field actions.

10 CFR 50 Appendix R: None. Fire Area CB-D is deterministically compliant with 10 CFR 50 Appendix R.

Table S-3 Implementation Items Item Description LAR Section / Source S-3.1 During the implementation of the NFPA 805 licensing basis, performance-Attachment A based surveillance frequencies will be established as described in Electric Power Research Institute (EPRI) Technical Report 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features". The performance-based surveillance frequencies will be evaluated in the monitoring program in accordance with NFPA 805 FAQ 10-0059.

3 Item S-2.9 was added in NLS2013016 (ADAMS Accession Number ML13051A539).

NLS2014015 Page 9 of 12 Table S-3 Implementation Items Item Description LAR Section / Source S-3.2 Enhanced transient and combustible controlled zones will be established in 4.5 and Attachment W high risk Fire Zones 8A and 9A. Enhanced transient and combustible controlled locations will be established in the following specific fire zone locations to address high risk transient fire scenarios: Fire Zone 2C above the TIP Room, and Fire Zones 3C and 3D in the areas around instrument racks 25-5 and 25-6.

S-3.3 Post-fire operating procedures will be updated to reflect new NSCA Attachments G and V strategies and training performed as necessary.

S-3.4 Technical, Operations, and administrative procedures and documents that 4.3.2 and Attachment D relate to non-power modes of plant operating states will be revised as needed for implementation of NFPA 805.

S-3.5 The Fire Protection Design Basis Document described in Section 2.7.1.2 of 4.7.1 NFPA 805 and necessary supporting documentation described in Section 2.7.1.3 of NFPA 805 will be created.

S-3.6 A confirmatory demonstration (field validation walk-through) of the 4.2.1.3 and Attachment G feasibility for the credited NFPA 805 RA will be performed. This will include field validation of:

(1) Transit times (i.e., travel times to/from recovery action manipulated plant equipment).

(2) Execution times (i.e., time required to physically perform the action, such as handwheel a valve open, open a breaker, etc.).

(3) Communications for adequacy between the controlling location and RA locations for areas which involve actions.

(4) Adequate lighting (either fixed or portable) for access/egress and local lights are provided for the component to be operated.

S-3.7 CNS calculations will be reviewed and updated based on the results of the 4.2.1.3 and Attachment G field walkdowns of the recovery actions from Implementation Item S-3.6.

S-3.8 The FPP procedures will be revised to specify application of the NFPA 805 Section 4.7.3 Section 2.7.3 quality requirements.

NLS2014015 Page 10 of 12 Table S-3 Implementation Items Item Description LAR Section / Source S-3.9 Procedure 0.23 will be revised to identify the Authority Having Jurisdiction Attachment A for the various areas of the Fire Protection Program.

S-3.10 Administrative procedures will be revised to control the use of portable Attachment A electric heaters and revised to document that portable fuel-fired heaters are not permitted in plant areas containing equipment important to nuclear safety, or where there is a potential for radiological release resulting from a fire.

S-3.11 Emergency procedures will be updated to allow use of Service Water (SW)

Attachment W pumps alone to provide cooling to Residual Heat Removal (RHR) heat-exchangers in the event RHR SW booster pumps are rendered unavailable.

S-3.12 Procedures will be revised to require new cable installations to meet the Attachment A requirements of IEEE-383, or similar.

S-3.13 Procedure 0.7.1 will be revised to include requirement that bulk gas Attachment A storage not be allowed inside structures housing systems, equipment, or components important to nuclear safety.

S-3.14 Procedures will be revised to include the requirement for the inspection of Attachment A the transformer spill containment area.

S-3.15 For personnel performing fire modeling or Fire PRA development and 4.7.3 evaluation, NPPD will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work.

S-3.16 Procedures will be revised to inventory which pre-fire plans are contained Attachment A in the fire lockers, and ensure that updates of the pre-fire plans include replacing the updated pages in each of the inventoried locations throughout the plant.

NLS2014015 Page 11 of 12 Table S-3 Implementation Items Item Description LAR Section I Source S-3.17 Procedures will be revised to ensure that pre-fire plan drawings are Attachment A maintained in the Control Room and to ensure that the latest revisions are available.

S-3.18 The fire brigade training program will be updated to include guidance to Attachment A ensure fire drills are conducted in various plant areas, especially in those areas identified to be essential to plant operation and to contain significant fire hazards.

S-3.19 Not-u6ed. The Fire PRA analysis will be updated, prior to performing self-NLS2014015, PRA RAI 40, Rev. 1 approval evaluations using the Fire PRA, to incorporate the model changes identified in Supplement 3 of the response to PRA RAI 40.

S-3.20 Pre-fire plans and training materials will be revised to address radioactive Attachment E release requirements of NFPA 805.

S-3.21 Revise Procedure 0.7.1 to include the requirement for ventilation duct Attachment A materials to be non-combustible or listed by a nationally recognized testing Laboratory, such as Factory Mutual or Underwriters Laboratory, Inc., for flame spread index of 25 or less and a smoke development index of 50 or less.

S-3.22 Procedures will be revised to ensure that the fire protection system is not Attachment A to be used for non-emergency usage.

S-3.23 The NFPA 805 Monitoring Program will be developed and implemented, as Section 4.6 described in Section 4.6.

S-3.24 The Fire PRA database will be controlled as an electronic document in the Attachment V same way the Internal Events PRA model (CAFTA model) is controlled.

S-3.25 Fire Zone 9A (Cable Spreading Room) and Fire Zone 8A (Auxiliary Relay Attachment V Room) will be designated as enhanced transient and hot work controlled fire zones.

S-3.26 The CNS Updated Safety Analysis Report will incorporate the applicable Section 5.4 subject matter described in FAQ 12-0062, at a level of detail consistent with NEI 98-03, "Guidelines for Updating Final Safety Analysis Reports."

NLS2014015 Page 12 of 12 Table S-3 Implementation Items Item Description LAR Section I Source S-3.27 The configuration control procedures which govern the various CNS Section 4.7.2 documents and databases that currently exist will be revised to reflect the new NFPA 805 licensing bases requirements.

S-3.28 Control procedures and processes for NSCA supporting information, Non-Section 4.7.2 Power Mode Review, Fire Modeling Calculations, Fire Safety Assessments, risk evaluations, etc., will be developed.

S-3.29 System level Design Criteria Documents will be revised to reflect the NFPA Section 4.7.2 805 role that the system components now play.

S-3.304 Upon completion of all Fire PRA credited implementation items in NLS2013104, PRA RAI 19.01 Transition report Tables S-2 and S-3, verify the validity of the change-in-risk (total modifications) provided in Attachment W. If this verification determines that the risk metrics have changed such that the risk metrics from LAR Attachment W are exceeded, additional analytical efforts, and/or procedure changes, and/or plant modifications will be implemented to assure the Regulatory Guide 1.205 acceptance criteria are met.

Basis for Change: During the NRC onsite audit conducted January 28 and 29, 2014, a consolidated version of Tables S-2 and S-3 was requested which included changes made since the original submittal of the LAR. Additionally, Item S-3.19 was added at the request of the NRC during the audit in order to ensure the Fire PRA model will be appropriately updated before evaluating risk-informed changes for self-approval.

4 Item S-3-30 added in NLS2013104 (ADAMS Accession Number ML13353A073).