L-2014-071, Response to Request for Additional Information License Amendment Request No. 216 - Transition 10 CFR 50.48(c)-NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants(2001 Edition)

Response to Request for Additional Information License Amendment Request No. 216 - Transition 10 CFR 50.48(c)-NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants(2001 Edition)

ML14113A176
Person / Time
Site:	Turkey Point
Issue date:	04/04/2014
From:	Kiley M Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2014-071
Download: ML14113A176 (103)

Text

April 4, 2014 FPL. 10 CFR 50.90 L-2014-071 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Re: Turkey Point Nuclear Generating Station Units 3 and 4 Docket Nos. 50-250 and 50-251 Response to Request for Additional Information Regarding License Amendment Request No. 216 - Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

By Florida Power and Light Company (FPL) letter L-2012-092 dated June 28, 2012, in accordance with the provisions of 10 CFR 50.90, "Application of License or Construction Permit," FPL requested an amendment to the Renewed Facility Operating License (RFOL) for Turkey Point Nuclear Generating Station Units 3 and 4. The license Amendment Request (LAR) will enable FPL to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision 1 of Regulatory Guide (RG) 1.205.

On February 10, 2014, the NRC Staff requested additional information regarding the LAR. The attachment to this letter provides the response to the request for additional information (RAI).

The additional information does not impact the 10 CFR 50.92 evaluation of "No Significant Hazards Consideration" previously provided in FPL letter L-2012-092.

This letter adds the following new commitments:

Turkey Point will replace the Reactor Coolant Pump (RCP) seals with Flowserve RCP seals. As documented in the attached RAI Probabilistic Risk Assessment (PRA) 29 response, a new plant modification item reflecting this commitment has been added to FPL letter L-2012-092, Attachment S, Table S-2, Plant Modifications Committed. In addition, a new implementation item is added to Attachment S, Table S-3, Implementation Items, to confirm consistency between the PRA model logic used for modeling the Flowserve seals with the Topical Report for these seals, upon its approval by NRC.

The attached RAI PRA 18.01 response added a new implementation item to FPL letter L-2012-092, Attachment S, Table S-3, Implementation Items, to revise procedures for Control of Transient Combustibles, and Control of Ignition Sources, to restrict storage of transient combustibles/ignition sources in specific locations of the plant.

The attached RAI PRA 29 response added a new implementation item to FPL letter L-2012-092, Attachment S, Table S-3, Implementation Items, to implement all changes discussed into the PRA model and that these changes will be evaluated in accordance with the PRA standard to determine the need for a focused-scope peer review. Any findings as a result of a focused-scope peer review, if required, will be resolved before self-approval of post-transition changes.

Florida Power & Light Company 9760 SW 344 St Homestead, FL 33035

L-2014-071 Page 2 This letter deletes the following commitments:

Installation of the FlowServe RCP seals allows elimination of Item 24 of FPL letter L-2012-092, Attachment S, Table S-2, Plant Modifications Committed. The attached RAI PRA 29 response provides a mark-up of FPL letter L-2012-092, Attachment S, Table S-2, Plant Modifications Committed reflecting deletion of Item 24.

As previously discussed in FPL letter L-201 3-164 in response to RAI PRA 01 .r, credit for the incipient detection (very early warning fire detection system (VEWFDS)) in the main control room has been eliminated. The attached RAI PRA 29 response provides a mark-up of FPL letter L-2012-092, Attachment S, Table S-2, Plant Modifications Committed reflecting deletion of Items 3 and 4 in their entirety, since in-cabinet detection is no longer credited, as discussed in the response to RAI PRA 01.r.02.

This letter revises implementation item 18 to FPL letter L-2012-092 Attachment S, Table S-3, Implementation Items, regarding update of the Fire PRA model. The revised implementation item requires review of the results of the final updated fire PRA after all modifications and procedural changes are complete and as-built and all implementation items affecting the fire PRA results are complete. These results are to be compared to the final updated version in the LAR after resolution of all RAIs and treated as a change evaluation in accordance with procedure EN-AA-202-1004 which includes the RG 1.205 risk acceptance criteria guidelines and provides a path to request a license amendment should those limits be exceeded.

If you should have any questions regarding this application, please contact Robert Tomonto, Licensing Manager, at 305-246-7327.

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

Executed on April 4, 2014.

Michael Kiley Vice President Turkey Point Nuclear Generating Station Attachment cc: Regional Administrator, Region H, USNRC Senior Resident Inspector, USNRC, Turkey Point USNRC Project Manager for Turkey Point Ms. Cindy Becker, Florida Department of Health

Attachment to L-2014-071 Page 1 of 101 L-2014-071 ATTACHMENT FPL's Turkey Point Nuclear Power Plant Units 3 and 4 NFPA 805 LAR RAI

Response

PRA 01.a.01 PRA 01.v.01 PRA 01.d.01 PRA 01.y.01 PRA 01.dd.01a PRA 01.z.i.01 PRA 01.dd.01b PRA 01.z.ii.01 PRA 01.e.01 PRA 03.01 PRA 01J.i01 PRA 07.01 PRA 01.j.01 PRA 08.01 PRA 01.I.01 PRA 11.01 PRA 01.m.01 PRA 13.01 PRA 01.m.02 PRA 13.02 PRA 01.p.01 PRA 16.01 PRA 01.q.01 PRA 18.01 PRA 1 .r.01 PRA 19.01 PRA 01.r.02 PRA 22.01 PRA 01 .t.01 .a PRA 23.01 PRA 01 t01 .b PRA 27.e.01 PRA 01 1t.0c PRA 28 PRA 01.t.01.d PRA 29 PRA 01 t01 .e

Attachment to L-2014-071 Page 2 of 101 RAI PRA 01.a.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 01 .a and indicated that the use of guidance specified in frequently asked questions (FAQ) 12-0064, "Hot Work/Transient Fire Frequency: Influence Factors," results in an increase in Unit 3 large early release frequency (LERF); however, a decrease in LERF is reported for Unit 4. Provide an explanation for this asymmetry.

In addition, the Unit 3 and 4 Fire Risk Evaluations for Fire Area HH indicate that a factor of 0.01 was applied to transient fire scenarios to credit the additional administrative controls (e.g.,

"physical barriers") to reduce the probability of a transient fire. If additional credit for administrative controls to reduce the likelihood of a transient fire is taken beyond that specified in FAQ 12-0064, then provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, using FAQ-12-0064 or other accepted methods, and identify the methods used.

RESPONSE

The asymmetry is due to an error in the calculation where the complete set of transient scenarios was not used in the risk summation for the updated Unit 4 LERF calculation which incorporated the updated transient ignition frequency. This resulted in'an incorrect statement that the Unit 4 LERF total decreased, when in fact it increased. The effect of the updated transient influence factors on the total pant risk, in accordance with FAQ 12-0064 for all plant areas, will be incorporated into the aggregated risk analysis results to be submitted subsequent to completion of the second-round RAIs.

The treatment of transient ignition frequency for Fire Area HH will be updated to be in accordance with FAQ 12-0064. A 'low' influence factor in the transient ignition frequency calculation for Fire Area HH will be applied while eliminating the 0.01 factor applied for addressing the probability of failure of administrative controls.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 d round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 01.d.01 In letter dated March 18, 2013, the licensee responded to PRA RAI 0l.d and discussed parameter uncertainty rather than addressing sources of LERF modeling uncertainty. Identify and characterize the LERF sources of model uncertainty and related assumptions.

RESPONSE

The primary sources of LERF model uncertainty are the same as those associated with the CDF model uncertainty since the core damage model is a primary input to the LERF model. See the response to RAI PRA 27.e.01 for a discussion of the internal events PRA CDF model uncertainty and its impact on the Fire PRA.

Attachment to L-2014-071 Page 3 of 101 RAI PRA 01.dd.01.a In letter dated May 15, 2013, the licensee responded to Fire Modeling RAI 01 .r and defined a larger ZOI for enclosures where hot gas layer effects are of concern; however, in letter dated April 16, 2013, the licensee responded to PRA RAI 0l.dd and indicated that while the provisions for a larger ZOI exist, the current analysis does not employ them.

a. Provide clarification on the statement that "a larger ZOI [is] used in order to allow credit for a longer time to hot gas layer."

RESPONSE

The generic zone of influences (ZOIs) used to identify impacted targets for a fire scenario require a hot gas layer (HGL) analysis to evaluate the time at whichthe ZOI: begins to be affected by the HGL. The initial evaluation uses ZOls associated with an 80'C-HGL temperature. The HGL analysis treats the fraction of fires that are not suppressed prior to reaching the HGL limitation for the ZOI as impacting all targets within the fire zone (full room burnout). This type of treatment is conservative and bounding.

In situations where this treatment produces high risk results, a larger ZOI can be. used in conjunction with a higher HGL temperature. The larger ZOI incorporates a lower damage heat flux limit and allows for a larger heat flux input from the HGL (higher. HGL temperature) with the combination resulting in a damage heat flux. The expanded ZOI coupled with the elevated temperature HGL results in a heat flux at the ZOI boundary that is equal to the damage threshold for thermoplastic cables (5.7 kW/m 2) in this case. This is essentially a trade-off between the ZOI and the HGL temperature (evaluation of higher hot gas layer temperatures will increase the time to hot gas layer which will decrease the manual suppression non-suppression probability) which is useful for scenarios where a larger ZOI does not significantly increase the risk but where the risk due to a hot gas layer scenario is significantly higher. The application of a higher HGL temperature allows more time for fire suppression prior to the room reaching the HGL temperature. For instance, a thermoplastic ZOI associated with a 5.7 kW/m2 damage heat flux can be replaced with a 3 kW/m 2 damage heat flux. In doing so, the HGL temperature .used to evaluate the ZOI limitation is now 131 'C, which will produce a longer time to HGL and therefore a smaller fraction of fires contributing to HGL. This additional refinement, more realistic analysis, is applied where it will provide a significant reduction in the scenario risk.

This approach was used in the Turkey Point Cable Spreading Room for ZOI and HGL evaluations for transient fires which did not impact secondary combustibles.

RAI PRA 01.dd.01.b In letter dated May 15, 2013 (ADAMS Accession No. ML13157A011), the licensee responded to Fire Modeling RAI 01 .r and defined a larger ZOI for enclosures where hot gas layer effects are of concern; however, in letter dated April 16, 2013 (ADAMS Accession No. ML13109A008), the licensee responded to PRA RAI 01 .dd and indicated that while the provisions for a larger ZOI exist, the current analysis does not employ them.

b. Clarify why the risk analysis does not use the larger ZOIs and describe how this is consistent with acceptable methods or provide updated risk results as part of the

Attachment to L-2014-071 Page 4 of 101 aggregate change-in-risk analysis requested in PRA RAI 29, addressing this apparent non-conservatism.

RESPONSE

As discussed in the response to RAI PRA 01 .dd.01 .a, the larger (Zones of Influence) ZOIs are used as a method for reducing the fraction of fires that contribute to HGL. The use of either of the two ZOI-HGL temperature combinations is a conservative approach for evaluating risk impacts of a fire scenanio.

The approach taken to use the ZOI associated with a 131 'C HGL threshold in the Turkey Point Cable Spreading Room and the ZOI associated with an 80'C HGL everywhere else in the plant is conservative and does not constitute a new method. Therefore, the Turkey Point Fire PRA does not need to be updated with the new ZOIs.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with all second round RAI responses will be provided following NRC initial review and feedback on this response.

RAI PRA 01.e.01 In letter dated April 16, 2013, the licensee responded to PRA RAI 01.e.iii and stated that "a walkdown was conducted of all accessible fire zones to ensure that all barriers credited are of substantial build and capable of confining heat and products of combustion" and that inaccessible zones were verified using fire barrier drawings; however, the criteria utilized and the basis that each fire area meets these criteria are not provided. Provide a table of all the non-rated barriers credited in the PRA and justification for this credit. Alternatively, describe the specific criteria utilized during walkdowns and review of plant drawings to arrive at the conclusion that barriers were "substantial", and provide justification for each physical analysis unit (PAU) that describes how these criteria were met.

RESPONSE

The walkdown criteria considered any physical analysis unit (PAU), fire zone boundary, without visible openings (opening through thickness of barrier that can be confirmed as open by visible inspection) as a substantial barrier. Any such barrier would be sufficient to prevent the transmission of radiant heat and flow of hot gases. Fire zone boundaries including visible openings or fire zones which were separated from adjacent fire zones by an imaginary line which facilitated fire zone definition were not credited as being capable of truncating the zone of influence of an ignition source. For any such barriers, targets within the zone of influence which extended via the opening to an adjacent zone were included in the list of impacted targets. The Multi-Compartment Analysis assumes a 1.0 failure probability between fire zones with normally open boundaries. The results of the walkdown, and a review of the Fire Hazards Analysis fire barrier description, are documented in Table 2-2 of the Multi-Compartment and Hot Gas Layer Analysis. Table 2-2 identifies the assumed barrier failure probability for each boundary separating adjacent fire zones in the "Barrier Failure Probability" column. As discussed in Section 5.2.2, a 1.0 barrier failure probability identifies barriers in which openings are present.

Attachment to L-2014-071 Page 5 of 101 RAI PRA 01.i.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 01.i and stated that a subset of circuit failure probabilities within the fire PRA (FPRA) lacks a circuit analysis basis and that "[a]ll components that were. not subject to a circuit analysis will have corresponding hot-short-induced spurious operation credit removed in a future revision of the Fire PRA documentation." Provide updated risk results as part of the current aggregate change-in-risk analysis requested in PRA RAI 29 after removing hot-short-induced spurious operation credit from components without a documented basis substantiated by circuit analysis (i.e., by setting the failure probabilities to 1.0).

RESPONSE

All credited circuit failure probabilities in the Turkey Point Fire PRA will be updated to provide a circuit analysis basis or will be removed (i.e., set to a failure probability of 1.0).

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI, will be provided following NRC initial review and feedback onthis response.

RAI PRA 01.j.01 Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides change in risk guidelines that are to be applied to the mean estimates of the change in risk associated with a proposed change. In letter dated May 15, 2013, the licensee responded to PRA RAI 01 .j but did not provide the requested confirmation that the values in LAR Attachment W are mean values, and the licensee seemed to imply that the values are point estimates but that some sensitivity study has been done. Nor does the response include the requested information about which types of parameters were considered correlated. Clarify which types of parameters were considered correlated. Additionally, as part of the aggregate change-in-risk analysis requested in PRA RAI 29, provide updated risk results that can be characterized as mean values for comparison with the RG 1.174 acceptance guidelines, and explain how those estimates are developed.

RESPONSE

The values provided in the License Amendment Request (LAR) Attachment W are point values.

A Monte Carlo parametric uncertainty analysis has been performed using EPRI's UNCERT software. The mean value associated with this analysis has been determined to be no more than 1% higher than the point values. Therefore, use of the point value is considered acceptable. The final quantification, post-round-2 RAI-resolution uncertainty analysis will incorporate the correlation of the following fire-specific parameters in the parametric uncertainty analysis:

1. Ignition Frequencies

2. Non-Suppression Probabilities

3. Severity Factors

4. Hot-Short-Probability-Related Altered Events

Attachment to L-2014-071 Page 6 of 101 Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2n6 round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 01.1.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 01.1 and provided a summary of a method to assign screening values to human error probabilities (HEPs) that is described as "similar" to the method in NUREG-1921, "EPRI/NRC-RES [Electrical Power Research Institute/NRC Office of Research] Fire Human Reliability Analysis Guidelines," to assign screening values. As noted in the part of the RAI response, when the two methods are compared, there are substantive differences between the "PTN ["PTN" is the licensee's acronym for Turkey Point 3 and 4] HRA [human reliability analysis] method" and "that specified in NUREG-192 1."

For example:

" The PTN HRA method is inherently less conservative than the NUREG- 1921 screening approach given that values below 0.1 and multipliers below 10 are applied.

"The PTN HRA method does not address recommended screening criteria for Sets 1, 2, and 3 outlined by NUREG- 1921. For example, the PTN HRA method does not explicitly consider possible damage to safe shutdown equipment as indicated in NUREG- 1921 for Set 1 screening values.

"While some modifications were made to account for the general categories of time available, accessibility, and complexity through multipliers, the performance shaping factors (PSFs) explicitly addressed through the flowcharts documented in the Human Failure Evaluation report do not encompass all of those PSFs that the NUREG- 1921 scoping approach characterizes as capable of inducing "significant variability in crew performance and response times."

• The NUREG-1921 scoping approach is predicated on the performance of a feasibility assessment, which, according to the RAI response," is not explicitly prescribed" by the PTN HRA method.

• The PTN HRA method considers the time available to perform an action; however, it does not appear to consider the time required to perform the action (i.e., the concept of time margin).

" Multipliers are directly applied to detailed HEPs from the internal events PRA (IEPRA),

whereas NUREG-1921 sets an HEP value of I E-3 for the base fire scenario in which the conditions represent the best possible for the fire context. In this manner, NUREG- 1921 deems 1 E-3 to be the best achievable HEP in the scoping fire HRA approach.

Furthermore, the response to the RAI failed to provide the requested sensitivity study where the new method is replaced by an acceptable method. Use an accepted method, such as NUREG- 192 1, to estimate HEP values, and use the method in the aggregate change-in-risk analysis requested in PRA RAI 29, or provide the following response in order for the staff to continue its review of the new method:

Attachment to L-2014-071 Page 7 of 101

a. Provide a complete description of this new method and its associated assumptions, addressing, at a minimum, the above differences.

b. Detail the steps performed to support the implementation of this new method.

c. Provide the results of a sensitivity analysis (i.e., core damage frequency (CDF), LERF, ACDF and ALERF) using screening/scoping approaches in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," and/or NUREG- 1921 to quantify HEPs or use the acceptable methods in the aggregate change-in-risk assessment requested in PRA RAI 29.

RESPONSE

The Turkey Point Fire Human Reliability Analysis is being revised to address the above concerns.

The new methodology is consistent with that specified in NUREG-1921. The revised analysis methodology will be incorporated into the post-second-round-RAI quantification effort.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 01.m.O1 In letter dated May 15, 2013, the licensee responded to PRA RAI 01 .m and stated that the PRA HRA assumes complete dependency amongst screening/scoping HEPs. While this treatment is conservative for the post-transition plant, it may produce non-conservative results for ACDF and ALERF when applied to the compliant case (e.g., in cases where an HEP is associated with initiation of a mitigating system made available in the compliant plant due to resolution of variances from deterministic requirements (VFDRs)). Explain whether this treatment is conservative considering the fact that conservative estimates of the compliant plant risk lead to underestimates of the change in risk when subtracted from realistic estimates of the variant plant risk. If the change-in-risk evaluation is underestimated, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, removing this conservatism from the compliant case, and explain how these estimates are developed.

RESPONSE

The Turkey Point Fire Human Reliability Analysis and joint HEP methodology willbe revised to incorporate the methodology defined in NUREG- 1921. This change will eliminate the screening/scoping HEPs and will apply the same methodology used for HEP dependency that is used in the Full-Power Internal Events PRA.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

Attachment to L-2014-071 Page 8 of 101 RAI PRA 01.m.02 In letter dated May 15, 2013, the licensee responded to PRA RAI Ol.m and indicated that dependency between multiplier-adjusted HEPs (i.e., those for which dependency was addressed using the EPRI calculator) and screening-based HEPs was not considered. Discuss how dependency between these two types of human failure events (HFEs) was addressed. If not considered, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, evaluating these dependencies.

RESPONSE

The Turkey Point Fire Human Reliability Analysis and joint HEP methodology will be revised to incorporate the methodology defined in NUREG-1921. This change will eliminate the multiplier-adjusted HEPs and will apply the same methodology used for HEP dependency that is used in the Full-Power Internal Events PRA.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 01.p.01 In letter dated May 15, 2013, the licensee responded to PRA RAI Ol.p but did not clarify whether a delay to damage or combustible ignition, as discussed in the disposition to LAR Attachment V, Table V-3, Fact and Observation (F&O) 9-5, is assumed for targets impacted by a high-energy arcing fault (HEAF). If such a delay is credited, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, treating targets within the zone of influence (ZOI) for HEAF scenarios as damaged and/or ignited at time 0 per Appendix M of NUREG/CR-6850.

RESPONSE

The Turkey Point Fire PRA will assume that all HEAF fires contribute to the hot gas layer scenario for the associated fire zone with all components and cables damaged and no credit for a time delay.

No ZOI will be applied to evaluate the impacted targets. Table I identifies those fire zones where HEAF ignition frequency was postulated and how it is applied to all the targets in the fire zone.

Attachment to L-2014-071 Page 9 of 101 Table 1.

Fire Zones with HEAF Ignition Frequency Contributions Fire Zone Treatment of HEAF Ignition Frequency 025 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

067 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

068 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

070 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

071 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

093 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

094 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

095 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

096 Applied full HEAF ignition frequency to hot gas layer fraction summation in the Hot Gas Layer report.

134 A base scenario is used to quantify fire zone 134, which applies the total fire zone ignition frequency (including the HEAF contribution) to the base fire zone CCDP.

A base scenario is used to quantify fire zone 139, which applies the total fire zone ignition frequency (including the HEAF contribution) to the base fire zone CCDP.

No credit will be taken in the updated Fire PRA, based on input from the PSL NFPA 805 audit, for a fire in the initiating compartment or for a multi-compartment fire scenario, based on the above summary, for delay of damage or combustible ignition due to HEAF fires. This revision to the Fire PRA, contrary to the original Fire PRA approach or the approach identified in responses to the first round RAIs, will not credit any delay in damage or ignition of combustibles due to HEAF fires.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers, including the treatment of HEAF scenarios as specified by this RAI will be provided following NRC initial review and feedback on this response.

RAI PRA 01.q.01 Justify that the collection period for fire protection system reliability and availability data for only the last 3 years is representative of plant-specific operating experience, and provide a basis for excluding past data. In addition, confirm that credited fire detection and suppression systems have not experienced outlier behavior relative to system unavailability that would indicate that actual systems are unavailable more frequently than would be indicated by the generic values utilized. If generic values are not representative of plant-specific operating experience, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29.

Attachment to L-2014-071 Page 10 of 101

RESPONSE

The period for data collection was chosen based on availability of data in a readily retrievable format. This data is considered representative of current and older data. System unavailability is not considered a concern given the compensatory actions that are implemented when systems are unavailable. No indication of outlier behavior was identified in the review. Therefore, the generic detection and suppression system failure probabilities are considered appropriate.

RAI PRA 01.r.01 Given that the approach in Appendix L was not originally employed and that the responses to PRA RAI 01.r in letter dated May 15, 2013, and PRA RAI 08 in letter dated May 15, 2013, do not describe the fire scenarios postulated for the main control board (MCB) in detail, discuss the fire scenarios postulated for the MCB, and provide their quantified results. Additionally, discuss whether the Bin 4 frequency is further apportioned by panel or scenario as Attachment H of the Fire Scenario report and the response to PRA RAI 08 appear to indicate. If so, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, applying the full Bin 4 frequency to each MCB scenario postulated per Appendix L of NUREG/CR-6850.

Alternatively, provide justification that the overall approach for evaluating MCB fires bounds the results (i.e., CDF, LERF, ACDF and ALERF) that would be obtained had an approved method (e.g., Appendix L of NUREG/CR-6850) been employed.

RESPONSE

The main control board scenarios are being revised to eliminate in-panel detection credit and to combine the MCB scenarios to address the application of the total MCB frequency in conjunction with the use of the Appendix L factor. Two sub-scenarios are defined with each scenario: a non-abandonment and an abandonment (-A) scenario. The MCB scenario is the scenario associated with a fire at the MCB for the unit quantified while the OPMCB scenario is the scenario for the opposite unit's MCB. The total Bin 4 frequency is used for each scenario. The NSP column below represents the NUREG/CR-6850, Appendix L factor for a zero distance. The SF column represents the control room abandonment probability. The CCDP for the MCB case is a 1.0 CCDP x annual plant availability. The highest CCDP for the original panel scenarios was applied to the single-panel MCB scenario for application of the total MCB frequency to the Appendix L factor.

For the opposite unit, the maximum CCDP for the opposite-unit impact on the quantified unit is used for the OPMCB scenario. For the delta risk calculation, the 5.6E-02 compliant case CCDP is substituted for the MCB-PTB-A case and the OPMCB-PTB-A case, without alteration (since the CCDP is less than 5.6E-02) is added to it. Only the abandonment scenarios contribute to the compliant case risk calculation. This update of the main control board quantification is consistent with the guidance of NUREG/CR-6850. The quantification below is considered preliminary since the quantification may be impacted by other RAI responses.

Attachment to L-2014-071 Page 11 of 101 NSP OCDF (per 7ZONE . Scen IGF . NSP SF . CCDP reactor-year),

106 MCB-PTB 2.78E-03 9.00E-03 8.68E-01 2.17E-05 106 MCB-PTB-A 2.78E-03 9.00E-03 4.17E-03 8.68E-01 9.04E-08 106 OPMCB-PTB 2.78E-03 9.00E-03 4.83E-04 1.21 E-08 106 OPMCB-PTB-A 2.78E-03 9.00E-03 4.17E-03 4.83E-04 5.03E-11 Total Risk 2.18E-05 Total Delta Risk 5.88E-09 RAI PRA 01.r.02 In letter dated May 15, 2013, the licensee responded to PRA RAI 01.r and PRA RAI 08 and indicated that a 0.19 non-suppression probability (NSP) credit is applied to MCB and non-MCB fire scenarios for panels with in-cabinet smoke detectionin lieu of credit for incipient detectioh.

Address the following:

a. Given that the MCB and some main control room (MCR) electrical panels have open backs, provide technical justification to support the conclusion that an additional 5 minutes of advanced warning is available for each panel to which the 0.19 NSP credit is applied.

b. The 0.19 NSP credit does not address the availability and reliability of the in-cabinet smoke detectors. Provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, accounting for detector availability and reliability, and justify the estimate used.

c. Propagation between electrical (non--MCB) panels within the MCR was excluded from the original analysis and is not addressed as requested in PRA RAI 01 .r. Explain and justify how propagation (including between open-back panels) is addressed. If propagation was not considered, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, considering propagation between non-MCB panels within the MCR.

d. The response to PRA RAI 01 .r appears to indicate that smoke detectors are credited to limit the extent of internal fire damage within electrical (non-MCB) panels they monitor; however, per Appendix P of NUREG/CR-.6850, the 0.19 NSP credit may only be applied to targets beyond the ignition source (e.g., adjacent panels, cables trays, etc.). If smoke detectors are credited to prevent damage to non-MCB panels, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, only applying the 0.19 NSP credit to targets beyond the ignition source (e.g., adjacent panels, cable trays, etc.).

RESPONSE

a. Per RAI PRA 01 .r.01, credit for in panel detection is being removed from the MCB and open back panels. The credit for the 0.19 NSP will be maintained in closed panels that

Attachment to L-2014-071 Page 12 of 101 impact adjacent panels as needed. In-panel detection will not be credited for precluding damage in the panel in which it is installed.

b. Where the 0.19 NSP factor is credited, for a fire impacting multiple panels, the NUREG/CR-6850, Appendix P, failure probability of Halon suppression is used (there is no halon in the Turkey Point control room, this NSP is used to derive a bounding detection failure probability) as the bounding value for the failure probability of the detection system, 0.05. This means that the percentage of fires that are extinguished prior to escaping the panel is equal to the factor for in-panel detection times the success ratio plus the probability the smoke detector will fail to respond, i.e. 0.19*0.95+1 *0.05=0.2305. The remainder of the frequency will be included in the non-severe scenario. The failure of the in-panel detection system will be included in the final quantification in this manner.

c. Scenarios have been created for the back panels in which adjacent panels are assumed to be impacted from fire propagation. Additionally, transient scenarios between the open back panels are assumed to damage nearby Bin 15 panels. The total impact of these scenarios will be determined as part of the final quantification.

d. In-panel detection credit will be taken for scenarios associated with fire propagation between panels. In panel detection will not be credited for precluding damage in the panel in which it is installed.

See Attached LAR Attachment C and Attachment S markup for removal of incipient detection, associated with the response to the original RAI (RAI PRA 01 .r).

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

Attachment to L-2014-071 Page 13 of 101 Security-Related Information - Withhold Under 10 CFR 2.390 Florida Power & Light Attachment S - Plant Modifications and Items to be Completed Table S-2 Plant Modifications Committed DELETE nit Problem Statement Proposed Modification InRisk-Informed FPRA Measure RikIoreCarcrzan Characterization 3 High 3 Reduce risk contribution by installing Fire Area MM - Install incipient Yes Yes Allows early detection which Incipient detection, detection in panels 3C01, 3C02, provides 3C03, 3C04, 3C05, 3C06, 3QR8, additional risk benefits for CDF and 3QR20, 3QR42. 30R44, 3QR46, LERF.

3QR47, 3QR50. 3QR51. ComPensatory Measure: Incipient detection is designed to detect combustion byproducts prior to actual combustion (i.e., a fire). The Control Room is a continuously occupied space; therefore, detectionz of a fire would occur as soon as reasonably practical due to either smoke/odor from a fire or failure of panel functions, No additional compensatory measures are required.

4 High 4 Reduce risk contribution by installing Fire Area MM - Inslall incipient Yes Yes Allows early detection which incipient detection, detection in panels 4C01, 4C02, provides additional risk benefits for 4C03, 4C04, 4C05, 4C05, 4QR8. CDF and LERF.

4QR20, 4QR26, 4QR42, 4OR44. Com*.nsatporv Measure: Incipient 4QR46, 4QR47, 4QR50 4QR51, detection is designed to detect 4QR66. combustion byproducts prior to actual combustion (i.e., a fire). The Control Room is a continuously occupied space; therefore, detection of a fire would occur as soon as reasonably practical due to either smokelodor from a fire or failure of panel functions. No additional compensatory measures are required.

Page S-7 Revision 00 Revision Page S-7

Attachment to L-2014-071 Page 14 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: *L (Unit 4) - Units 3 and 4 Auxillary Building Fen Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone I Description Syterm Systam Protection Feature Required Fire Protection Feature and System Details 028 Units 3 and 4 Auxiliary Building None None E Combustible Loading: E Fan Room - Fire Area L Physical separation: E Fire Area ID: LL (Unit 3) - Unit 3 A DC Equipment Room Compliance Basis; NFPA 805. Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 104 Unit 3 ADC Equipment Room - None R, b R, S, N Detection Sysiem, Alarm Point 8: R 0 Fire Area LL ERFBS, 104-1: R S N ERFBS, 104-2: R S N ERFBS, 104-3: R S N ERFBS, 104-4: R S N ERFBS, 104-6: R S N Fire Area ID: LL (Unit 4) - Unit 3 A OC Equipment Room Compliance Basis: NFPA 805, Seclion 4,23.3 (a). 3-hr rated ERFBS Required Required Suppression Detection Required Fire Fire Zone ID Description System Sptem Protection Feature Required Fire Protection Feature and System Details 104 Unit 3 ADC Equipment Room - None None S, N ERFBS, 104-1: S N Fire Area LL ERFBS, 104-2: S N ERFBS, 104-3: S N ERFBS, 104-4: S N ERFOS. 104-5: S N Fire Area ID: MM (Unit 3) - Control Roam Complex Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protcteion Feature Required Fire Protection Feature and System Details (AI) Area Wide None None 097 Units 3 and 4 Mechanical None D 0 Detection System. Alarm Point 5: D IDELETE Equipment Room - Fire Area MM -I -

106 Units 3 and 4 Control Room - Fire None -r

ýj ED D Detection System. Alasn. Point 5: D Area MM Det ion sem. Mo, i -M:

Extinguishers: D Fire Safety Analysis Data Manager (4,1) Tu~rkey Point Run: 0612(V2012 13:39 Page: 21 of 54

Attachment to L-2014-071 Page 15 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: MM (Unit 3) - Control Room Complex Compliance Basis: NFPA 805, Section 4,2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and SysteM Details 106R Units 3 and 4 Control Room Roof None None E, D Combustible Loading: E

- Fire Area MM Extinguishers: D Fire Area ID: MM (Unit 4) - Control Room Complex Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Reoulred Fire Protection Feature and Sstem Details (AN) Area Wide None None E]D ý uoete -- D ELET E 097 Unlis 3 and 4 Mechanical None ) D Detection Sf5 .t.OI Equipment Room - Fire Area MM f . xtingulstlers: D 106 Units 3 and 4 Control Room - Fire None 0Dlection D SYslem. Alarm Point 5":1 Ara ,MM Y

Deection System, Modification - M xtingullhers: 0 R

IOR Units 3 and 4 Control Room Roof None None E, D Combustible Loading: E

- Fire Area MM Extinguishers: 0 Fire Area ID: N (Unit 3) - Unit 4 Charging Pump Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description. System. System Protection Feature Rocluired Fire Protection Feature and System Details (Alt) Area Wide None None R ProcedureslGuidance: R 045 Unit 4 Charging Pump Room - E, D E. 0 None Detection System, Alarm Point 7: E D Fire Area N Water Suppression, 4-10-830: E D Fire Area ID: N (Unit 4) - Unit 4 Charging Pump Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fl!rn zonn ID Description SMtorm System Protection Feature Required Fire Protection Feature and System Details (Alt) Area Wide None None R Procedures/Guldance! R 0J45 Unit 4 Charging Pump Room - E E, D None Detection System. Alarm Point 7: E D Fire Area N Water Suppression, 4-10-830; E Fire Safety Analysis Data Manager (4. 1) Turkey Point Run: 06/26t2012 13:39 Page: 22 of 54

Attachment to L-2014-071 Page 16 of 101 RAI PRA O1.t.01.a In eliminating the use of panel factors, a new fire scenario development methodology discussed in the responses to PRA RAI 01.t in letter dated May 15, 2013 and PRA RAI 01.z.i in letter dated March 18, 2013 is implemented that differs from the ZOI-based approach originally used by the FPRA to support the LAR submittal. Provide a description of how this method, including its associated assumptions, impacts FPRA scenario development, including any effect on the previous quantification (these may be provided as part of the response to PRA RAI 29). In particular, address the following:

a. Describe and justify the method used for converting target damage times presented in Appendix H of NUREG/CR-6850 to percent damage as a function of heat flux, referred to here as the damage accrual function. The NRC staff has concerns with the use of this function because it makes assumptions about how damage occurs that do not appear to be connected to physical phenomena. Provide updated risk results removing credit for the damage accrual function. If any alternate treatment is used, provide similar description and justification.

RESPONSE

The methodology used to evaluate thermoplastic cable percent damage is based on the use of the time to damage data provided in Appendix H of NUREG/CR-6850 and applying an Arrhenius methodology, which is used extensively for environmental qualification (EQ) of components such as cables in a contai*nent accident environment, to determine the time to damage of the cables. An NRC internal evaluation of the Arrhenius methodology for equipment qualification is provided in a February 24, 2000 NRR Memo from Samuel J.

Collins to Ashok Thadani (ML003701987). The use of this data is therefore not considered an "unapproved method".

The NUREG/CR-6850 Appendix H, Table H-8 data provides times to target damage for thermoplastic cables for a set of steady state incident heat flux values. This provides a time delay for target damage beyond the damage heat flux of 6 kW/m^2.. For instance, Table H-8 provides a 19-minute time-to-damage delay for a thermoplastic cable with a steady state incident heat flux of 6 kW/m^2.

In order to apply the NUREG/CR-6850 Appendix H data to a fire with a tA2 growth rate, the EQ methodology of damage accrual is applied. The times to damage provided in NUREG/CR-6850 Appendix H were converted to damage rates by taking the reciprocal of the time to damage. For instance, the 19 minute time to damage for a 6 kW/mA2 incident heat flux in Table H-8 is converted to a 191 min-' damage rate. This provides a discrete set of damage rates for the heat flux values provided in Appendix H. An exponential regression is applied to these data points to generate a damage rate - heat flux profile. This regression analysis provides the Arrhenius curve for these cables based on the NUREG/CR 6850 Appendix H data.

The methodology used in the LAR-submitted Fire PRA model used a damage rate profile that assumed no damage before a critical incident heat flux was reached, directly applying the Appendix H data which states that no damage occurs prior to critical heat flux. The updated methodology that is proposed to update the model will assume a damage rate equal to the critical heat flux damage rate for, incident heat flux values up to and including the critical heat flux. This approach bounds any degradation of the cable target for heat flux values below the critical heat flux. Beyond the critical heat flux, the Arrhenius curve damage rates are applied

Attachment to L-2014-071 Page 17of 101 with no maximum damage rate applied, making this approach more conservative than that defined by the NUREG/CR-6850, Appendix H data. This ensures the use of a bounding damage rate curve without extrapolating. data to lower heat flux values, using the critical heat flux damage rate as a minimum damage rate, and not imposing damage rate limits beyond a maximum heat flux, thereby providing a conservative, bounding analysis. Figure 1 below shows a plot of the damage rate - heat flux profile that models this approach..

Damage Rate vs. Heat Flx

/

1

~' :p~:U

/

I.

/

V0 5

Heat FluŽ [kw~m2]

Figure 1. Damage rate - heat flux profile.

For an example calculation implementing this approach see the response to PRA RAI 01.t.0l.e.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with all 2 nd round RAI responses will be provided following NRC initial review, and feedback on this response.

RAI PRA O1.t.O1.b In eliminating the use of panel factors, a new fire scenario. development methodology discussed in the responses to PRA RAI O0.t in letter dated May 15, 2013 and PRA RAI 01.z.i in letter dated March 18, 2013 is implemented that differs from the ZOI-based approach originally used by the FPRA to support the LAR submittal. Provide a description of how this method, including its associated assumptions, impacts FPRA scenario development, including any effect on the previous quantification (these may be provided as part of the response to PRA RAI 29). In particular, address the following:

b. Describe and justify how the method accounts for pre-heating of targets that occurs at heat fluxes prior to reaching the peak heat flux for the fire being analyzed, including those below the target damage threshold. Note that in Section H. 1.5.2 of NUREG/CR-

Attachment to L-2014-071 Page 18 of 101 6850, the failure times reported in Table H-8 assume steady-state fire exposure conditions and are therefore not well suited for cases where exposure conditions evolve over time. Provide updated risk results that appropriately account for pre-heating or that conservatively do not credit the time delay associated with the pre-heating period.

RESPONSE

The methodology used in the LAR-submitted FPRA model used a damage rate profile that assumed no damage before a critical incident heat flux was reached, directly applying the Appendix H data which states that no damage occurs prior to critical heat flux. The updated methodology that is proposed will assume a damage rate equal to the critical heat flux damage rate for incident heat flux values up to and including the critical heat flux. This approach bounds any degradation of the cable target prior to the critical heat flux. Beyond the critical heat flux, the Arrhenius curve damage rates are applied with no maximum damage rate applied. This ensures the use of a bounding damage rate curve without extrapolating data to lower heat flux values, using the critical heat flux damage rate as a minimum damage rate, providing a conservative, bounding analysis. Figure 1 in the response to subpart a of this question shows a plot of the damage rate - heat flux profile that models this approach. As Figure 1 shows, for heat fluxes below the critical heat flux of 6 kW/m^2, the damage rate associated with 6 kW/mA2 is applied.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with all second round RAI responses will be provided following NRC initial review and feedback on this response.

RAI PRA 01.t.01.c In eliminating the use of panel factors, a new fire scenario development methodology discussed in the responses to PRA RAI 01.t in letter dated May 15, 2013 and PRA RAI 01.z.i in letter dated March 18, 2013 is implemented that differs from the ZOI-based approach originally used by the FPRA to support the LAR submittal. Provide a description of how this method, including its associated assumptions, impacts FPRA scenario development, including any effect on the previous quantification (these may be provided as part of the response to PRA RAI 29). In particular, address the following:

c. Describe and justify how the overall manual non-suppression probabilities developed for scenarios that employ this method are calculated to ensure any dependencies between fire scenario development branch points are appropriately taken into account.

Specifically, where multiple NSPs (e.g., those associated with time to target damage and time to HGL) are being applied to. fire scenarios associated with a particular ignition source, justify why these NSPs are considered to be independent. If not independent, provide updated risk results that appropriately accounts for the dependency. Also, confirm that, for any scenario, the resulting NSP applied is no less than 0.001, the NUREG/CR-6850-recommended lower bound.

RESPONSE

The Turkey Point Fire PRA model implements two types of scenario manual suppression factors for an ignition source: time to target damage and time to hot gas layer (HGL). The

Attaclment to L-2014-071 Page 19 of 101 time to target damage evaluates the direct heat flux incident on a target due to the fire. The time to HGL evaluates the volume temperature effects due to the fire.

The LAR-submitted model used the approach that these two analyses were independent of each other and therefore one was not conditioned on the other. The updated approach, which will be incorporated into the integrated quantification for the second-round RAIs, will consider the two analyses as dependent. Since the time to target damage in most cases is less than the time to HGL, the time to HGL will be conditioned on the time to target damage. For example, consider a time to target damage of 5 minutes and a time to HGL of 30 minutes, in the context of the event tree in Figure 1. The first node, event MS1, represents the time to target damage. Using the manual non-suppression (MS) distribution from NUREG/CR-6850, Supplement 1, Chapter 14 with a lanmbda value of 0.1.02 (electrical fires), the MS I probability is 0.602. The second node, event MS2, represents the time to HGL and is conditioned on the first node. In order to condition MS2 on MS 1, the time credited for. MS 1 is subtracted from the time available for MS2. In this example that would leave 25 minutes available for MS2, which, using NUREG/CR-6850, Supplement 1, Chapter 14, has an MS value of 0.079. Figure 1 shows the resulting fire scenario MS values for the three respective fire scenarios applying the node MS1 and MS2 values. The HGL fire scenario gets a 0.0468 MS value, which corresponds to a 30-minute non-suppression probability.

A detailed example of the overall calculation, including the manual suppression calculation, is provided in the response to RAI PRA 0l.t.0 .e.

MST Ir Sor

____________ __ M

_ __ _ jY IG=I

'* e4E5*z* , b&.**. ,

41-C

• 4..

Figure 1. Event tree with no detection or suppression reliability modeled.

The minimum manual non-suppression probability used is 0.001.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with all second round RAI responses will be provided following NRC initial review and feedback on

• this response.

Attachment to L-2014-071 Page 20 of 101 RAI PRA 01.t.01.d In eliminating the use of panel factors, a new fire scenario development methodology discussed in the responses to PRA RAI 01.t in letter dated May 15, 2013 and PRA RAI 01.z.i in letter dated March 18, 2013 is implemented that differs from the ZOI-based approach originally used by the FPRA to support the LAR submittal. Provide a description of how this method, including its associated assumptions, impacts FPRA scenario development, including any effect on the previous quantification (these may be provided as part of the response to PRA RAI 29). In particular, address the following:

d. Describe and justify how automatic detection and suppression systems, including treatment of system reliability and unavailability, are credited by scenarios that employ this method. If not included in the analysis, provide updated risk results that appropriately account for both reliability and unavailability where these systems are credited.

RESPONSE

Reliability and unavailability of automatic detection systems were assumed in the LAR-submitted model to be incorporated in the manual non-suppression probabilities specified in NUREG/CR-6850, Appendix P, as revised in NUREG/CR-6850, Supplement 1. Reliability of automatic suppression systems was based on values specified in NUREG/CR-6850, while availability was not considered to impact the reliability data given that plant procedures specify compensatory actions to be implemented when the systems are not available. In order to address the concern that the infonnation inherent in the NUREG/CR-6850 data may not be bounding, the model will be updated to incorporate this additional failure potential. The scenario development event tree will incorporate an additional node, before any suppression (manual or automatic) is credited. The event tree detection failure path will include a 15-minute time delay before manual suppression is allowed to be credited (using SDP guidance for detection time for locations without detection systems). Figure 1 shows an event tree without consideration of detection failure. Figure 2 shows the updated approach which incorporates detection failure. Note that the MS1/MS 15 and MS2/MS_15 values will be bounded to a maximum value of 1. This will result in zero ignition frequency being applied to the success branch for instances where the time to target damage or time to hot gas layer is less than 15 minutes.

NUREG/CR-6850 Appendix P suggests a bounding failure probability for smoke detection based on the halon suppression failure probability. The data used to develop the halon suppression failure probability included detection failure (smoke detection), so the detection failure probability by itself is bounded by the halon failure probability. Since no guidance on thermal detection failure probabilities is given in NUREG/CR-6850 Appendix P, the same approach, use of the associated suppression system failure probability, will be applied to the corresponding detection system. The NUREG/CR-6850 failure probability specified in Appendix P for deluge or pre-action sprinkler systems is conservatively applied as the failure probability for thermal detectors associated with actuation of a pre-action system.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 "dround RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial. review and feedback on this response.

Attachment to L-2014-071 Page 21 of 101

e. IGF FrASovan Iankon imqreiecy: or to tanet danr-aae ~ layer~bo IA!S I

^;4 I Now-Sv~e' 1.4 44).S2 I Seveie 1AS2 4SO 2 -32 Figure 1. Event tree with no detection or suppression reliability modeled.

Figure 2. Event tree with detection reliability modeled.

RAI PRA 01.t.01.e In eliminating the use of panel factors, a new fire scenario development methodology discussed in the responses to PRA RAI 01.t in letter dated May 15, 2013 and PRA RAI 01.z.i in letter dated March 18, 2013 is implemented that differs from the ZOI-based approach

-originally used by the FPRA to support the LAR submittal. Provide a description of how this method, including its associated assumptions, impacts FPRA scenario development, including any effect on the previous quantification (these may be provided as pdrt of the response to PRA RAI 29). In particular, address the fo:llowing: '

e. For the response, provide a complete example of how the method was applied on a specific fire ignition source, addressing Parts (a) through (d).

RESPONSE

Attached is a detailed calculation for a NLUREG/CR-6850 Appendix E, Table E-1, Case 3 fire, including the damage accrual for target damage (part a and b), the time-to-hot-gas-layer evaluation (part c), and the manual suppression calculation (part c). The response to part d of the question includes an example of treatment of detection/suppression systems. In order to keep the example calculation as straightforward as possible, only the Table E-4, Bin 3 fire is evaluated.

Attachment to L-2014-071 Page 22 of 101 PTN RAI PRA O1.t01.e" Example Case 3. Bi.n 3 Calculation PTN RAI PRA 01101I.e: Example Case 3 Bin 3 calculation This calculation performs a simplified eoaluation for the time to damage of a target using the damage accrual methodology and the time to hot gas layer (HGL). This example will evaluate a NUREGCR.-6850 Appendix E Case 3, single cable bundle thermoplastic, bin 3 ignition source fire impacting a taroet at a vertical distance of 3 feet, in a 6000 cubic foot volume. The calculation is organized into 6 sections: input data., heat flux profile damage rate profile, correlation of the heat flux and damage rate profiles, time to HGL, and manual non-suppression probability.

Input Parameters This section provies a tabulation of the input data to be used.

" HRR - thie HRR distribution row for a Case 3 bin 3 fire from NUREGJCR-6850 Appendix E

" HF_distance - The Generic Fire Modeling Treatments Table 5-11 provides vertcal damage distances for an Appendix E Case 3 bin 3 fire impacting thermoplastic (5.7 kWim^2), class A combustible (9 kWVm^2), and thermoset (11 kWMm'2) targets. These three heat flux - distance points are used for the first three columns: The Generic Fire modeling Treatments Table 5-4 pro-,4des the flame height distance for the an Appendix E Case 3 bin 3 fire. The half flame height is assumed to be the point at which the maximum heat flux: of 120 kWlmi2 is incident. The max heat flux - distance point for a medium sized cabinet is used as the fourth column in the matrix.

" distance - the verticat distance from the fire to the target

" qdata - tabulates the four heat flux points 5.7. 9t 11, and 120 kWm'm-2

" damage data - tabulates the damage time from NUREGICR-6850 Appendix H Table. H-8

" tHGL_ provides times to an 80 degree Celsius hot gas layer temperature for 3530 cubic foot and 8820 cubic foot volumes, for fires corresponding to 69 kW, 211 kW, and 464 kW. HRR sizes.

" Volume - the volume of the fire zone within which the scenario is postulated. For this example the fire zone has a volume of 6000 cu. ft.

(Tvaw&Boumd" "Upper Boimdý "Point Vake~' "SF"'

_3 79 65 0.192) d"5.7 kW.';*' . kW~m'2 "l I kW.m; 2" 120 kWlm*2' Iff-distce = 6 4"3 5.381 4.905 1.1693

-- qkt  !=

~6 19> 5.7 69 810 9 HGL BMRý27 ID0 6 11.4 t464, d~mg_*a=11 damageý data :t= 4 i 120) 20 14 2 vi. 16 .1 Page 1 of 10

Attachment to L-2014-071 Page 23 of 101 PTN RAJ RA Ol.tOl.e- Example Case 3 Bin 3 Calculation

"'530 cu. -" '-8N20ci. " Volume:= 6000 di-tmnce

.ft. 3 12.3 60 tH-C-LISG: *7K- 7.8 10.7 1

ý,"464 kW' 5.7 7.8 Heat Flux Profile This section produces the heat flux - time profile by evaluating the peak incident heat flux at a distance of 3 feet and then using a tA2 growth from NUREGICR6BSO Appendix G, replacing peak HRR with peak heat flux.

f6.433 Extract from input data the relevant distance data.

r (HF dL-tare

.1-169 Perform a power regression

.of the heat flux and distance data to generate a heat flux -

distance function coefficients A, B and C, for Case 3 tan 3.

Generate a generic power function that accepts A, B, and C coefficients, and a 120 qf,,x(dtABz -fF~dq 1 .C *t distance d.

Apply the power regression coefficients to the generic power 12 iff.dqiO~ftrft ):120 function, using a mad.mum heat flux of 120 kWArr02. This generates a function which evaluates the incident heat flux given Fý'4 qfj , q fijqt) otews a vertical distance, d, from the fire- The graph belomw shows a plot of q.(d) *vs d.

Page 2 af 10

Attachment to L-2014-071 Page 24 of 101 PTNFRAI PRA 01 .tO I.e: Example Case 3 Bin 3 Calculation Heat Flux vs. Vertical Distance. for Select HRR Bins

. . .. I

%,(d)

Sc \

a 4 8 d

Distance [ft]

qmnaxz o.7 q(3) = 2&6137 Madmum incident heat flux to a target at 3 feet by a Case 3 bin 3 fire.

To determine the heat flux as a function of time it was assumed that the heat flux wRlI increase proportionally to the HRR increase of the ilnition source. Therefore, to determine the heat flux as a function of time, the equation from NUREGICR-6850 Appendix G was used. The peak HRR was replaced wvth the pesak heat flux, qmax, as determined by the function qa(d).

q;(r) = .{410 4.ýLj10..q__max.,.(-, if K 1 mah.II0 ,mmt1_O._maz..I othem'se Pag. 3 of Ita

Attachment to L-2014-071 Page 25 of 101 PTK RA* PRA Ol.tOl.e: Example Case 3 Bin 3 Calculation The graph below plots the incident heat flux vs. time for a target 3 feet vertical feet from a Case 3 bin 3 fire.

Incident Heat Flux vs- Time at a Distance of 3 feet 40 0 . .".

0 /1 Co .,_ _ __ _ ___________/____

0 6 15 34-0 Timne t [minutes]

Pa?. 4 of 10

Attachment to L-2014-071 Page 26 of 101 PTN RA* PRA 01.01.e: Example Case 3 Bin 3 Calculation Damage Rate Profile This section produces the damage rate vs heat flux profile (damage rate profile) based on the NUREGICR-6850 Appendix H damage data. This damage rate profile vill be correlated to the incident heat flux profile to produce a damage rate vs. time profile for the scenano.

Pull relevant data from the Appendix H 6 input data tabulated above.

dm-e :=d~amzg*dta ý) 1 damape-time:= damage-dt q) 6 4

14

,16) 0.053' Convert the damage times to a damage rate by evaluating the reciprocal 0.1. of the d*amage time for each heat flux.

dhamage_rate 1 ,!6 dam e tme 0.25 0.5

ýC:= ez;f &it~~maaE

[

tid=Mae~rate,.1 1 I 4.702 x10 0.3.32 3

Perform an exponential regression of the heat flux and damage rate data to generate a heat flux -

" 0.034 damnage rate fut:rction coefficients A, B and C, for the Case 3 bin 3.

Generate an exponential function DR cu C 0 .ezp~ + C, that uses the A, B, and C coefficients, and a heat flux q.

This applies boundary conditions on the heat flux -

D-t(c:= y if q 5.7 v D 1.

c e(q) damage rate function. The first line ensures a 1119 minimum damage rate for a heat flux (q) less than 5.7 19 OR for damage rates less than 1/19. The second line I~k I,,,(cj ifq > ,-7A. D- -(q 1 applies the damage rate curve, -K.O (q), if the heat flux is greater than 5.7 AND the damage rate is creater than 1/19.

Psae 5 f 10

Attachment to L-2014-071 Page 27 of 101 ITN RAJ PR,&.DI.LO..e: Example Case 3 Bin 3 CalcujlaionI The plot below shows the damage rate vs. heat flux function produced from the exponential regression and botmdary conditioning aboe. As the plot show,, a damage rate equal to the critical damage heat flux damage rate of 1119 per minute is applied to heat fluxes up to and including the damage heat flux of 5.7 kWlm'.2 No maaximum damage rate boundary is applied.

Damage Rate vs. Heat Flux

°--

D (q)

-a 0 15 20 q

Heat Flim [kinvmn2]

Page 8 of 10

Attachment to L-2014-071 Page 28 of 101 PTN RA* PRA 01.t0l.e: Example Case 3. En 3 Calculation Correlation of Heat Flux and Damage Rate Profiles The plot below shows the damage rate curve over time applicable to a Case 3 bin 3 fire impacting a thermoplastic target at 3 feet vertical distance. In order to capture the cumulative damage overtime, this function n to be integrated.

Dama~ge Rate vs- Time at a Distance of 3 feet D3Jpq3ý.t)I 0.s

~rJ 0 5 10 15 Time, t [minutes]

The function below integrates the danmage rate vs. time function using a time step of 0.1. The integration terminates vven the accrued damage equals t1-. The time step at which this occurs is the time to target damage.

tdrnmg 0 whie DmL- < I

'd- td-. -,e Dmg Din5 4 Dk`3tml tdar ý-- tdmn tdam = S.6 Page 7 of 10

Attachment to L-2014-071 Page 29 of 101 PTN RAI PRA 01.tO1.e: Example Case 3 Bin 3 Calculation Time to Hot Gas Layer The approach taken to calculate the time to hot gas layer (HGL) is to perform a linear interpolation between the data for the two volumes that bound the scenario fire zone volume, for each HRR value. For this example, the scenario fire zone volume is 6000 cu. ft-, which is bounded ty volume data from 'Evatuation of the Development and Timing of Hot Gas Layer Conditions in Generic PTN Fire Compartments with Secondary Combustibles Rev 1' (HGL Secondary Combustible Report) for volumes of 3530 cu- ft.

and 8820 cu. ft. Once a scenario fire zone volume specific time to HGL is calculated using linear regression, a power regression is performed on the three HRR - time to HGL points in order to generate a time to HGL values for the Case 3 bin 3 HRR (65 kW).

Define a linear fegression function to interpolate between CHGL(VL V'H,H&LLHGLH) - ir 1%' )I HGLH, volume data points. This function produces the A and B coefficients for the linear function A+xB.

Define a linear function based on the linear regression function tHGL(`Vol, AB): Vol-B A-i- with a volume input parameter and a time to HGL out parameter Define power fit function which accepts power regression coefficients HiGLtnme(H, A IBI Q A-IC A, B, and C.

19ST&b:= 65 Define the applicable HRR for use in the time to HGL calculation.

Volhmei -. 3530 Volumehizh:= S820 Low and high volume variable definitions used for the linear regression.

Time to 80 degree Celsius

• 23" 60 60 degree Celsius HGL temperature vector columns

%= S.Jg 1 0 -7! corresponding to the low and high vohrmes for 69kW, 237 IVW, and 464 kW.

69 kW Linear Regression CHiGL69 =CHGL(Volumeio0 . Voluehj, r-G~t:1GL~ Generate linear regression coefficients for 69 kW using generic function defined above.

t8069"= tHGI4 Vollue,CCHGL69 0., CHGL691 Using linear regression coefficients and the fire scenario fire zone volume, generate a fire zone specific time to HGL 237 kW Linear Regression CHGL237:= CiiGL.{Voiumelow.VotmneLhjtHGLjoW .tHGLjL" Generate linear reg:ression coefficients for 237 kW using generic function defined above-t0 ,_.

2  := tHGVolume,{C-IGL237., CHGL2371) Using linear regression coefficients and the fire scenario fire zone volume, generate a fire zone specific time to HGL 464 kW Linear Regression Generate linear regression coefficients for 464 CiGL464 := CHGL(Vo,*we- ohu.eL -, tHGL dItGLh,- VIW using generic function defined above.

t804,4 := .HG4(VolumeCHGL464D, CHGL464 1) Using linear regression coefficients and the fire scenario fire zone volume, generate a fire zone specific time to HGL Pap_ 8 of 10

Attachment to L-2014-071 Page 30 of 101 FTt RA] PR PRA 01 le: Example Case 3 Bin 3 Calculation Times to HGL for the specific PAU volume for 8OC L

8t6P = -345721 Interpolated times to HGL pmduced from linear regression function.

t:,GLo c tBO, 9.154) tsyj464., 6.6S1 f {G2..L2 Generate power regression function relating HRR ,.s time

-f= t R.t-GL.SO Ha*FG [-

to HGL for the specific volume ard temperature threshold-The (32, -1, 1) are guesses for each parameter of the power function-tHGL_SCenaxio := HGLime" HER \0 bin., CS 0 .C 0 * .C 0 1 = 37.41 Evaluate time to HGL for the scenario.

Page r of 10

Attachment to L-2014-071 Page 31 of 101 PTN RPA .PRA 01101.e: Example Case 3 Bin 3 Calculation Determination of NSP value for each end state value in scenario decision tree This section inputs the times to target damage and HGL into the manual non-suppression distribution from FAO O8-0050. The lambda value used is 0-126 for transient fires and 0 102 for electrical fires per guidance from the FAQ. Once the manual non-suppression probabilities (MSs) are calculated, tney are normalized based on the event tree provided below. This accounts for the fact that a fraction of the time to suppress the fire before a HGL is reached has been used up by the time to target damage_ Essentially, each time to suppress is calculated independently, when in fact they are part of the same timeline, andl thus when aggregating the timelines, they must be normalized or conditioned by the previous suppression times.

l~blrIoNFR5QjENcY <i~sr~' .~ tt~*2 ~e~ceraNee Fire ser stn Fttstimn s armrsc I ft e-ý LVleve I SF

- IMM.4ýn'vEM.wfl A 145I Rot 3 e' fi eer n5 c.:

SF := 0.192 Case 3 distribution probability for bin 3 fire.

(I- 0a 'lda,,j NIS1 0.001 if e :0.001 = 0.416 Calculate the manual non-suppression probability for the time to target damage, MS1, using a minimum boundary condition of t(- O0V2"Kam) ie' oherme 0.001.

MS2 =0.001 if e(- 0.1*.Co- _ O) :0.001 = 0.022 Calculate the manual non-suppression probability for the time to Ie(- 0.10O2.t:-:L_tce-an)or-e HGL, MS2, using a minimum boundary condition of 0.ODI.

-"S2*°o := MSi if MS1 :. 0 = 0.052 Determine the conditioned MS2 by conditioning MS2 on. MSi.

MSI IMS2 Othr-a-e NSP Evaluation This section evaluates the end state non-suppression probaNlities (NSPs) in the event tree above, using the MS values evaluated above.

NS?-A i=1 - MSI).SF =0.112 NSPA applies to the non-severe fire TSPsB. - MS2,

.is(I ~SF = 0.076 NSPB applies to the 98th percentile NSPC applies to the HGL scenario NSP.C:- MS]-MS2,,,dSF 4.119 x 107 Page 10 of 10

Attachment to L-2014-071 Page 32 of 101 RAI PRA O1.v.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 01.v and stated that non-suppression credit for automatic suppression systems was only credited in PAUs where automatic suppression systems are actuated by smoke detection. However, the responses to PRA RAI 01 .bb from letter dated May 15, 2013 and PRA RAI 01 .cc from letter dated March 18, 2013 state that modes of actuation include thermal detectors and fusible links. Clarify this discrepancy, and summarize how the fire suppression system activation times for each PAU with a non-smoke-detector-actuated system are estimated and credited.

RESPONSE

Automatic suppression was credited in the Fire PRA in eight Fire Zones identified in Table 1 below. There are three fire scenario configurations for which credit was taken for automatic suppression - zone of influence (ZOI), 80 'C hot gas layer (HGL), and 205 'C HGL.

For instances where automatic suppression is credited with mitigating impacts on the ignition source ZOI using halon suppression(Fire Zones 98, 108A, 108B), the detection actuation system is ionization smoke detection. Fire zone 79A credits the installed deluge suppression system for mitigation of transient fire ZOI impacts using a thermal detection system, requiring a plant modification (3-CC-02/4-CC-02) to install thermal detecting wire within cable tray risers which could be ignited by transient combustibles and could propagate fire to trays of concern in the upper elevations of the zone. No timing evaluation was performed for this thermal detection credit since the detection wire will be physically installed within the trays which must ignite in order to cause propagation of fire to the fire trays of concern in the upper elevations of the fire zone.

For 80'C HGL scenarios, the halon suppression systems were credited as part of the event tree which incorporates target damage and HGL impacts. See response for RAI PRA 01 .t.01 for further discussion of the scenario event tree. In this instance, the halon is credited in the ZOI evaluation as well as the HGL evaluations.

For 205'C HGL scenarios, the halon suppression systems and the pre-action sprinklers in fire zones 45, 55, 74, 75, 98, 108A and 108B were credited. No timing evaluation was performed for the thermal detection actuation since the detection system setpoint 93.3 'C (200 F) is much lower than the 205'C HGL and is assumed to always actuate before the HGL temperature is reached unless the detection system fails. The failure of the detection system is incorporated into the analysis as discussed in the response to RAI PRA 01 .t.01 .d.

A markup to LAR Attachment C, Table C-2, NFPA 805 Required Fire Protection Systems and Features, is attached to this response to reconcile suppression systems credited by PRA (for Risk, "R"). This markup is based on a revised Attachment C, Table C-2, provided in the response to FPE RAI 14, via FPL transmittal letter L-2014-003, dated 01-07-2014.

Attachment to L-2014-071 Page 33 of 101 TABLE 1. CREDITED AUTOMATIC SUPPRESSION SYSTEMS Credit Taken Fire Zone Suppression Type Detection Type ZOI 80 OC HGL 205 °C HGL 45 - U4 Charging Pump Deluge Thermal Room X 55 - U3 Charging Pump Deluge Thermal Room X 74 - U3 Train B DG Day Deluge Thermal Tank Room X 75 - U3 Train A DG Day Deluge Thermal X Tank Room 79A - U3 and 4 Aux Deluge Thermal Bldg North-South X Breezeway 98 - UJ3 and 4 Cable Halon Ionization X X x Spreading Room 108A U3 and 4 Train A Halon Ionization ' ' "

DC Equipment Room 108B U3 and 4 Train B Halon Ionization X X x DC Equipment Room

Attachment to L-2014-071 Page 34 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: C (Unit 4) - Unit 4 RHR Compliance Basis: NFPA 805, Section 4.2.4. Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description system System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide None None R Procedures/Recovery Actions: R 014 Unit 4 RHR Heat Exchanget None R, D, A E Detection System, Alarm Point22: R.D A Room - Fire Area C Physical separation: E 015 Unit 4 RHR Pump A Room - Fire None R. D, A E Combustible Loading: F Area C Detection System, Alarm Point 22: R D A 016 Unit 4 RHR Pump B Room - Fire None R. D, A None Detection System, Alarm Point 22: R D A Area C Fire Area ID: CC (Unit 3) - Units 3 and 4 Auxiilary BuIlding North-South Breezeway Compliance Basis: NFPA 805, Section 4.2.4. Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide Non/e None R, D ProcedutesiRecovery Actions: R D K

079A Units 3 and 4 Auxiliary Building North-South Breezeway - Fire Area CC E, D) 7 R Combustible Centiol - Transient Restrictions: R Detection System. Alarm Point 39: R D Detection System, Modification - CC U3: R ERFBS, Modification CC U3: R FireBarrier, MODIFICATION - CC U3T R Water Suppression, 10-850: E Fire Area ID: CC (Unit 4) - Units 3 and 4 Auxiliary Building North-Sbulh Breezeway Compilailce Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protect*ro Feature Required Fire Protection Feature and System Details (All) Area Wide None None ) ProcedureslRecovery Actionrs: D 079A Units 3 and 4 Auxiliary Building North-South Breezeway - Fire E.015 R. D R Combustible Control - Transient Restrictions: R Detection System. Alarm Point 39: R D Area CC Detection System, Modification -CC U4: R ERFBS, Modification CC U4: R FireBarriet, MODIFICATION -CC :R Water Suppression, 10-860: IED VJ Fire Safety Analysis Data Manager (4.2) FPL_ - Turkey Rum 12J1912013 16:34 Page; 8 of 68

Attachment to L-2014-071 Page 35 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: GG (Unit 4) - Unit 3 480V Load Centers C and ) Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Firo Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide None None R Procedures/Recovery Actions. R 09B Unit 3 480V Load Centers C and None R. D, A None Detection System. Alarm Point 1: R D A DRoom - Fire Area GG Fire Area ID: H (Unit 3) - Unit 3 West Electrical Penetration Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 019 Unit 3 West Electrical Penetration None R, D R, S, N Detection System, Alarm Point 9: R D Room - Fire Area H ERFBS. 019-1: R S N ERFBS, Q19-2; R SN ERFBSq 09-t3: R S N ERFBS, 019-4: R S N Fire Area ID: -H (Unit 4) - Unit 3 West Electrical Penetration Room Compliance Basis: NFPA 805. Section 4.2.3.2, Separate Fire Area Required Required-Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 019 Unit 3 West Electrical Penetration None None None None Room - Fire Area H Fire Area ID: HH (Urit 3) - Units 3 and 4 Cable Spreading Room and Chase Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone t0 Description System System Protection Feature Required Fire Protection Feature and System Details (Ail) Area Wide N None R, D ProcoduresYRecovery Actions: R D 098 Unils 3 and 4 Cable Spreading R, D, A R Gombustirbe Control -Transient Restrictions: R Room - Fire Area HH Detection System, Alarm Point 6: R D A Detection System, Modification - 1I1 U3: R ERFBS, Modification - HH U3: R Gaseous Suppression, CV- 1450A&B CV- 145 IA&B: L@ ý 132 Units 3 and 4 Control Room D. A R Combustible Control - Transient Restrictions: R Electrical Cable Chase - Fire Detection System, Alarm Point 5: D A Area HH Gaseous Suppression. CV-1450A&B CV-1451A&B: Dt FPL - Turkeft Run: 12/19,2013 15:34 Page: 17 of 58 Fife Safety Ana"Is Data Manager (42)

Attachment to L-2014-071 Page 36 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features HH (Unit 4) - Units 3 and 4 Cable Spreading Room and Chase Fire Area ID:

Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone tD Description System System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide No~p* None R, D Procedures/Recovery Actions: R D 098 Units 3 and 4 Cable Spreading D R, D, A R Combustible Control - Transient Restrictions: R Room - Fire Area IH Detection System, Alarm Point 6: R D A Detection System, Modification - HH U4: R ERFBS. Modification - HH U4: R Gaseous Suppression, CV-1450A&B CV-1451A&B:

132 Units 3 and 4 Control Room D D. A R Combustible Control- Transient Restrictions: R Electrical Cable Chase - Fire Detection System. Alarm Point 5: 0 A Area ViH Gaseous Suppression. CV-1450A&B CV-1451A4&B: D Fire Area ID: I (Unil 3) - Unit 3 South Electrical Penetration Room Compliance Basis: NFPA 805, Section 4.2-4, Performance Elased Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Reguired Fire Protection Feature and System Details 020 Unit 3 South Electrical None R, D, A None Detection System, Alarm Point 9: R D A Penetration Room - Fire Area I Fire Area ID: I (Unit 4) - Unit 3 South Electrical Penetration Room Compliance Basis: NFPA 805, Section 4.2.3.2, Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 020 Unil 3 Soulh Electrical None A None Detection System, Alarm Point 9: A Penetration Room - Fire Area I Fire Safety Analysis Data Manager (42) FPL - Turkey Run*: 12/19/2013 15:Z4 Page: 18 of 58

Attachment to L-2014-071 Page 37 of 101 Attachment C Table C-2 NFPA 806 Required Fire Protection Systems and Features Fire Area ID: NN (Unil 3) - Urnis 3 and 4 A DC Equipment Room Compliance Basis: NFPA 805. Section 4,4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zonu ID Description System System Protection Feature Required Fite Protection Feature and System Details (All) Area Wide None None R Procedures/Recovery Actions: R 108A Unils 3 and 4 A DC Equipment R, S, R,D. S0 N, A R, S. N Delecion System, Alarm Point 15: R D S N A Room - Fire Area NN ERFBS, 108A-1: R S N ERFBS, 110A-2: R S N ERFBS, 108A-3: R S N ERFBS. 108A-4: RS N Gaseous Suppression, CV-1452 CV-1453: R S N Fire Area ID: NN (Unil 4) - Units 3 ind 4 A DC Equlpmenl Room Compliance Basis: NFPA 805, Section 4.2.4. Performance Based Approach Required Required Suppression Detection Required Fire Fire Zonei ID Description System System Protection Feature Reqtired Fire Protecli.- Fcaturp and Sysiewm Detaiis (All) Area Wide None None R Procedures/Recovery Actions: R 108A Units 3 end 4 A DC Equipmenl R, S, R, D, S, N, A R, S, N Detection System, Alarm Point IS: R D S N A Room - Fire Area NN ERFBS, 108A-1: R S N ERFBS, 10BA-2: R S N ERFBS, 10BA-3: R S N ERFBS, 108A-4: R S N Gaseous Suppression, CV-1452 CV-1453: R S N Fire Area ID: 0 (Unit 3) - Unit 3 and 4 Boric Acid Tanks and Pumps Rooms Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description system System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide None None R ProcedureslRecovery Actions: R 041 Units 3 and 4 Boric Acid Tanks Norne None E Physical separation: E and Pump Room.- Fire Area 0 055 Unit 3 Charging Pump Room - E, A E, D,A None Deteclion System, Alarm Point 7: E DY Fire a 0 . 7 Water Suppression, 3-10-841: E.A /

Fire Sateky Analysis Data Manager (4.2) FP'L - Turkey Run: 1211912013 15:34 Page: 24 of 58

Attachment to L-2014-071 Page 38 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: 00 (Unit 3) - Units 3 and 4 B DC Equipment Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Descrption System System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide Noney* None R Procedures lRecovery Actions: R 108B Units 3 and 4 DC Equipment S. N 6.ý. R, D, S, N, A R. S, N Detection System, Alarm Point 15: R D S N A Room -Fire Area 00 ERFBS, 108B-1: R S N ERFBS, 1081-2: R S N Gaseous Suppression, CV-1454 CV-1455: S Fire Area ID: 00 (Unit 4) - Units 3 and 4 8 DC Equipment Room Compliance Basis: NFPA 805, Section 4,2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone lb Description System System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide' None None R ProceduresIRocc'ery Actions: R 10Ba Units 3 and 4 DC Equipment D, S, Na R. D, S, N, A Rk 5, N - Detection System, Alarrm Point 15: R D S N A Room - Fire Area 00 ERFBS, 0813- 1: R ERFBS, 108B-2: RSSNN Gaseous Suppression, CV-1454 CV-1455: 0 S Fire Area ID: P (Unit 3) - Unit 4 Containment Building Compliance Basis: NFPA 805, Section 4,2.3.2, Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 059 Unit 4 Containment Building - Fire None A E Combustible Loading: E Area P Detection System, Alarm Point 43: A Detection System, Alarm Point 44: A Fife Safety Analysis Data Manager (4.2) FPI.,- Tu~rkey Run: 12119/2013 15:34 Page: 45 of 58

Attachment to L-2014-071 Page 39 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area 11: RR (Unit 4) - Unit 4 Train A Emergency Diesel Generator Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 138 Unit 4 Train A Emergency Diesel E, R, D R, D None Detection System, All available - RR: R D Generator Room - Fire Area RR Water Suppression, 4-10-1 112: E R D Fire Area ID: S (Unit 3) - Units 3 and 4 Computer Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 052 Unils 3 and 4 Computer Room - None Rk D, A None Detection System. Alarm Point 31: R D A Fire Area S Firm Are 1: S (Unit 41)- Unitts 3 and 4 Computer Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fite Zone ID DescriPtiorI System System Protection Feature Required Fire Protection Feature and System Details 062 Units 3 and 4 Computer Room - None R. D, A None Detection System. Alarm Point 31: R D A Fire Area S Fire Area ID: SS (Unit 3) - Unit 4 Train B Emergency Diesel Generator Room Compliance Basis: NFPA 805. Section 4.2.3.2, Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 133 Unit 4 Train B Emergency Diesel FO A None Detection System, All available - S*: A Generator Room - Fire Area SS Water Suppression, 4-10-1113: EM Fire Salety Analysis Data Manager (4,2) FPL - Turksy Run: 12119/2013 15:34 Page: 49 of 58

Attachment to L-2014-071 Page 40 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features ___________________

Fire Area ID: SS (Unit 4) - Unit 4, Train B Emergency Diesel Generato*r Room Compliance Basis: NFPA 805, Section 4.2,4. Performance Based Approach Required Required Suppression Detection Required Fire Fire Zono ID Description System System Protection Feature Required Fire Protection Feature and System Details 133 Unit 4 Train B Emergency Diesel E/ jO* R. D. A None Detection System, All available - S,,1,R D A Generator Room - Fire Area SS > Water Suppression, 4-10-11 13:

Fire Area ID: T (Unit 3) - Unit 3 Reactor Control Rod Equipment Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and S.stem Delails (All) Area Wide None None 0 Procedures!Recoveny Actions: D 063 Unit 3 Reactor Control Rod None D0,A None Detection System, Alarm Point 6: D A Equipment Room - Fire Area T Fire Area ID: T (Unit 4) - Unit 3 Reactor Control Rod Equipment Room Compliance Basis: NFPA 805, Section 4,2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 063 Unit 3 Reactor Control Rod None A None Detection System. Alarm Point 6: A Equipment Room - Fire Area T Fire Area ID: TT (Unit 3) - Train A SD - Unit 3 Swdchgear Room 3D Complearnce Basis: NFPA 805, Section 4.2.3.2, Separate F're Area Required Required Suppression Detection Required Fire Fire Zonc ID Description System System Protection Feature Required Fire Protection Feature and System Details 134 Unit 3 Switchgear Room 3D - Fire None A E Combustibie Loading: E Area TT Detection System, Detection 134: A Fire Safety Analysis Data Manager (4.2) FPL - Turke.y Run- 12119/2013 15:34 Page: 50 of 58

Attachment to L-2014-071 Page 41 of 101 Attachment C Table C-2 NFPA 806 Required Fire Protection Systems and Features Fire Area 10: WtN (Unit 3) - Unit 4 Train A Diesel Fuel Oil Handling Areas Compliance Basis: NFPA 805. Section 4.2.3,2, Separate Fire Area Required Required Suppression Detection Required Fire Firo Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 141 Unit 4 Train A Diesel Oil Transfer A None Detection System. Detection 141: A Pump Room - Fire Area ¶W E& Water Suppression. 4-10-1122 (4A): E&

142 Unit 4 Train A Diesel Oil Storage None None None None Tank - Fire Area WWV Fire Area iD: WW (Unit 4) - Unit 4 Train A Diesel Fuel Oil Handling Areas Compliance Basis: NFPA 805, Section 4.23.2, Separale Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 141 Unit 4 Train A Diesel Oil Transfer A None Detection System. Delection 141: A /'7 Pump Room. Fire Area WW water S ...... 4-^1,0-122 (4A);..g 142 Unit 4 Train A Diesel Oil Storage None None None None Tank - Fire Area WVN Fire Area ID: X (Unit 3) - 4160V Switchgear 3A Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details (All) Area Wide None None R, O Piocedures/Recovery Actions: R D 071 4160V Swltehgear 3A Room - None D. A R, S. N Detection System, Alarm Point 1: D A Fire Area X Detection System, Modification - XU3: D ERFBS, 071-1: RSN ERFBS, 071-2: R S N Fire Safety Analysis Data Manager (4.2) FPL - Turkey Run: 12119/2013 15:34 Page: 55 of 58

Attachment to L-2014-071 Page 42 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: X (Unit 4) - 4160V SvAthgear 3A Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protectlin Feature Required Fire Protection Feature and System Details (All) Area Wide None None R Procedures/Recovery Actions: R 071 4160V Switchgear 3A Room - None R, 0. A None Detection System, Alarm Point 1: R D A Fire Area X Detection System, Modilicatton -X U4: 0 Fire Area ID: XX (Unit 3) - Unit 4 Train B Diesel Fuel Oil Handling Areas Compliance Basis" NFPA 805, Section 4.2.3.2, Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System -System Protection Feature Required Fire Protection Feature and System Details 136 Unit 4 Train B Diesel Oil Transfer -A None Detection System, Detection 136: A Pump Room - Fire Area XX 10 . Water Suppression, 4-10-1122 (48):

137 Unit 4 Train B Diesel Oil Storage None None None None Tank - Fire Area XX Fire Area ID: XX (Unit 4) - Unit 4 Train B Diesel Fuel Oil Handling Areas Compliance Basis: NFPA 805, Section 4.2.3.2, Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 136 Unit 4 Train B Diesel Oil Transfer A None Delection Syslem, Detection 136: A Pump Room - Fire Area XX Water Suppression, 4-10-1 122 (4B):FE 137 Unit 4 Train B Diesel Oil Storage None None None None Tank - Fire Area XX Fire Area ID: Y (Unit 3) - Unit 3 Train B Emergency Diesel Generator Building Compliance Basis: NFPA 806, Section 4.2.4. Performance Based Approach Required Required Suppression Detection Required Fire System System Protectiun Feature Reaulrod Fire Protection Feature and System Details Fire Zone ID Descr1otion 072 Unit 3 Train B Emergency Diesel ') A None Detection System, Alarm Point I r-sD A Generator Building - Fire Area Y Water Suppression, 3-10-844: DV5 Fire Safety Analysis Data Manager (4.2ý FPL - Turfkey Run: 12119t2013 15:34 Page: 55 of 58

Attachment to L-2014-071 Page 43 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area I1: Y (Unit 4) - Unit 3 Train B Emergency Diesel Generator Building Compliance Basis: NFPA 805, Section 4.2.3.2. Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 072 Unit 3 Train 8 Emergency Diesel $ 1 - A None Detection System, Alarm Point 16 A Generator Building - Fire Area Y Fire Area ID: YY (Unit 3) - Unit 4 Train B Emergency Diesel Generator Control Room Compliance Basis: NFPA 805, Section 4,2.3.2, Separate Fire Area Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 135 Unit 4 Train B Emergency Diesel None A E Combustible Loading, E Generator Control Room - Fire Detection System, Detection 135: A Area YY Fire Area ID: YY (Unit 4) - Unit 4 Train 8 Emergency Diesel Generator Control Room Compliance Basis: NFPA 805, Section 4.2.4, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 1135 Unit 4 Train B Emergency Diesel None R, D. A E Combustible Loading: E Generator Control Room - Fire Detection System, All available - YY: R D A Area YY Fire Area ID: Z (Unit 3) - Unit 3 Train A Emergency Diesel Generator Building Compliance Basis: NFPA 805, Section 4.24, Performance Based Approach Required Required Suppression Detection Required Fire Fire Zone ID Description System System Protection Feature Required Fire Protection Feature and System Details 073 Unit 3 Train A Emergency Diesel D, A None Detection System, Alarm Point IQ. D A Generator Building - Fire Area Z Water Suppression, 3-10-847.

Fire Safety Ajislysis Data Manager (4.2) FPL - Turkey Run: 12J19/2013 15:34 Page: 67 of 58

Attachment to L-2014-071 Page 44 of 101 Attachment C Table C-2 NFPA 805 Required Fire Protection Systems and Features Fire Area ID: Z (Unit 4) - Unit 3 Train A Emergency Diesel Generator Building Compliance Basis: NFPA 805, Section 4.2.3.2, Separate Fire Area Required Requited Suppression Detection Required Fire Fire Zane ID Description System System Protection Feature - Required Fire Protection Feature and System Details 073 Unit 3 Train A Emergencv Diesel 6z - A None Detection System. Ala oiPoint 16: A Generator Building - Fire Area Z a 81 7 Fire Safety Analysis Data Manager (4.2) FPL - Trurkey Run: 12119/2013 15:34 Page: 58 of 58

Attachment to L-2014-071 Page 45 of 101 RAI PRA 01.y.01 In letter dated May 15, 2013, the licensee responded to PRA RAI Ol.y and indicated that the ZOI of an ignition source was not treated as being impacted by secondary combustibles when evaluating target damage distances; however, this response and the response to Fire Modeling RAI 01 .j in letter dated May 15, 2013 state that the impact of this nonconservatism is still being evaluated. Provide a summary and the results of the identified evaluation. Include an estimate of the impact of not including secondary combustibles on the risk results (i.e., CDF, LERF, ACDF and ALERF), or provide updated risk results using acceptable methods as part of the aggregate change-in-risk analysis requested in PRA RAI 29.

RESPONSE

The evaluation of the impact of secondary combustibles on the ignition source zone of influence (ZOI) involved updating the Fire PRA model using expanded fire ZOIs and updated applicable hot gas layer times. Fire scenario targets have been updated using the ZOIs provided in 'Combined Ignition Source -- Cable Tray Fire Scenario ZOIs for Turkey Point Nuclear Power Plant Applications.' The document provides generic ZOI dimensions for ignition sources impacted by a generic set of secondary combustible configurations. The range of ignition source - secondary combustible combinations were generated in order to bound specific Turkey Point configurations. The hot gas layer analysis has been updated using the time to hot gas layer provided in 'Evaluation of the Development and Timing of Hot Gas Layer Conditions in Generic PTN Fire Compartments with Secondary Combustibles'..

The document provides time-to-hot-gas-layer data for the same configurations provided in the ZOI document.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response (as well as RAI FM 01 .j) will be provided following NRC initial review and feedback on this response.

RAI PRA 01.z.i.01 In letter dated March 18, 2013, the licensee responded to PRA RAI 01.z.i and stated that "if a transient scenario ZOI would not impact more than one piece of equipment or conduit, no transient scenario was developed." There is no basis to categorically exclude such scenarios.

Clarify this statement, and explain how transient scenarios that may impact risk-relevant targets (e.g., those with high Conditional Core Damage Probabilities (CCDPs), pinch points, etc.) are appropriately reflected in the results. Provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, appropriately reflecting the contribution of transient fires that would impact only one piece of equipment.

RESPONSE

Transient scenarios were 'postulated' throughout every fire zone at Turkey Point where scenarios were defined. However, only those 'postulated' transient scenarios that would introduce a unique CCDP, not already captured by a defined fixed ignition source scenario, were defined and quantified for use in the Fire PRA risk totals. The intent of the PRA 01 .z.i response was to indicate that transient scenarios would not be postulated next to a fixed ignition source if that ignition source was the only equipment impacted by the transient scenario.

Attachment to L-2014-071 Page 46 of 101 To ensure that hot gas layer contributions from transient fires were captured, all fire zones quantified using individual ignition source fire scenarios had an additional 'dummy' transient scenario analyzed to which the fire zone total transient ignition frequency was applied. This captures the fraction of transient ignition frequency that is not directly modeled in a defined transient scenario but that could still cause a hot gas layer formation.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requanififcation. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 01.z.ii.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 01.z.ii and stated that transient fire scenarios were postulated behind open-back MCBs; however, it is unclear whether transient fire scenarios were postulated behind open-back back panels. If transient scenarios were not postulated behind all open-back panels, describe how this is consistent with acceptable methods; if not consistent, provide an estimate of the impact of not including secondary combustibles on the risk results (i.e., CDF, LERF, ACDF and ALERF).

Alternatively, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, including the placement of transient fires behind all open-backed panels in the MCR.

RESPONSE

Transient scenarios have been defined behind all open-back panels and Main Control Boards (MCBs) located in the Main Control Room (MCR).;

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI 03.01 In letter dated March 18, 2013, the licensee responded to PRA RAI 03 and stated that the seismic CDF estimate of I E-8/yr provided in the response "only addresses the seismic risk for an earthquake with a magnitude between the operating-basis earthquake and the design-basis earthquake (DBE), and does not include an estimate of the CDF due to earthquakes with a magnitude above that of the DBE." This makes use of frequencies obtained from the 1994 Lawrence Livermore National Laboratory seismic hazard curves (i.e., NUREG 1488, "Revised Livermore Seismic Hazard Estimates for 69 Nuclear Power Plant Sites East of the Rocky Mountains") in Iieu of the higher frequencies of the 2008 U.S. Geological Survey (USGS) seismic hazard curves reported in the September 2010 NRC staff memorandum as referenced in PRA RAI 03. In addition, it was noted that the response to PRA RAI 0l.h states in a recently performed seismic adequacy assessment that "[a]t plant HCLPF [high confidence in low probability of failure] levels of 0.3g and even 0.25g, the SCDF [seismic core damage frequency] is barely in the 10.6 range," implying at least a two order-of-magnitude difference.

An estimate of the total seismic risk is needed because the requested increase in risk is greater than "very small" in RG 1.174, and the total estimated risk without the seismic contribution is

Attachment to L-2014-071 Page 47 of 101 very close to the guideline that would indicate risk increase greater than "very small" may not be acceptable. Provide an estimate of the current seismic risk for all hazard magnitudes and summarize how that estimate was developed.

RESPONSE

A detailed, realistic seismic PRA does not exist for Turkey Point. The seismic CDF given in the original response used a frequency of a seismic event in the OBE to DBE range at the Turkey Point site of less than 1.OE-04 per year and combined this frequency with a plant-specific CCDP for a non-recoverable LOOP, to produce a seismic CDF. This is not considered to be a substitute for a detailed seismic PRA, but it was an attempt at producing a realistic, albeit crude, estimate of the seismic risk.

The September 2010 NRC staff memorandum, referenced in PRA RAI 03, reported a seismic CDF for the Turkey Point units of 5.9E-06 per year for the "simple average," using the 2008 U.S. Geological Survey (USGS) seismic hazard curves. Turkey Point was a Reduced Scope plant for IPEEE; therefore, the scope of the IPEEE seismic analysis consisted of seismic walkdowns which verified that the equipment could withstand the SSE-level seismic event.

Given this limited amount of seismic information, in the NRC staff memorandum, the seismic CDF was calculated assuming any earthquake with a magnitude above the SSE level would result in core damage, i.e., it took limited credit for any seismic robustness beyond the SSE level (credit which would be accounted for in a seismic PRA). Therefore, the seismic CDF in the NRC staff memorandum is a conservative,'bounding estimate. Neverthelessif the seismic CDF from the 2010 NRC staff memorandum is combined with the CDFs from the other contributors, the total estimated risk for Turkey Point remains below 1E-04 per year, keeping the requested risk increase in Region II as defined in RG .1.174. If the seismic LERF is assumed to be one-tenth that of the seismic CDF estimate (note that both the internal events and fire LERF-to-CDF ratios are less than one-tenth)', then the LERF estimate from the 2010 NRC staff memorandum is 5.9E-07 per year. Like the seismic CDF estimate, this is a conservative, bounding estimate, and, like the CDF estimate, if this seismic LERF is combined with the LERFs from the other contributors, the total estimated risk for Turkey Point remains below 1E-05 per year, keeping the requested risk increase in Region II as defined in RG 1.174. There is much work currently being done in the area of seismic risk, and it is expected that more realistic estimates of the seismic risk at Turkey Point will ensue.

RAI PRA 07.01 Relative to the counting and treatment of Bin 15 electrical cabinets, address the following:

a. Per Section 6.5.6 of NUJREG/CR-6850, fires originating from within "well-sealed electrical cabinets that have robustly-secured doors (and/or access panels) and that house only circuits below 440V" do not meet the definition of potentially challenging fires and therefore should be excluded from the counting process for Bin 15. By counting these cabinets as ignition sources within Bin 15 the frequencies applied to other cabinets are inappropriately diluted. Clarify that this guidance is being applied at Turkey Point.

b., In addition, all cabinets having circuits of 440V or greater should be counted for purposes of Bin 15 frequency apportionment based on the guidance in Section 6.5.6 of NUREG/CR-6850. Clarify that this guidance is being applied at Turkey Point.

Attachment to L-2014-071 Page 48 of 101

c. Per NUREG/CR-6850, cabinets below 440V that are not well-sealed and robustly-secured as well all cabinets above 440V are considered "potentially challenging" and should be assumed to propagate outside the cabinet. Clarify that this guidance is being applied at Turkey Point.

d. If the above guidance is not being followed, provide justification for deviating from the acceptable guidance and provide the results of a sensitivity analysis (i.e., CDF, LERF, ACDF and ALERF) , or provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, following the acceptable guidance.

RESPONSE

a. The counting methodology used did not exclude sealed panels below 440 V since excluding these panels from the counting process would indicate that the risk associated with a fire impacting the panel is not significant. This revision to the Bin 15 frequency will be incorporated into the final Fire PRA quantification.

b. All panels of 440 V or greater were counted for the Bin 15 frequency apportionment.

c. All panels below 440 V which are not well sealed and robustly-secured are counted and evaluated assuming propagation of a fire outside the panels. However, some panels of 440 V or greater which are well-sealed and robustly-secured in accordance with NUREG/CR-6850, Supplement 1 are not assumed to propagate a fire outside the cabinet. This is consistent with the guidance of NUREG/CR-6850, Supplement 1 which does not limit the definition of well-sealed and robustly-secured to any particular voltage level.

d. Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 08.01 In letter dated May 15, 2013 (ADAMS Accession No. ML13157A011), the licensee responded to PRA RAI 08 and did not provide the requested information to (1) justify the assumption that one half of MCR panels contain single-cable bundles and the other half contains multiple-cable bundles, (2) provide a quantitative basis that the assumption regarding heating, ventilation, air conditioning (HVAC) being unavailable for 10% of all MCR scenarios is conservative, or (3) address the use of NSP values less than 0.001 in the MCR abandonment analysis. Provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, (1) modeling all panels in the MCR as containing multiple-cable bundles, (2) failing the MCR HVAC control system only for those fire scenarios that will impact its operation, and (3) applying a minimum NSP of 0.001 per NUREG/CR-6850.

RESPONSE

Updated risk results will be provided in the final quantification incorporating the following model updates:

• The Main Control Room (MCR) abandonment frequency will be calculated utilizing the multiple cable bundle fire timing for all control room panels. However, where a

Attachment to L-2014-071 Page 49 of 101 walkdown confirms that a single cable bundle configuration is applicable to specific panels, the analysis will be revised to reflect single cable bundle for those panels.

The MCR abandonment frequency will be evaluated using the HVAC and door configuration that produces the most bounding abandonment times. However, where loss of HVAC impacts the results, a review of specific functions for individual panels will be performed to identify specific panel for which HVAC may be credited.

A review of the smoke temperature, optical density and layer height curves provided in the updated MCR Abandonment report Attachment 1 will be performed to determine which bins never reach the threshold value for each. Bins that never reach any of the three thresholds will have a 0 NSP value applied to reflect that the bin never causes an abandonment environment. Bins that do not meet this criteria will have an NSP applied based on the methodology of FAQ 08-0050, including a 0.001 minimum floor value.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2'd round RAI's would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 11.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 11 and did not provide the requested results of an HFE quantification process, such as that described in Section 5 of NNUREG-i921, or an analogous method to justify CCDP and conditional large early release frequency screening values used. Instead. the response provided a general description of a method to estimate the CCDP after MCR abandonment. The described method indicates that only one value (i.e., 0.056) is used as a CCDP for all MCR abandomnent scenarios for the compliant case. For the variant case, three values (i.e., 0.1, 0.2, and 1.0) of CCDP are used representing limited, moderate, and severe fire damage scenarios respectively. In the Table in the response, the "Basis for the CCDP used" column states that the bounding CCDPs (other than 1.0) are derived by assuming they are approximately double or four times the compliant case CCDP of 0.056. While this is mathematically correct, it is not a basis.

a. The method implies that the risk from every ignition source in the compliant case will be less than the risk from that ignition source in the variant case source because the compliant CCDP (i.e., 0.056) is always less than the variant CCDP (i.e., 0.1, 0.2, and 1.0), and the frequencies remain the same. Is this correct? If not, why not?

b. The response states that the variant case evaluation begins with a "calculated CCDP associated with the fire impacts ... [from] ... the fire event." The first entry in the Table states that some calculated CCDPs are less than .103. This implies that there are scenarios for which much more equipment is available than the minimal set assumed in the compliant case. Is this equipment being made available due to plant modification(s)? If not, clarify why the much larger compliant plant CCDP of 0.056 does not overestimate the compliant case risk and thereby underestimate the change in risk.

c. Provide the frequency (i.e., before applying the CCDPs) for the fire-induced main control room abandonment scenario bins. These would include the total abandonment

Attachment to L-2014-071 Page 50 of 101 frequency for the compliant case, and one total frequency for each of the three variant case bins. Presumably, the three variant case ignition frequencies will sum to the total variant case abandonment frequency. If not, please provide the total variant case abandonment frequency and explain why the sum differs from the total.

d. Summarize how the probability of the operators failing to shut down the plant is reflected in both the variant and the compliant MCR abandonment CCDPs.

e. In general, conservative evaluations to reduce unnecessary analytic effort are acceptable. In the change in risk calculation, overestimating the variant case risk and/or underestimating the compliant case risk yields a conservative result. If the value of 0.056 assumed for the compliant case is overestimated because alternative shutdown pathways that are available are not modeled, a nonconservative (underestimated) change in risk will result. Evaluate the change in risk calculations for the MCR to identify conservative and nonconservative assumptions, and discuss whether the net effect is conservative or nonconservative. Recognize the availability of a bounding approach where the delta-risk is equated to the total risk in lieu of detailed estimates.

RESPONSE

a. Yes. The compliant case CCDP is always lower than the associated variant case CCDP. However, for scenarios wherethe variant case CCDP is less than 0.056, a variant case CCDP of 0.1 is applied and the calculated variant case CCDP (value less than 0.056) is used for the compliant case CCDP. For these scenarios, the complexity of the shutdown is low and it is reasonable to consider the delta risk to be negligible or zero.

b. The scenarios with a low CCDP are associated with fires in panels with limited damage with respect to equipment required for post fire shutdown. As noted in the response to item a above the low calculated variant CCDP is used as the compliant CCDP and the variant CCDP is set to 0.1 for these scenarios.

c. This data is provided in the Fire Risk Evaluation report Table 2-1 for the U3 FRE for Fire Area MM and U4 FRE for Fire Area MM. Table 2-1 provides the data for the variant case. The ignition frequency, non-suppression probability and severity factor, are the same for the corresponding compliant case scenario. Only the abandonment scenarios (PTB-A) apply to the compliant case. The analysis and the individual scenario contributions will be updated based on the impact of other RAIs and will be included in the final post-second-round.RAI quantification. The sum of the abandonment variant case frequencies for all scenarios does sum to the total abandonment frequency.

d. A CCDP/CLERP is quantified for each variant case control room abandonment scenario assuming. failure of control room operator actions and crediting only outside control room actions. HEPs for these actions along with the extent of damage for the scenario are the basis for the CCDP/CLERP values. These CCDP/CLERP values are scaled up, as indicated in the response to round one RAI PRA- 11. The compliant case CCDP/CLERP is based on the use of the lower bound deterministic analysis compliant case CCDP of 0.056 specified above, with a lower CCDP/CLERP (equal to the variant case CCDP/CLERP) used for scenarios where the variant case CCDP/CLERP is less than 0.056.

Attachment to L-2014-071 Page 51 of 101

e. The question regarding conservatism of the methodology used for the PTN control room abandonment scenario appears to be primarily focused on the use of the 0.056 CCDP for the compliant case. This CCDP is considered conservative since the compliant case, based on current plant procedures, focuses on a shutdown path that relies on EDG power and one train of power remaining energized. The use of a compliant case CCDP that is comprised of the major failure probabilities related to an EDG and an AFW pump provides a bounding compliant case for this shutdown method.

There is no specific guidance on how to model the compliant plant in the Fire PRA. It has been recognized as a challenge by the industry and the NRC in FAQ 08-0054, Revision I which states:

"The definition of the variant condition may not always be easily defined.

Judgment may be necessary in order to calculate a change in risk. For example, pre-transition operator manual actions not taken at the Primary Control Station that are currently characterized as alternative shutdown (pre-transition) may not have a single, 'deterministically compliant condition' for comparison purposes, therefore some judgment may be necessary. One option would be to define a

'compliant case' that is not based on the actual fire area configuration, but based on a configuration that meets the deterministic criteria of Section 4.2.3 of NFPA 805. Regulatory Position 2.2.4 of RG 1.205 Rev. 1 provides clarification on this topic."

Given lack of specific criteria and challenges in correlating deterministic criteria to a MCR abandonment scenario, FPL believes the approach is reasonable and appropriate.

RAI PRA 13.01 In letter dated March 18, 2013, the licensee responded to PRA RAI 13 but did not provide a description of the methods used to calculate the changes in fire area risk reported in Appendix W, in exception for Fire Area MM (i.e., the main control room). For instance, for Fire Areas CC and HH, Section 5. 7 of the Fire Risk Evaluation report, as referenced in response to PRA RAI 11 from letter dated March 18, 2013, includes a discussion of two additional methods of detenmining delta risk not identified in response to PRA RAI 13; for other fire areas, it remains unclear how delta risk values were calculated. Given that the LAR does not describe either generically or specifically how delta: risk values were calculated:

a. Provide a description of the methods utilized to determine ACDF and ALERF for each fire area, including Fire Areas CC and HH, within the response to this RAI.

b. Discuss those cases for which the PRA model lacks sufficient resolution to model a VFDR, including any VFDRs not specifically modeled in the FPRA.

c. Elaborate whether methods utilized to determine ACDF and ALERF effectively bound ACDF and ALERF.

RESPONSE

RAI PRA 13.01.a.

The Fire Risk Evaluation Section 5.7 provides the following description of the calculation of delta risk for alternative shutdown fire areas (other than the control room, the methodology

Attachment to L-2014-071 Page 52 of 101 used for the control room was addressed in RAI PRA 13 in the first-round RAIs, with additional clarification on the compliant case provided in the response to RAI PRA 13.02).

Two PTN fire areas, other than the control room, utilized alternative shutdown strategies prior to NFPA 805 transition. These two fire areas are:

" Fire Area CC - Units 3 and 4 Auxiliary Building North-South Breezeway

" Fire Area HH - Units 3 and 4 Cable Spreading Room Cable Spreading Room:

For the cable spreading room the compliant-case CCDP was calculated based on a comparison of the variant case CCDP to the control room abandonment CCDP (see response to round-one RAI PRA 13 for control room abandonment CCDP basis). Scenarios with variant-case CCDP > 0.056 (control room abandonment compliant case CCDP) used 0.056 as the compliant-case CCDP. This ensures that credit for the Primary Control Station (PCS) as conservatively estimated by the abandonment CCDP is used for all cases where the combination of failures indicates the potential for a higher variant-case CCDP. For LERF quantification, the same process used above was applied to the scenario CCDP value and this CCDP value was multiplied by the ratio of CLERP to CCDP (thus incorporating the extent to which containment isolation versus core damage contributes to the quantification of LERF).

Units 3 and 4 Auxiliary Building North-South Breezeway The variant-case CDF for this area was treated in the same manner as non-alternate-shutdown areas with the quantification based on the fire impacts in each specific scenario. No compliant case was defined for this area with the delta CDF/LERF conservatively assumed to be equal to the variant-case CDF/LERF, thus assuming a compliant-case CDF/LERF of zero.

Non-Alternative Shutdown Fire Areas The methodology for non-alternative shutdown fire areas is provided in the Fire Risk Evaluation (FRE), Section 5.3. The discussion below is extracted from this section of the FRE.

Performing the FRE is an iterative process which compares a proposed risk-informed variant case to a deterministically compliant case. The compliant condition was created by manipulating the Fire PRA model to 'remove' the VFDRs, thereby creating a compliant condition. Fire PRA manipulationsmay involve 'toggling off or excluding specific PRA basic events to remove the potential fire-induced failure associated with the VFDR. The variant case represents the as-found condition or may include a risk-informed strategy that utilizes failure probabilities for recovery actions, plant modifications, or combinations of the two to mitigate the risk of the VFDRs. The variant condition represents the 'post-transition' plant configuration, not necessarily the "as built" plant. The necessary Fire PRA manipulations were documented in the FRE-specific Fire Area attachments.

The difference in risk between the variant case and the deterministically-compliant case was evaluated against the risk acceptance criteria of RG 1.174. The iterative analysis process includes the following steps.

1. For each area, the variant case represents the Fire PRA model, pre-FRE. The variant case provides the point of reference for the FRE, from which the difference in risk between full compliance with NFPA 805 deterministic requirements will be measured. Screening criteria for Fire Area risk were established to 'simplify' the evaluation process. The screening criteria initially established were less than I E-

Attachment to L-2014-071 Page 53 of 101 07/yr for CDF and less than 1E-08/yr for LERF. If the total area risk was less than the screening criteria, then the Fire Area CDF/LERF was considered as a surrogate for the delta CDF/LERF. In some cases fire areas with CDF values greater than the criteria may also be 'screened'. These cases are discussed in the specific fire area attachments to the FRE.

2. For each area, targets that were: failed in the Fire PRA model were reviewed, along with scenarios where the VFDR targets(s) were damaged. If any scenarios contained the VFDR targets, then Step 3 was followed. If no scenarios for a given' Fire Area resulted in damage to VFDR targets, then the variant. case (Step 1) became the post-transition baseline case and the remaining analysis steps were not needed (See Figure 5-3). The acceptability of the results and the recovery actions already in the Fire PRA model that mitigate the risk of the VFDRs were documented, followed by evaluation of defense-in-depth (DID) and safety margin.

3. For scenarios that were determined to damage the VFDR targets, the variant case was modified to reflect a deterministically-compliant case for each scenario that impacted the listed VFDRs. This 'compliant' case provided the fire risk if the plant configuration was modified to reroute or protect all of the associated, VFDR components and cables. The quantification of this case wAs:defined by setting the basic events in the Fire PRA model that are associated with.the VFDRs to their nominal, no-fire, random failure probability (i.e., the basic events associated with components related to the VFDR Were treated as if they were unaffected by fire).

4. The difference in CDF and LEI" from Step 3 and Step. 1 was calculated. If the total difference in CDF and LERF values between. scenarios in Step 3 and Step 1 for the Fire Area was less than 1E-07/yr for CDF and less than lE-08/yr for LERF, the risk presented by the VFDRs was typically considered acceptable without additional refinement to the model. In this 'instance,.the post-transition baseline case is the same as the variant case and the: analysis was considered complete and the remaining analysis steps were not needed. The acceptability of the results and the recovery actions already in the Fire PRA model that mitigate the risk of the VFDRs were documented, followed by evaluation of defense-in-depth (DID) and safety margin.

5. If the difference in CDF and LERF values between scenarios in Step 3 and Step 1 was greater than 1E-07/yr for CDF or.greater .than 1E-08/yr for LERF but less than the acceptance criteria, additional reviews. were typically performed.

6. If delta CDF and delta LERF in Step 4 were greater than the acceptance criteria, considerations for risk reduction wer-e p'ursued, such as model refinements, recovery actions, modifications, or combination thereof.

This iterative process was completed as many times as necessary to result in a post-transition baseline case that meets the acceptance criteria for risk.

Note that it is possible for the change in risk, following the consideration of modifications or other risk-reduction measures, to be a negative number, if the post-transition configuration has lower risk values than the compliant case.

Attachment to L-2014-071 Page 54 of 101

RESPONSE

RAI PRA 13.01.b.

The Fire Risk Evaluation documents the VFDRs not modeled in the Fire PRA and provides a basis for the lack of modeling of the VFDR. See Sections 2.2.2 of the FRE for each fire area.

Typical reasons for not modeling a VFDR include:

a. The first is for control room and cable.spreading room fire areas where the compliant case risk is based on a pre-defined, lower bound, CCDP for the compliant case, as discussed in the response to item a of this response, above.

b. The second reason that VFDRs are not considered in the Fire PRA is applicable to the Fire PRA model success criteria and is applicable for any fire area. These VFDRs deal with components or specific failure modes that are not modeled in the. Fire PRA as they do not correspond to core damage scenarios. An example of this, would be the failure to energize the pressurizer heaters which results in VFDRs for the deterministic analysis where the functional requirements, include the need to maintain pressurizer level within the pressurizer level instrumentation indicating range. The Equipment Selection task, as documented in the Component/Cable Selection Report, evaluates the Fire PRA model against the SSD analysis and addresses differences between these analyses. The PRA does not consider this a core damage sequence and therefore this VFDR would not be included in the delta risk as measured by the Fire PRA model. This variance in the Fire PRA success criteria also includes VFDRs related to Cold shutdown actions which are beyond the PRA mission time.

c. A third reason for not modeling a VFDR is related to a conservative limitation of redundant components modeled in the PRA. In some instances VFDRs were not modeled because. the specific component was not credited in the Fire PRA. This includes:

i. Diesel fuel oil transfer system is not modeled in the PRA due to the low risk and multiple options available for fuel oil supply given the time available for tank refill.

ii. The PRA credits the containment emergency coolers while the SSD analysis credits the normal containment coolers. PRA credit is taken for ability to isolate'the norMal containment coolers as long as the emergency containment coolers are available.

RESPONSE

RAI PRA 13.01.c.

The methods used for evaluating the ACDF and ALERF, as described above and in the first-round RAI response for the control room fire area, provide a conservative bounding analysis of the delta risk between the post-transition plant and a "compliant" plant configuration.

RAI PRA 13.02 In letter dated March 18, 2013 (ADAMS Accession No. ML13099A441), the licensee responded to PRA RAI 13 and stated that components associated with VFDRs were "allowed

Attachment to L-2014-071 Page 55 of 101 to be failed with a failure probability equal to the ignition frequency of the fire zone.. .in the variant case." Clarify this statement, elaborating on the failure probability utilized within the FPRA model for components associated with VFDRs. If the probabilities other than 1.0 (or TRUE) for fire-affected equipment were used to calculate scenario CCDPs, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29, assigning failure probabilities of 1.0 (or TRUE).

RESPONSE

Components associated with VFDRs were given a failure probability of 1.0 in the post-transition baseline (i.e., variant) case in the Fire PRA delta risk evaluations. The intent of the statement "allowed to be failed with a failure probability equal to the ignition frequency" was an attempt to specify that the frequency of the fire multiplied times the subsequent failure of the component is equal to the fire ignition frequency, implying a failure probability for the component of 1.0. Other factors impacting the failure probability are associated with severity factors (addressing the fraction of fires impacting the component given a split fraction associated with fire size/location which would impact the cable/component) and non-suppression probabilities (addressing the potential for fire suppression prior to fire damage to the component) which also impact the component's failure probability. In addition, for components where the failure mode was found to be associated with a hot short, a probability of spurious operation was applied per NUREG/CR-6850 (without CPT credit) as discussed in the response to RAI PRA-01 .i.01. Given that failure probabilities of 1.0 were used for fire damage, as modified by other factors described above, update of aggregate change is risk results is not required.

PRA RAI 16.01 LAR Attachment S, Table S-3, Item 18 does not provide a plan of action should the change in risk exceed risk acceptance guidelines. Revise this implementation item to include a plan of action to notify the NRC if risk acceptance guidelines are exceeded subsequent to completion of PRA-credited modifications and implementation items.

RESPONSE

The original LAR Attachment S, Table S-3, Item 18 stated "Update the Fire PRA Model, as necessary, after all modifications are complete and as-built" (Reference FPL Letter L-2012-092 dated June 28, 2012). This was revised in response to PRA RAI 16 (Reference Request for Additional Information dated March 15, 2013 ML13038A310) to read "Update the Fire PRA Model after all modifications and procedural changes are complete and as-built and all implementation items affecting the fire PRA results are complete. Review the results of the fire PRA compared to the final updated version in the LAR after all RAIs have been responded and accepted. Any CDF increase greater than I E-07/yr or LERF greater than I E-08/yr shall generate a corrective action to determine the cause of the risk change and if that risk change impacts the conclusions in the LAR, (Reference FPL Letter L-2013-086). This was in response to the original RAI that recluested that the validity of the reported risk and change in risk be verified after all items that could impact risk in the implementation of NFPA 805 were complete. The request included that an action plan should the as-built change in risk exceed the estimates reported in the LAR. The above quoted change in LAR Attachment S, Table S-3 Item 18 provides that action plan by stating that any risk increase larger than the self-approval limits in RG 1.205 Revision 1.

Attachment to L-2014-071 Page 56 of 101 This is similar to the response to a Duane Arnold RAI (RAI PRA 84 NextEra Letter NG 0287 ML13191A035) and accepted in the safety evaluation for Duane Arnold dated 9 2013 (ML13210A449). However, to clarify the plan to validate the risk estimates in the LAR, Attachment S Item 18 will be revised as follows: "Update the Fire PRA Model after all modifications and procedural changes are complete and as-built and all implementation items affecting the fire PRA results are complete. Review the results of the fire PRA compared to the final updated version in the LAR after all RAIs have been responded and accepted. This will be treated as a change evaluation and evaluated in accordance with procedure EN-AA-202-1004." This procedure contains the instructions for evaluating changes to the fire protection program and includes the RG 1.205 risk acceptance guidelines and provides a path to request a license amendment should those limits be exceeded. The part of the procedure dealing with NFPA 805 change evaluation was not available at the time of the original RAI response. The revised wording for Table S-3, Implementation Items, Item 18 is provided as part of the response to PRA 29.

RAI PRA 18.01 In letter dated March 18, 2013, the licensee responded to PRA RAI 18 and did not address location-specific attributes and considerations (e.g., physical congestion, radiological restrictions, limited floor space, etc.) and did not provide a sufficient basis for why postulated transient combustible fires have a heat release rate (HRR) distribution similar to that of an electrical motor fire. Further, the response does not indicate that a review of plant records was performed. As a result, address the following per the guidance endorsed by the memorandum dated June 21, 2012 (ADAMS Accession No. ML12171A583), from Joseph Giitter to Biff Bradley ("Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires"'):

a. Provide a discussion of location-specific attributes and considerations applicable to each individual fire area/zone that supports use of a 69 kilowatt (kW) transient fire HRR.

b. Provide justification that the 69 kW HRR distribution for electrical motor fires bounds the postulated transient fires in fire areas/zones with administrative controls.

In the response, address the full range of types and quantities of combustibles that are expected to be in each location and how administrative controls will enforce this range to preclude a greater than 69 kW transient fire.

c. Perform and document a review of past transient fire experience at Turkey Point as well as a review of iecords related to any violations of the transient combustible controls that may include both internal plant records (e.g., condition reports) and NRC inspection records (e.g., by residents or during triennials) to inform the development of the administrative controls.

Note that FAQ 12-0064 provides guidance on adjusting influence factors to reflect possible reductions in the likelihood of certain transients being located in particular areas; as a result, this guidance (and not a surrogate reduction in the HRR of such transients) is the appropriate way to account for any potential reductions in likelihood.

RESPONSE

The fire zones/areas crediting the reduced transient heat release rate are:

Attaclhnent to L-2014-071 Page 57 of 101 Fire Zone 058/Fire Area F - Unit 3 and 4 Auxiliary Building 18' Elevation Hallway Fire Zone 079A/Fire Area CC - Unit 3 and 4 Aux Building North-South Breezeway Fire Zone 098/Fire Area HH - Unit 3 and 4 Cable Spreading Room

a. Fire Zone 058 and Fire Zone 079A are hallways whichpreclude storage of significant quantities of combustibles given the relatively high traffic in these hallways. The cable spreading room is known to be a "high risk" fire zone and, although it contains ignition sources which will require periodic maintenance, the application of the stringent transient controls is considered viable given the limited activities and the high level of awareness of the potential risk in this zone:.

b. The controls to be imposed will be to restrict all transients in these zones with specific compensatory actions to be in place during timeframes where transients-must be in the zone to support a particular maintenance or testing activity. Therefore .the expected transient combustibles in the zone will be negligible, but a 69 kW HRR is assumed to bound potential violations of the new transient controls.

The 69 kW HRR is based on an evaluation of a potential violation of the administrative controls to be implemented in these zones where administrative procedures will implement a zero transient combustible control criteria. The expectation is that the implementation of a zero transient Corhbustiblelimit will significantly reduce the size of.potential transients which could be placed in the zone in violation of the applied limits. This type of transient control is a newly imposed criterion that will require a monitoring programto address future adherence to the requirements. The results of post transition. nionitoring with respect to these controls will be the basis for implementation of appropriate corrective actions should violations occur. This review will also asses .the configuration and potential size of any such violations and determine if the use of the 69 kW HRR criteria is appropriate.

Additional bases for the use of the 69 kW HRR are provided below:

" PTN is implementing additional administrative controls such as a fire watch for conditions in which transients are stored in these areas.

" Areas that have transient administrative controls will not have stock piles of paper, cardboard, scrap wood or trash stored in these areas.

" The transient fire heat release rate distribution specified in NUREG/CR-6850 as a 317 kW (300 Btu/s) 98th percentile peak heat release rate fire is considered to be generically applicable to nuclear power plants. The PTN plant does not differ in any significant manner with respect to its transient combustible controls to warrant a significant increase or decrease in the applicable heat release rate profile. However, for areas that have been designated as "no transient combustible areas," to address the potential for violation of these controls., a 69 kW (65 Btu/s) 98th percentile peak heat release rate fire was applied. This heat release rate is considered appropriate given the unlikely event that transients are stored in these areas contrary to the controls imposed.

Any such violations are expected to be of a smaller size than the typical transient HRR configuration.

" The 69 kW (65 Btuls) heat release rate was defined based on the heat release rate specified in NUREG/CR-6850 for a motor fire given that the most likely transient fire in a zone with limited transients would be associated with temporary cabling since .this

Attachment to L-2014-071 Page 58 of 101 configuration would provide both the ignition source (energized temporary cabling) and combustible (cable insulation). The motor configuration would resemble such a transient fire.

0 Monitoring of the controls and evaluation of their effectiveness will provide a basis for assessing the appropriateness of this HRR as will the monitoring of other transient fires at PTN and industry wide with respect to the use of the nominal 317 kW (300 Btu/s) peak heat release rate transient fire.

0 A letter dated September 27, 2011, from NEI to NRC, B. Bradley to D. Harrison, "Recent Fire.PRA Methods Review Panel Decisions: Clarifications for Transient Fires and Alignment for Pump Oil Fires," Attachment 1, "Description of Treatment for Transient Fires," and Attachment 3, "Panel Decision," allows a lower heat release rate to be chosen for transient fires based on "the specific attributes and considerations applicable to that location." The letter suggests that "plant administrative controls should be considered in the appropriate HRR for a postulated transient fire" and that "a lower screening HRR can be used for individual plant specific locations if the 317 kW value is judged to be unrealistic given the specific attributes and considerations applicable to that location.' The use of this method was endorsed by the June 21, 2012 letter from the NRC to NEI (ML12171A583), with minor exceptions unrelated to the PTN treatment. This endorsement came in response to EPRI 1011989 which states that from a practical standpoint that a plant can have a "range of HRR values being applied in a nuclear power plant fire PRA. Locations within the plant that are under more rigorous controls or that have greater restrictions with respect to the introduction, handling, and placement of combustibles and/or the performance of hot work would be expected to have a lower HRR applied as compared to locations that have less rigorous controls and/or restrictions." Theuse of the lower heat release rate in the areas with significantly increased transient controls is considered to be consistent with this guidance.

c. A review of transient controls during the period for November 2009 to April 2012.

Violations of transient controls during this period were found to be primarily related to outage activities, many of which were related to significant scope Extended Power Uprate.(EPU) outages during this period. A review of documentation to detennine if any transient fires had been experienced was performed. No transient fires were identified in this review. A review of NRC inspection findings did not identify any transient control issues.

A markup of LAR Table S-3 which identifies the implementation of the controls specified in the discussion above is provided in the response to PRA 29.

RAI PRA 19.01 In letter dated March 18, 2013, the licensee responded to PRA RAI 19 and indicated that the availability of instrumentation for operator actions credited in the FPRA is predicated on the safe shutdown analysis. Describe how fire-induced instrument failure (including no readings as well as incorrect or misleading readings) is addressed in the FPRA HRA for systems, functions and/or equipment that are credited within the PRA but that were not addressed by the safe shutdown analysis.

Attaclunent to L-2014-071 Page 59 of 101

RESPONSE

As part of the task of replacing the screening HEP values with detailed FPRA HEP failure probabilities, FPRA-specific HEPs are being added to the quantification fault tree including instrumentation cues. The required cues are correlated to SSD analysis instrumentation which is identified as available instrumentation in the post-fire shutdown procedures. This imposes a failure of the HEP in any scenario where all associated cues are lost due to fire damage. The treatment of cues is consistent with NUREG- 1921, specifically discussion regarding failure of cues due to fire in accordance with NUREG-l1921 section 4.5.5.

During operator interviews, it was confirmed that an operator would always verify a cue through other redundant instrumentation or equipment cues prior to taking an action. Per the ASME/ANS Standard RA-Sa-2009, Supporting Requirement ES-C2, Capability Category II, only a single spurious indication is to be assumed. Therefore, incorrect/lmisleading readings would not impact post-fire shutdown.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response, updated HRA analysis, will be provided following NRC initial review and feedback on this response.

PRA RAI 22.01 In letter dated April 16, 2013, the licensee responded to PRA RAI 22 and provided neither the documentation of gap assessments performed nor their surmnarized results. In addition, it is unclear from the response whether the most recent gap assessment considers the clarifications and qualifications of RG 1.200, Revision 2, "Ani Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." Describe whether the internal events PRA gap analyses discussed in response to PRA RAI 22 utilized the Nuclear Energy Institute (NEI) document, NEI 00-02, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," self-assessment process as supplemented by clarifications and qualifications contained in Appendix B to RG 1.200, Revision 2. If the aforementioned process was implemented,, provide the summarized results of the latest gap assessment in a manner analogous to the format in which F&Os are presented and dispositioned in Attachments U and V of the LAR, including an assessment of each identified gap's impact on the FPRA. If not, perform and provide the similarly summarized results of a self-assessment of the internal events PRA model that addresses the clarifications and qualifications in Appendix B of RG 1.200 on the use of NEI 00-02 peer review process guidance.

RESPONSE

As described in the original response, after the ASME PRA Standard and RG i.200 were

.issued, the results of the NEI 00-02 global peer review of the Turkey Point PRA were re-evaluated using the Standard and RG 1.200 using the RG 1.200 Thble B-4 as a guide to ensure that for each Standard Supporting Requirement (SR), the, appropriate NEI'00-02 guidance and their associated NEI 00-02 peer review findings were reviewed and applied to the appropriate Standard SR. The Industry Self-Assessment Actions and Regulatory Position from RG 1.200, Revision 2, for each SR were also considered for each of the Standard SRs. No new F&Os were created during this process, as the scope and requirements of the NEI 00-02 process and

Attachment to L-2014-071 Page 60 of 101 the Standard were very similar. Focused peer reviews were conducted in 2011 and 2013, and their findings were integrated into the overall gap analysis.

Two tables are provided below. The first lists those ASME PRA Standard supporting requirements that do not meet the Category II requirements, and their anticipated impact on the fire PRA. The second lists the findings (F&Os) that have yet to be resolved, and their anticipated impact on the fire PRA.

Attachment to L-2014-071 Page 61 of 101 Turkey Point Gap Analysis - ASME PRA Standard (ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2 Supporting Requirements Not Met SeclD

" ;IF SMECategory -_______ II

_ I_ t~

Met?_M Co '

Co..in mentss FPR Comments. .mpa '..... *, i cIt on Fire PRA IE-C14 In the ISLOCA frequency No The ISLOCA rhis SR is not met. ISLOCA modeling has analysis, INCLUDE the analysis PTN-BFJR- :a) The PTN ISLOCA evaluation considers the no effect on the fire following features of plant 07-003 addresses configurations of the potential pathways. However, PRA results.

and procedures that these features. several F&Os were identified to improve the evaluation influence the ISLOCA and allow IE-C14 to be met.

frequency: :al) Pen. #43 - The TBCCW penetration evaluation is (a) configuration of :erformed for one thermal barrier cooler, yet there are potential pathways :hree RCPs, each with a TB cooling coil. The IE including numbers and frequency for the TBCCW should address the total types of valves and their frequency from all three TB cooling coils (refer to F&O relevant failure modes EI-C14-1).

and the existence, size, :a2) Pen. #43 - The evaluation of this penetration and positioning of relief ncludes credit for a local manual operator action to valves solate the penetration. This action should not be (b) provision of protective credited without demonstrating the time available and interlocks iabitability for success (refer to F&O EI-C14-2).

(c) relevant surveillance :a3) Pen. #3 - The thermal barrier cooling supply line test procedures vas screened from detailed evaluation without a (d) the capability of proper basis. This penetration should be evaluated secondary system piping further as a potential ISLOCA pathway (refer to F&O (e) isolation capabilities E-C14-3).

given high ýa4) Pen. #1 - Accident sequence cutsets for the small flow/differential pressure SLOCA involving the RHR SDC suction line did not conditions that might consider that the common suction piping beyond the exist following breach of RHR pumps could affect the reliability of HHSI and the secondary system RWST, which are credited for mitigation. These sequences should be re-assessed without credit for the associated Unit HHSI and RWST (refer to F&O IE-C14-4).

(a5) Pen. # 58/59/60/18 - These penetrations are screened from detailed evaluation based on 9quivalency to screening criterion (c). The basis for this screening should be strengthened (refer to F&O IE-C14-5).

(a6) The updated analysis was performed in the 2007-2008 timeframe and does not take advantage of some

f the latest industry practices (refer to F&O IE-C14-6).

(a7) Table 1 "Potential ISLOCA Flow Paths" of the ISLOCA evaluation should include more detail relative to screening, particularly for penetrations #13 and #14 (refer to F&O IE-C14-7).

(b)Section xx of the evaluation describes the protective interlocks available for control of the PIVs - particularly for RHR (Pen. #1). These interlocks and controls reduce the likelihood of spurious opening and are considered in the IE evaluation. No issues identified.

(c) Section xx of the evaluation describes the relevant surveillance test procedures that control testing and alignment of the PIVs considered in the IR evaluation.

No issues identified.

(d) The capability of secondary system piping is explicitly included in the IE evaluation. The pipe failure frequencies used to model secondary pipe failure were reviewed and judged reasonable and defendable for modeling these failures. No issues identified.

(e) The isolation capabilities given high flow/differential pressure conditions was verified. The only valve isolation mitigation credited is MOV-*-626. According to the CCW DBD, this valve is designed to close under full RCS shutoff dp. No issues identified.

Attachment to L-2014-071 Page 62 of 101 SeciD Met? Comments iASMECategoryjI FPR Comments lImpact on Fire PRA SY-C3 DOCUMENT the sources Draft The uncertainty The uncertainty of model uncertainty and analyses for the analysis only needs to related assumptions (as system analyses be formalized. No identified in QU-E1 and have been impact on fire PRA.

• QU-E2) associated with completed and are the systems analysis. documented in the Uncertainty Analysis

_________ ___ ~Notebook_ _ _ _ _ _ _ _ _ _

DA-D6 USE generic common No Documentation of This SR is not met. The review of plant-cause failure probabilities, the CCF data for the The revised CCF model uses appropriate alpha factor specific data for CCF consistent with available PTN model is values from the NRC website, "CCF Parameter events is not expected plant experience. included in PTN Estimations, 2007 Update", http://nrcoe.inl.gov/results/ to result in the EVALUATE the common calculation PTN- CCF/ParamEst2007/ccfparamest.htm, September discovery of any such cause failure probabilities BFJR-08-012. 2008. The CCF probabilities are evaluated consistent events.

in a manner consistent Generic alpha with the component boundaries. However, two findings The CCF modeling in with the component factors are used to *were identified that are significant enough so that this boundaries. generate the SR is not met. the system initiating common cause Two F&Os were identified: the absence of a event fault trees has failure probabilities. documented review of plant failure data for common been reviewed and The CCF data is cause events (see DA-D6-01), and the absence of found to be applied consistent CCFs in system initiating event models (see DA-D6- appropriate.

with the PRA 02). No impact on fire PRA.

component boundaries.

1QU-D5 IREVIEW a sampling of No This has been done Review needs to be nonsignificant accident for the PTN PSA ,formalized. No impact cutsets or sequences to model quantification on fire PRA.

determine they are results, but not reasonable and have formally physical meaning. documented. Add review to the model update.

LE-F1 PERFORM a quantitative No Section 5.2 of the SR Met: (CC I). IThis finding only evaluation of the relative Level 2 calculation Reviewed PTN-BJFR-99-010, Rev. 1, specifically, addresses the contribution to LERF from PTN-BFJR-99-010. Section 5.2. Endstate frequency totals are given in categorization of LERF plant damage states and Table 5. And results by release category are given in results. This will be significant LERF Table 6. However, results using the Plant Damage done in the next model contributors from Table State definitions of Section 4.2 are not provided, update, but will have 2-2.8-3. Strictly speaking by the letter of the SR, CC II is not no effect on the fire met because relative contribution to LERF by PDS is PRA.

not shown, although information is available to provide such data.

LE-G3 DOCUMENT the relative No Section 5.2 of the SR Met: (CC I) This finding only contribution of Level 2 calculation Reviewed PTN-BJFR-99-010, Rev. 1. Endstate addresses the contributors (i.e., plant PTN-BFJR-99-010. frequency totals are given in Table 5. And results by categorization of LERF damage states, accident release category are given in Table 6. However, results results. This will be progression sequences, using the Plant Damage State definitions of Section 4.2 done in the next model phenomena, containment are not provided. Strictly speaking by the letter of the update, but will have challenges, containment SR, CC II is not met because relative contribution to no effect on the fire failure modes) to LERF. LERF by PDS is not shown, although information is PRA.

available to provide such data. So, similar to F&O LE-F1-01, strictly speaking by the letter of the SR, CC II is not met because contribution to LERF by PDS is not documented.

Attachment to L-2014-071 Page 63 of 101 SecilD SMECategory, 1 Met? Comments FPR;Comments"ý, impact on Fire PRA LE-G5 IDENTIFY limitations in No Section 5.2.1 of the SR Not Met. This finding addresses the LERF analysis that Level 2 calculation Reviewed PTN-BJFR-99-010, Rev. 1 Section 5.2. The uncertainty in the would impact PTN-BFJR-99-010. only limitation cited regarding the impact on LERF calculations.

applications. applications is with respect to low Fussell-Vesely (FV) The stated I and risk achievement worth (RAW) values. That does uncertainties are found not meet the intent or letter of this SR. There can be in all LERF potentially large sources of uncertainty on physical calculations. No phenomena such as thermally induced SGTR, and the impact on fire PRA. I break size and location in piping for ISLOCA.

Additionally, there are uncertainties in the determination of radionuclide release fractions and the assignment of CET endstates. The Level 2 LERF analysis is a representation of known severe accident phenomena, but uncertainties in the LERF results may impact certain applications such as the use of RG 1.174.

Attachment to L-2014-071 Page 64 of 101 Turkey Point Gap Analysis - ASME PRA Standard(ASME/ANS RA-Sa-2009) and RG 1.200, Revision 2 Facts and Observations (F& Os) Not Resolved rF&O ssue* Basis for Significance Possible Resolution Impa on Fire PRA DA-D5-01 For several CCF groups, a "global The missing CCF wo alternatives. The missing CCF The peer reviewers did not common cause event" (as described :contribution from the 5- terms could be added to the expect this to have a at the end of Section 4.2 of PTN- 0f-6 term (or the 2-of-4 CAFTA fault trees and CCF basic signification effect on the BFJR-2008-012, Rev. 0) is used. and 3-of-4) should not events calculated for the new results.

While this is a reasonable be significant since the terms. A simpler alternative is to simplification, the global common 6-of-6 term (or 4-of-4 revise the calculation of the a6 term cause event needs to account for the term) is included and to include the missing a5 value.

common cause combinations that are should dominate the Thus, a6' = a5 + a6. This not included explicitly. However, for OCF contribution. overestimates the a5 contribution, several 6-component groups (AFW since it is applied to the case where AOVs FTO, AFW CVs FTO, AFW all 6 components fail, but this MOVs FTO), the 5-of-6 term was not should be a small and conservative included and the 6-of-6 term was not approximation. (Similar correction adjusted. A similar issue appears to for the 4-component group, a4' =

be present for SG SVs FTO (4- a2 + a3 + a4).

component group), where only the 4-jof-4 term is included (the 2-of-4 and 3-of-4 terms are missing and the 4-of-4 term was not adjusted).

DA-D6-01 The CCF notebook did not include a The SR includes a Review plant-specific component The review of plant-specific review of plant failure data for check to assure the failure events from the most recent data for CCF events is not common cause events. CCF parameters are :data update to identify any common expected to result in the consistent with available cause failures. If CCFs are discovery of any such events.

plant-specific identified, verify that the CCF is experience, modeled for the specific component Not expected to have an impact and failure mode. If this data on the fire PRA.

indicates a significantly larger fraction of failures are CCFs than the generic CCF parameters would predict, plant-specific CCF parameters should be calculated. If the data is limited (one or two failures in a specific component group), this would not be sufficient evidence to justify plant-specific CCF parameters.

DA-D6-02 Section 3.0 of the CCF Notebook. CCFs may be significant Provide a basis for excluding CCFs CCFs are appropriately includes the assumption that CCFs contributors to system from system initiating events and included for the components in are not included in fault tree initiating initiating events. include CCFs where a basis for the initiating event fault trees.

events with year-long mission times exclusion cannot be established.

due to excessive conservatism in For example, include CCF in No impact on fire PRA applying CCF factors that are system initiating event models only developed for 24-hr mission time. for active components that are in However, this is not sufficient basis fori the same configuration (i.e.,

excluding CCFs for fault tree IE between normally operating pumps models. in the same system but not between operating and standby pumps in the same system).

Attachment to L-2014-071 Page 65 of 101 F.... Issue Basis for Significance Possible Resolution .Impact I on Fire PRA..:

IE-C14-01 RCP TBHX rupture probability - The Depending on the Assess the tube rupture original ISLOCA frequency will have no IE frequency for tube rupture is based application of the tube source data and whether it is impact on the fire PRA.

on a Reference 5 value of 3.48E-08/hr rupture data, the applicable to each thermal barrier (peer review did not verify this thermal barrier heat cooler/RCP. Revise initiator reference) for "HX Tube External Leak exchanger ISLOCA %ZZISLTBCCW and document any Large >50 gpm". This hourly initiator frequency and changes or basis accordingly.

frequency is multiplied by 8760hr/yr associated core for an annual IE frequency of 3.05E- damage results appear 04/yr. Depending on the application to be low by a factor of of the data, this IE frequency could be three because the applied at each RCP, thus event tree ISLOCA event could top event "RCP TBHX Tubes Intact?" occur at each RCP. This would be multiplied by a factor of should not have a

3. Applicability of the TBHX data to significant effect on the one or all RCPs should be overall ISLOCA results examined/documented for impact on given that TBCCW the total %ZZISLTBCCW ISLOCA sequences are initiator/results, of very low frequency.

IE-C14-02 Manual operator action is credited for TBCCW isolation is Evaluate and document whether- ISLOCA frequency will have no local manual closure of MOV-*-626 credited in the ISLOCA the operator action should be impact on the fire PRA.

(should it fail to close) and/or to local evaluation. Success of credited and remove credit for the closure of manual valve *-736. isolation is important action if it cannot be justified Operator success ensures that the because it both COW piping remains intact. Although mitigates the ISLOCA the HEP for the local action is 0.5, the event and maintains time window basis should document CCW function for to ensure that the operator has cooling of essential sufficient time to perform these equipment.-The basis actions before the CCW piping for the operator action boundary fails. (adequate lime and habitability) should be demonstrated or no credit should be given for the local manual action.

E-C14-03 Thermal Barrier ISLOCA IE This TBCCW Evaluate and document the ISLOCA frequency will have no Frequency - RCP Thermal Barrier penetration could TBCCW supply penetration for impact on the fire PRA.

COW Supply Penetration #3 - This significantly contribute possible ISLOCA initiating events.

penetration is not evaluated for to the existing TBCCW Should also assess the impact on potential ISLOCA contribution. This ISLOCA CDF CCW return line from RCP motor penetration is protected by two sequences. cooling and lifting of RV-729 if V-normally open, active check valves 712A fails open. Ensure that these (717 and 721A/B/C) inside penetrations are also identified in containment and two normally open Table 1, list of penetrations.

MOVs (716A/B) outside containment. The associated piping inside containment appears to be designed for full RCS pressure. However, given a thermal barrier tube breach, the active check valves could fail to close (w/CCF). The active failure of the outboard MOVs (also w/CCF) may be highly unreliable due to low differential pressure design capability and lack of relevant closure signals, and there might not be sufficient time for manual action.

Failure of this penetration should be assessed for possible contribution to the TBCCW ISLOCA event frequency and sequences.

Attachment to L-2014-071 Page 66 of 101 F 0 Issue. Basis for Significancel PO6S.be.Resolution .. - Ipact on .FirePRA IE-C14-04 ISLOCA assessment of Penetration 1 Current RHR Small Evaluate and document the RHR ISLOCA frequency will have no (RHR SDC suction line) did not ISLOCA event 'small ISLOCA sequences taking no impact on the fire PRA.

consider that the common suction sequences improperly credit for associated Unit HHSI piping beyond the RHR pumps could credit the mitigation pumps and RWST.

be affected by the over-pressurization capability of the event. This would impact the function associated Unit's HHSI of the high head SI pumps and the pumps and RWST.

RWST (and Containment Spray Removing credit for pumps, which are not important in HHSI and RWST will ISLOCA scenarios). As a result, the increase some of the current RHR small ISLOCA event RHR small ISLOCA sequences apply too much credit for event contribution by 3 the associated Unit's RWST and HHSI orders of magnitude; pumps. from the range of E-12/yr to E-09/yr. Even with this significant increase, these sequences will still contribute only a small amount to the overall ISLOCA risk.

ilE-C14-05 Penetrations 58/59/60: (HHSI cold leg These penetrations Review these penetrations and The assumption here is that 3 injection) - These penetrations are have been qualitatively provide further basis for screening. closed check valves are not qualitatively screened from further screened, however it is believed to be equivalent to 3 detailed evaluation on the basis that not clear that their manually isolated manual

.... "the combination of three check physical configurations valves. Still, the probability of 3 valves is equivalent to three are equivalent to the closed check valves opening locked/closed isolation valves", for intended configuration against pressure is likely to be meeting NUREG/CR-5928 criterion defined in the screening adequately low for screening.

(c), systems isolated by redundant criterion. Additional normally closed and locked manual basis is needed to Not expected to significantly valves that are independently verified justify the screening or impact the fire PRA.

to be closed and locked before plant perform a detailed startup". This comment is also evaluation of these applicable to Penetration 18. penetrations.

Additional basis is needed to support this equivalency assertion for screening these penetrations.

IE-C14-06 The PTN ISLOCA analysis is based This is not a significant Consider updating the ISLOCA ISLOCA frequency will have no on early NUREG information and F&O. In general, the evaluation to current industry mpact on the fire PRA.

industry practice, which continue to ISLOCA evaluation ,practice and reference material. It is provide a reasonable source of method and notebook noted that there are limitations in inputs/practice for consideration in documentation could the WCAP-1 7154, Revision 1 ISLOCA modeling. In general benefit by methodology and its complete however, the evaluation might benefit reviewing/consulting adoption is not recommended.

from aspects of the latest industry aspects of the latest ISLOCA best practice/methodology ISLOCA industry presented in WCAP-17154, Rev.1. information and best practices including WCAP-17154, Revision 1, "ISLOCA Risk Model"

Attachment to L-2014-071 Page 67 of 101

.Issueasis for Signicance]Possible R~solution impact on Fire PRA IE-C14-07 Table 1 "Potential ISLOCA Flow Improve documentation Consider updating the ISLOCA ISLOCA frequency will have no Paths" - Consider adding more detail of ISLOCA scope and report to improve the details in impact on the fire PRA.

in the ISL Screening Results column. screening. Table 1, primarily the column For example, Penetrations 13 and 14 information under "ISL Screening (Letdown and Charging) may not Results" cleanly screen. Both systems interface with low pressure systems (letdown-purification piping and charging-pump suction). Typically there are redundant isolation means to isolate - thus IE frequency should be low. However, this cannot be concluded from the table details. Also, Penetration 3, "RCP CCW Supply" indicates that this penetration was screened based on "not connected to the RCS". However, this penetration provides the CCW supply to RCP thermal barrier cooling and should be assessed (refer to F&O IE-C14-2).

LE-D2-01 Electrical penetration assembly failure There is no explicit Perform a scoping assessment of For containment isolation, the modes have been found to be discussion of the the potential impact of electrical Level 2 update incorporated the important contributors to overall thermal-mechanical. :penetration thermal-mechanical . existing containment isolation containment fragility at other large dry impact of accident response to severe accidents, analysis; it did not revisit this PWRs, and in at least 2 instances, progression on Consider using some of the issue directly. As for tend to be the most limiting in terms of containment seals and following references: NUREG/CR- containment strength, this was ultimate failure pressure. Additionally, electrical penetration K944, CR-5083, CR-5096, CR-. also something that was early studies at Sandia National assemblies. ,15118, and CR-5334. provided and was not re-Laboratories have considered the investigated, so would not have potential impact of very high (beyond affected the analysis directly.

design basis) temperatures on The place. in the Level 2 model elastomer seals (this latter issue is where this would have an effect more critical for small volume would be the "Containment containments such as BWR Mark I). Failure at Vessel Breach" events, which were determined via NUREG sources to be minimal. It is not known whether these referenced NUREGs already factored such considerations into their containment strength estimates and failure probabilities, but it is not expected to have a significant effect.

Not expected to have a significant impact on the fire PRA.

LE-FI-01 Endstate frequency totals are given in PDS relative Perform summary calculation to This finding only addresses the Table 5 of the Level 2 notebook, PTN- contribution to LERF is quantify PDS relative contribution to categorization of LERF results.

BJFR-99-010, Rev. 1, and results by not provided as. LERF. This will be done in the next "elease category are given in Table 6. specified in the SR. model update, but will have no However, results using the Plant effect on the fire PRA.

Damage State definitions of Section 4.2 are not provided. CC II is not met 2ecause relative contribution to LERF

y PDS is not shown, although nformation is available to provide such data.

Attachment to L-2014-071 Page 68 of 101 F&O Is. e Basis for Significance Possible Resolution Impact on Fire PRA LE-G5-01 There is no discussion of limitations of Does not meet the Provide a discussion of possible This finding only addresses the severe accident understanding and intent of providing a limitations of the LERF analysis categorization of LERF results.

modeling. This includes such matters discussion regarding based on, for example, limitations This will be done in the next as the impact of uncertainty regarding limitations on the on the state of severe accident model update, but will have no thermally induced SGTR on understanding of severe understanding and level 2 PRA effect on the fire PRA.

quantification, the uncertainty of accident analysis. Briefly describe how key ISLOCA break size and location on phenomenology, and uncertainties in the LERF timing and source term, and the how the Level 2 quantification could impact risk-assignment of CET to endstates. modeling uncertainties informed changes to the licensing Conservative treatment of some could impact LERF basis under RG 1.174, for example.

phenomena can affect LERF quantification and quantification, which in turn impacts potential risk-informed LERF and delta LERF results when applications.

applying RG 1.174 guidelines in risk-informed changes to the licensing basis, for example.

RAI 23.01 In letter dated April 16, 2013, the licensee responded to PRA RAI 23 and simply states that the HRA reviewers' resumes did not refer to the proper experience and therefore did not provide sufficient information to demonstrate that, contrary to the available information, the reviewers were qualified. Explicitly summarize how the HRA reviewers meet the qualification requirements in the American Society of Mechanical Engineers/American Nuclear Society PRA Standard, as provided in Section 1-6.2, with particular emphasis on the following (below) as related to a focused-scope peer review (termed a "review for a PRA upgrade" in the Standard): The peer review team members individually shall be ...

experienced in performing the activities related to the PRA Elements for which the reviewer is assigned. When a peer review is being performed on a PRA upgrade, reviewers shall have knowledge and experience appropriate for the specific PRA Elements being reviewed.

However, the other requirements of this Section shall also apply. The peer reviewer shall also be knowledgeable (by direct experience) of the specific methodology, code, tool, or approach (e.g., accident sequence support state approach, MAAP code, THERP method) that was used in the PRA Element assigned for review. Understanding and competence in the assigned area shall be demonstrated by the range of the individual's experience in the number of different, independent activities performed in the assigned area, as well as the different levels of complexity of these activities. Include in this discussion actual experience and any publications or other activities that demonstrate the individuals' qualifications as experts for the review element.

RESPONSE

The revised resumes of the peer review team members, taken from Revision 1 of the focused peer review team report, are givenbelow, with references to HRA experience highlighted in bold font.

Attachment to L-20 14-071 Page 69 of 101 JODINE M. JANSEN VEHEC EDUCATION Texas A&M University: B.S., Nuclear Engineering (1990)

University of Tennessee: M.S., Environmental Engineering (1997)

EXPERIENCE

SUMMARY

Mrs. Jansen Vehec has over 20 years of experience peiforming PSA and safety assessment evaluationsfor both nuclear and non-nuclearfacilities. Mrs. Jansen Vehec also has experience in evaluating nuclear waste, hazardous waste, fissile waste, and mixed waste operations. Her safety assessment experience includes performanceof probabilistic risk analysis, reliability, availability, and maintainability (RAM) assessment safeo, assessment, accident sequence assessment and fire dispersion analysis. Mrs. Jansen Vehec has also been the project managerand lead analystfor seismic andfire risk assessments.

Mrs. Jansen Vehec has experience in the development and review o safety analysis reports, limiting conditionsfor operation,and other tipes of safetj assessments. She has knowledge of and experience with several of the most utilized PSA software packages, including the CAFTA and Risk Spectrum suite of codes, the EPRI calculatorand SAROS HRA toolbox, the FRANC and FRANX fire/flooding databaseapplication, and has an application-levelof experience with the Modular Accident Analysis Program (MAAP).

PROFESSIONAL EXPERIENCE February 2007 to Present, Reliability and Safety Consulting Engineers, Inc.

Mrs. Jansen is currently the lead engineer and project manager for the Waterford Steam Electric Station Unit 3 (WSES-3) internal events PRA update. As the project manager she is responsible for ensuring all technical elements are updated on schedule and are completed within budget. As the technical lead, she is responsible for perforning and/or reviewing the. updates to the accident sequence, initiating events, success criteria and LERF portions of the PRA. She is also responsible for ensuring the technical accuracy and completeness of the human reliability, data' system modeling, MAAP, and quantification portions of the PRA.

Mrs. Jansen Vehec was the lead engineer for the resolution of the NRCs Request for Additional Information (RAI) on the WSES-3 Radiological Release portion of their NFPA-805 License Amendment Request (LAR) submittal. In this capacity she was responsible for identifying all potential radiological areas at the site, the quantity and type of radiological materials and contamination potentially present in each area, and verifying that the potential offsite gaseous and liquid release associated with a fire in each area meets the Technical Specification limits for offsite releases.

Attachment to L-2014-071 Page 70 of 101 Mrs. Jansen Vehec was a key member of the WSES-3 RAI response team for the PRA portion of the NFPA-805 LAR. In this capacity, she was responsible for performing and/or reviewing sensitivity analyses associated with replacing unapproved analysis methodologies (UAMs) with approved methodologies. These analyses included sensitivities associated with secondary combustibles, higher heat release rates, removal of unapproved reduction factors from screening analyses, and removal of other unapproved factors to determine the individual and cumulative impact of fire scenarios at WSES-3 using approved methods versus UAMs.

Mrs. Jansen Vehec was one of the leads for the development of the Brunswick Nuclear Plant (BNP) Control Room evacuation HRA developed to support the BNP fire PRA. In this capacity she was responsible for identification of critical actions, the development of the control room evacuation and critical action timelines, review of the assumptions used in the analysis, and review of the HRA calculator inputs for each of the critical actions.

Mrs. Jansen Vehec was a primary member of the Calvert Cliffs Nuclear Power Plant Fire PRA development team. In this role, Mrs. Jansen was the lead engineer for the development of the plant partitioning and qualitative screening analyses and notebooks.

Mrs. Jansen Vehec was also the lead engineer for the fire risk quantification efforts for both Units, including the initial set-up of the FRANX scenarios, quantification of all scenarios, and documentation of the process and results of the scenarios. Mrs. Jansen Vehec other responsibilities on this project included identifying potential Dual Unit initiators, assisting with the equipment selection task, identifying equipment relied upon by operators in the control room to ensure the equipment was included in the cable selection task, reviewing fire AOPs to identify critical operator actions, assisting in the review of the HRA calculator assessments of the operator actions being credited under fire scenarios, assisting in the review of the HRA dependency analysis, reviewing the resulting recovery action file used in quantification, performing cut set reviews, and assisting with detailed fire model development, including identification of target sets associated with individual ignition sources. Mrs.. Jansen Vehec was responsible for the resolution of the Peer Review findings and observations associated with topics for which she was the lead engineer.

Mrs. Jansen Vehec was the technical lead and project manager for the development of the Labgene Internal Events PRA. This project involved developing an integrated at-power PRA model for a PWR test reactor that is being designed to mimic the operational capabilities of a nuclear submarine. The project included the development and documentation of initiating events, accident sequences, system models, data analysis, human reliability analysis, ISLOCA analysis, common cause failure analysis, quantification, and cutest reviews. As technical lead, Mrs. Jansen Vehec was responsible for the identification and development of the initiating events of concern, including their accident sequence progression and success criteria, overall integration of the individual

Attachment to L-2014-071 Page 71 of 101 system models, quantification of the models, and review of the results. In this capacity, she also provided mentoring to younger engineers on systems modeling and fault tree development, data analysis, HRA analysis, and the breaking of circular logic, was the lead engineer responsible for reviewing each of these tasks.

Mrs. Jansen Vehec participated in the -development of a focused scope seismic evaluation for an unrestrained RCP and steam generator condition as Arkansas Nuclear One. Her responsibilities included supporting the development of an event tree and supporting fault tree models to evaluate the potential for fuel damage due to a seismic event during fuel movements coincident with the RCP and/or S/G being non-seismically restrained, supporting the evaluation of operator actions following the postulated seismic event using the SAROS toolbox, and quantification of the resulting model.

Mrs. Jansen Vehec was'the project manager and technical lead for the Kemkraftwerk Miihleburg (KKM) Phase 2 fire risk assessment project. This project involved more than ten people, and involved the development of a complete fire PRA model for the plant, including the walkdown and documentation of .the KKM Operations and*Reactor Building fire zones to document cable types, and locations, ignition sources, fire protection capabilities and equipment, physical layouts, and potential fire propagation pathways for each fire zone in the buildings that were not screened out in the Phase 1 analysis. Following the walkdowns, detailed evaluations of fifteen fire zones, including identification of target sets associated with individual ignition sources, were performed and results were documented. In this capacity, Mrs. Jansen Vehec was primarily responsible for the development of the fire scenarios, the documentation of cable locations and loading, the identification of ignition sources, the review of operating procedures, the identification of potential fire-specific HRAs, the evaluation of which Internal Events HRAs were impacted by fire conditions, oversight of the HRAs developed including a review of the HRA dependency evaluation for reasonableness and completeness, the quantification of the fire scenarios, and the review of cutsets to ensure the validity of the fire PRA and credited post-trip recovery actions..

Mrs. Jansen Vehec was responsible for the integration of the Phase 1 KKM fire risk assessment response logic into the KKM internal events PRA model, and the incorporation of airplane crash impacts into the KKM probabilistic risk assessment model. Mrs. Jansen Vehec was the project manager and lead analyst for the KKM probabilistic risk assessment data analysis update project and was also the project manager for the update of the at-power PRA model for KKM. In addition to overall project management responsibilities, she: was responsible for the methods development and application for each portion of the update, including initiating events identification, success criteria and accident sequence determination, system modeling, data analysis, HRA, and quantification.

Mrs. Jansen Vehec has been a memberof the team for several PRA Peer Reviews, Fire PRA Peer Reviews, and focused scope Peer Reviews. As a member of the various PRA Peer Review and focused Peer Review teams, Mrs. Jansen Vehec has served as the

Attachment to L-2014-071 Page 72 of 101 subject matter expert for the Internal. Flooding, the Human Reliability, the Systems, and the Accident Sequence subject areas; and as a supporting reviewer for the success criteria, data, quantification, LERF, and maintenance and update subject areas. As a member of the Fire Peer Review teams, Mrs. Jansen Vehec served as the subject matter expert for the Plant Response Model, Fire Risk Quantification, and Uncertainty subject areas; and as a support reviewer for the Plant Partitioning, Equipment Selection, Cable Selection, Ignition Source, and Fire Scenario Selection and Analysis, and Human Reliability Analysis subject areas. Mrs. Jansen Vehec was also a team member for the Koeberg PRA Peer Review in South Africa which included both a review of the at-power PRA model against the requirements of the ASME PRA Standard, and a review of the low power and shutdown model against the Draft ANS Low Power and Shutdown Requirements.

Mrs. Jansen Vehec was responsible for the development and update of the H.B. Robinson Nuclear Plant success criteria and accident sequence notebooks, including the development and implementation of the WCAP- 15831-P methodology for ATWS modeling. Other recent work includes updating the H.B. Robinson auxiliary feedwater notebook, accident sequence analysis notebook, event trees and fault trees to reflect revised Steam Generator Tube Rupture success criteria and assisting in the conversion of the Waterford-3 MAAP deck conversion to version 4.0.6.

February 2003 to February 2007, American Electric Power At the Donald C. Cook Nuclear Plant, Mrs,. Jansen Vehec had several roles within the PRA group and plant operations.

As an engineering supervisor, she passed the NRC Generic Fundamentals Examination, and all portions of the Site's Audit examination, which included the written examination, three simulator scenarios (one for each position in the Control Room - Unit Supervisor, Balance of Plant Reactor Operator, and At the Controls Reactor Operator), and fifteen Job Performance Measures (JPMs) (3 in-plant JPMs, 5 Administrative JPMs, and 7 Simulator JPMs) to obtain her SRO Certification. She was assigned to an operations crew to participate in re-qualification training and written examinations to maintain and broaden Operations background.

She assisted unit supervisors and work control SROs with coordinating and evaluating control room and work control center activities, pre-job briefings, adequacy of clearances, initiating and resolution of conditions adverse to quality, and determination of technical specifications and/or report ability impacts. Mrs. Jansen Vehec also monitored and reported on the status of Operations-related outage activities, performed Containment Closeout inspections, participated in Refueling activities, and assisted in preparation of Clearance packages associated with the outage.

As the probabilistic risk assessment engineering supervisor, her accomplishments included guidance, direction, and leadership to a multidiscipline staff to achieve department and station PRA goals. This included:

Attachment to L-2014-071 Page 73 of 101

" Leading a major PRA Upgrade Project which was completed within time and budget constraints

" The establishment of a culture that integrated PRA techniques into daily plant operations and was accountable for the successful operation of this focus area

" Providing advice and counsel to senior management regarding risk informed opportunities for the site As the Supervisor of the PRA Upgrade project, Mrs. Jansen Vehec was responsible for the quality and completeness of all portions of the PRA including initiating events, accident sequence and success criteria development, plant specific thermal hydraulics analysis, plant specific data analysis, human reliability analysis, LERF analysis, internal flooding analysis, and quantification. The resulting PRA was used in support of various NRC submittals, as well the model for the station's on-line risk monitoring program.

She also functioned as primary interface with the NRC on all risk-related regulatory interactions, evaluated industry operating experience for applicability to the site and the PRA group, supported Root Cause evaluations associated with plant transients, and supported License Amendment Requests (LARs) and resolution of the NRCs Requests for Additional Information (RAIs) associated with conversion to Improved Technical Specifications and Plant Life Extension.

While leading the PRA group, she prepared the risk assessments and associated modeling changes required to support an LAR for converting from a 3-day Emergency Diesel Generator (EDG) Limiting Condition for Operation (LCO) time to a 14-day EDG LCO.

She was also an active key member of the EPRI Risk/Safety Management Technical Advisory Committee, and the WOG Risk Management Subcommittee.

September 1999 to February 2003, TXI_

While at the Comanche Peak Steam Electric Station (CPSES), she served as the CPSES site lead and Strategic Teaming and Resource Sharing (STARS) Team representative responsible for development and coordination of Life Cycle Management (LCM) activities. She served as the STARS Project Manager for the Main Generator LCM effort which received an EPRI Nuclear Power Sector Technology Transfer award. She also ensured the CPSES needs were being met'by the STARS LCM effort, and for identified and implemented STARS LCM strategies at CPSES.

As the chairperson of the WOG RBTWG, she worked with Combustion Engineering Owner's Group (CEOG) chairperson to integrate the two groups into a single, cohesive Risk Management Subcommittee as pait of the WOG/CEOG merger. In this capacity, she was also responsible for the review and approval of all WOG Peer Reviews performed while she was chairperson.

Mrs. Jansen Vehec also served as the project manager and technical lead for the CPSES PRA Certification Course (designed to teach comprehensive PRA fundamentals). She provided oversight and technical review of the development of the material, the

Attachment to L-2014-071 Page 74 of 101 presentation of the training, and identified potential future enhancements for the course.

She managed the initial development and presentation of the 13-week course and served as contract coordinator, as well as subject matter expert for review of the materials. This course has been taught to members of several utilities in addition to TXU personnel, and includes training on fault tree and event tree development, generic and site-specific data development, model integration, breaking of circular logic, performing HRA evaluations using various industry HRA tools, quantification of the system level fault trees and the integrated fault trees, performance of cutest reviews, and potential applications. Additionally, she contributed to the STARS Risk Informed - In-service Inspection initiative which developed a common, risk-informed inspection program that enabled STARS plants to reduce the number of required examinations and reduced inspection costs and worker radiation exposure.

She performed PRA model maintenance, supported plant operations, and interfaced with NRC Resident Inspectors. She developed and documented an upgraded and detailed internal flooding probabilistic risk model, updated and upgraded system models, top logic models, and operator response models and the associated documentation in support of a PRA model upgrade effort, and performed several Phase 3 SDP analyses.

October 1998 to September 1999, ENERCON Services, Inc.

She served as an on-site staff augmentation PRA Engineer at the Clinton Power Station supporting restart of the station. Tasks performed during the assignment included:

" Preparation of the site's responses to the Nuclear Regulatory Commission's (NRC) Requests for Additional Information on the Individual Plant Examination for External Events ([PEEE) fire submittal

" Development of and validation of a plant specific shutdown risk model for use with the Outage Risk Assessment and Management (ORAM) program

" Preparation of and review of PRA related technical documents and analyses, represented PRA on the Maintenance Rule Expert Panel

" Performance of 50.59 evaluations of proposed plant design changes

" She also was the interface with NRC regulators on Appendix R and Fire Protection Program open issues and resolved NRC concerns associated with the programs is support of Plant restart efforts September 1997 to October 1998, Carolina Power and Light Company While working at the H.B. Robinson Nuclear Plant as the site PRA engineer, she was responsible for maintenance of the Site's PRA model; including model updates, operator interviews to ensure model accuracy and response timing, and training on the uses of the PRA in day-to-day plant operations. She also evaluated potential safety impacts associated with scheduled maintenance activities, proposed plant modifications and procedure changes, performed PRA living model maintenance, and represented PRA on the site's Maintenance Rule Expert Panel, represented PRA on the site's Severe Accident

Attachment to L-2014-071 Page 75 of 101 Mitigation Alternatives (SAMA) Team and developed several SAMA procedures and provided PRA training and SAMA training to all the departments at the site.

January 1997 to September 1997, Ricky Summitt Consulting (RSC), Inc.

While working for RSC, Inc., Mrs. Jansen Vehec was responsible for management and execution of tasks associated with safety, PSA and reliability projects for both commercial and governmental organizations.

She was responsible for the development of system fault tree models and associated documentation notebooks for the Miihleberg PSA upgrade being performed for BKW Energie in Switzerland. This involves developing system documentation for several BWR systems including the auxiliary cooling water and reactor building cooling water systems.

Another completed task performed by Mrs. Jansen Vehec was an update to PSA documentation for the Robinson plant service water system and component cooling water system as a part of the overall PSA update.. As part of this effort, the documentation and model were examined and updated as appropriate. *Theoverall documentation was streamlined to enhance future updates. i . ,

January 1990 - April 1997, Science Applications International Corporation, Los Alamos, New Mexico and Oak Ridge, Tennessee Mrs. Jansen Vehec, as an independent contractor to Science Applications International Corporation, supported the Special Initiators development activities for the. Vogtle Nuclear Power Plant. She was responsible for developing the special initiator models associated with the 120 VAC power, 125 VDC power, Nuclear Service Cooling Water, Instrument Air, and Alternate Core Cooling Water systems.

As an SAIC employee, Mrs. Jansen Vehec supported the Facility Safety and Operations Organizations at the East Tennessee Technology Park (ETTP) formerly known as the Oak Ridge K-25 Site. In this capacity, she was responsible for developing facility safety and authorization basis documentation, evaluating the operations and hazards associated with various types of facilities including laboratories, Environmental Restoration sites, and waste management. facilities, and.fo:r performing design reviews and safety evaluations, and for developing authorization basis documents for various demonstration operations and facilities proposed for use at the ETTP, including the Transportable Vitrification System and the Transportable Compressed Gas Recontainerization System.

She was also responsible for developing, reviewing and evaluating processes, procedures, work instructions, and Nuclear Criticality Safety Assessments for treating, storing or disposing of various materials including compressed gases, low-level waste, fissile materials, hazardous waste, TSCA-waste, RCRA-waste, and mixed wastes. Mrs. Jansen Vehec was responsible for helping coordinate the segregation, packaging, and transporting of high-enriched materials firom the ETTP to storage at the Y- 12 Site.

Attachment to L-2014-071 Page 76 of 101 As a nuclear/safety engineer, Mrs. Jansen Vehec was involved with the Site-wide Safety Analysis Report (SAR) for the Rocky Flats Environmental Technology Site. She was responsible for performing the bounding radiological and chemical hazard analysis for seismic events and range fires at the Site.

Mrs. Jansen Vehec was involved with upgrading the Savannah River Technology Center (SRTC) SAR to meet current Department of Energy Orders and Standards. Her involvement included facility hazards classifications, accident analysis, radiological hazard analysis, and chemical hazards analysis. Mrs. Jansen Vehec was also responsible for providing technical support for the development of the Basis for Interim Operation (BIO) for the SRTC. This effort involved providing detailed radiological source term and dose calculations and a comprehensive chemical hazard analysis.

Mrs. Jansen Vehec was involved with the preparation of the safety documentation required for the proposed 105-K Tritium Reservoir Storage Repository at the Savannah River Site. She was responsible for helping prepare the Hazard Assessment Document, Preliminary Hazards Analysis, and BIO for the proposed facility. Mrs. Jansen Vehec also supported the Savannah River Site Level 1 PRA for the K-Reactor. She was responsible for modifying fault trees to create a "living model" of the reactor.

Mrs. Jansen Vehec was responsible for helping perform a validation of the RAM program and related computer models developed and used by the Strategic Petroleum Reserve.

This effort included developing and reviewing RAM logic model structures, verifying the methodology employed, and helping establish and review an updated critical item equipment lists.

Mrs. Jansen Vehec was also involved with the development of a manual outlining the requirements associated with determining which components should be included in the RAM logic model structures, and the reasoning for each components inclusion or exclusion.

Mrs. Jansen Vehec served as an analyst for the Nuclear Regulatory Commission Accident Sequence Precursor Program, which uses nuclear plant operational experience data to assess risk-related performance. Mrs. Jansen Vehec was responsible for identifying, documenting, categorizing, and evaluating precursors to severe core damage sequences involving historic safety-related failures and operator errors at commercial nuclear reactors.

Mrs. Jansen Vehec was responsible for fault tree modeling, system notebook development, and/or review of various Shearon Harris Nuclear Power Plant, H. B.

Robinson Nuclear Plant, and Grand Gulf Unit 1 front-line systems and support systems.

Mrs. Jansen Vehec has also provided training and guidance to nuclear power plants employees in probabilistic risk assessment techniques, including training in the development of fault trees and supporting system notebooks.

Mrs. Jansen Vehec has also been responsible for analyzing the effectiveness of nuclear power plant emergency operating procedures, including evaluating and

Attachment to L-2014-071 Page 77 of 101 timing operator responses to various simulator scenarios to identify and examine potential human errors during accident sequences, and developing the estimated HRA values for the identified errors.

1989 - 1990, Research Assistant, Texas A&M University, College Station, Texas Mrs. Jansen Vehec was responsible for data coordination and analysis for a research project involving boron-li. She prepared reports detailing the use of boron-Il as a possible reflector and moderator substitute for beryllium and graphite in fission and fusion reactors.

1987 - 1989, Engineering Aide, TU Electric, Dallas, Texas Mrs. Jansen Vehec performed fault tree modeling of various systems, including auxiliary feedwater and essential AC electric power for the Comanche Peak Unit 1 Individual Plant Examination.

She was also involved with analyzing the sensitivity. of different thermal-hydraulic parameters within the VIPRE computer code and the effect of several axial and radial noding designs on the minimum departure from nucleate boiling ratio for the Comanche Peak reactor.

ADDITIONAL TRAINING AND PROFESSIONAL ACTIVITIES Senior Reactor Operator (SRO) Certification - D.C. Cook Nuclear Plant Plant Certification - Comanche Peak Inactive "Q" Clearance Applying Human Performance Error Reduction Tools, D.C. Cook, Bridgman, MI Hazard Evaluation: Quantitative Frequency Analysis Methods, Process Safety Institute in Knoxville, TN Engineering Error Prevention, Performance Improvement International, Hartsville, SC Seven week Leadership Academy, D.C. Cook, Bridgman, MI Advanced Project Manager Training, D.C. Cook, Bridgman, MI Two week Engineering Supervisory Skills Training course at INPO Headquarters in Atlanta, GA Supervisory Management Training Program at University of North Texas Biz, Bucks, & Best-in-Class course i:n Business Fundamentals at TXU in Dallas, TX Mrs. Jansen Vehec's publications are available upon request.

Attachment to L-2014-071 Page 78 of 101 ANDREA MAIOLI Senior Engineer, Risk Application and Methods I Westinghouse Nuclear Services Division Education:

MS Nuclear Engineer Politecnico di Milano, Italy PhD Nuclear Engineer Politecnico di Milano, Italy Summary:

Dr. Maioli is a Senior Engineer in the Risk Applications and Methods Group of Westinghouse's Nuclear Services Division. He has over 4 years of experience in Westinghouse and over 7 years in the nuclear safety area.

Initially at the Westinghouse Science and Technology Department (R&D), he worked on risk methods and applications for advanced reactors as support of the design phase, while developing his PhD thesis in risk-informed methodology for Off-site risk management.

Within the Westinghouse R&D organization he has been technical leader for PRA development of the Westinghouse IRIS advanced reactor PRA, including the development of the preliminary HRA for the plant. In collaboration with the ENEA Italian National Lab, he has been the technical lead for the application of seismic isolators on advanced plants to evaluate their potential reduction on the residual seismic risk.

He is currently supporting the PRA activities for the Westinghouse SMR.

In the Risk Applications and Methods group, Dr. Maioli contributed to a number of risk related projects including support for the Beaver Valley power uprate, Surry PRA model updates, Ringhals fire and flood risk assessment, Wolf Creek ISLOCA model update, AP1000 and EPP availability analysis and is currently supporting the Vogtle ESF/LOOP staggered test risk-informed program.

Dr. Maioli has been the technical leader for the Watts Bar and then the Comanche Peak Internal Flooding PRA (IF-PRA) updates, leading in both cases the entire effort from the initial plant walkdowns to the final IF-PRA model quantification; this included the HRA re-evaluation for the inclusion of flood-specific performance shaping factors. He participated and lead industry peer reviews on IF-PRA. He participated to the IF-PRA for the Palo Verde Plant and he is currently the technical lead for the AP1000 IF-PRA update.

Dr. Maioli is a subject matter expert in Seismic PRA (S-PRA). He supported the API1000 Seismic Margin Assessment (SMA) development and he is leading multiple seismic- related activities of the PWROG Risk Management Sub-Committee, being the technical lead for the PWROG RMSC External Events core team. He is currently the technical lead for the AP1000 S-PRA update.

Attachment to L-2014-071 Page 79 of 101 Dr. Maioli has been leading Westinghouse innovation activities aimed at the development of modeling tools for Seismic and internal flooding PRA activities and he has experience with the CAFTA R&R workstation code suite, including XINIT for external events modeling, RISKMAN, WinNUPRA and RiskSpectnim.

Dr. Maioli is a registered Professional Engineer in Italy. He is an adjunct faculty member at University of Pittsburgh, School of Engineering, where he teaches PRA modules of the graduate course on Nuclear Operation and Safety. He is a member of the ANS/ASME Joint Committee of Nuclear Risk Management, where he sits on the Standard Maintenance Subcommittee, chartered with the maintenance of the ASME/ANS PRA standard and in particular in the External Event writing group. He is also a member of the Advanced Non-Light Water reactor PRA standard working group. Dr. Maioli is also an ANS member and recently coordinated the seismic PRA sessions of the ANS PSA-1 I conference.

Dr. Maioli is author and co-author of more than 15 technical publications for conferences, journals and IAEA technical documents.

Attachment to L-2014-071 Page 80 of 101 Daniel Sadlon Engineer, Risk Applications and Methods - II Westinghouse Nuclear Services Division Education:

B.S. Mechanical Engineering University of Connecticut M. Eng. Mechanical Engineering Rensselaer Polytechnic Institute Summary:

Mr. Sadlon has been working in the Risk Applications and Methods group at Westinghouse Electric Company (WEC) four years and has experience with a wide range of Probabilistic Risk Assessment (PRA) tools and methods. Mr. Sadlon has been involved in a variety of projects related~to Level 1 PRA modeling and is currently the technical lead on the Fort Calhoun Station internal flooding PRA (IF-PRA), the Arizona Public Service company IF-PRA, and the McGuire IIF-PRA. Mr.-Sadlon also participated on the teams performing IF-PRAs for Watts Bar and Comanche Peak. Mr. Sadlon lead the development and incorporation of the internal flooding human reliability analysis (HRA), which included dependency analysis, isolation action development and performance shaping factor adjustments for all the previously mentioned IF PRAs.

Mr. Sadlon has been an important contributor to many projects to support the Fort Calhoun Station (FCS) PRA. Mr. Sadlon has been extensively involved with FCS co-sourcing efforts and has contributed to various activities, including system notebooks, CCFs and eRoom coordination. Sadlon has assisted in the development of new operator actions as well as evaluating existing operator actions for the FCS PRA. Additionally, Sadlon has assisted with data updates and appropriate data incorporation into the FCS PRA.

Mr. Sadlon has assisted with the Internal Flooding EPRI guidance document revision. He has presented lessons learned from performing the FCS IF-PRA at an EPRI Risk and Reliability users group meeting, as well as co-authoring and presenting an ASME paper on the same topic.

Mr. Sadlon has supported multiple EPUs in which his key areas of contributions focused on Success Criteria, Human Reliability, Margin Enhancement and Potential Modification evaluations.

Attachment to L-2014-071 Page 81 of 101 RAI PRA 27.e.01 In letter dated May 15, 2013, the licensee responded to PRA RAI 27.e and clarified that sources of model uncertainty and related assumptions associated with the internal events PRA were identified, documented, and characterized; however, the response does not indicate that these sources were reviewed to identify those relevant to the FPRA application. Using guidance provided in NUREG 1855, "Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision Making," assess the importance of candidate IEPRA sources of uncertainty identified as relevant to the FPRA, and include a discussion of the criteria utilized to assess their importance. For those determined to be key sources of uncertainty for the FPRA, evaluate their impact on FPRA CDF, LERF, ACDF and ALERF.

RESPONSE

A review of the Turkey Point Units 3 and 4 Uncertainty Analysis Notebook was performed. This analysis documents the uncertainty evaluation in a manner consistent with the guidance of NUREG 1855. Table 12-1 of the notebook summarizes, the significant uncertainties identified in this analysis. This table is reproduced below with a column added to far. right identifying the potential impact on the Fire PRA model. None of the items identified have a significant impact on the Fire PRA as discussed in the last column of the table below. A similar review was performed for the Assumptions and Sensitivity analyse5 performed in support of the Turkey Point Level 2 Internal Events PRA analysis in order to6 determine if these assumptions/sen'sitivity evaluations reflect a potential LERF model uficertainfy that could impact the Fire PRA. These items are documented in Table 27.e.01 - I below.

The overall impact of these internal events model uncertainties on the Fire PRA will be significantly less than the impact on the internal events model given the high probability of fire induced failures governing the Fire PRA quantification..Therefore, the impact of these uncertainties is considered a second order impact and will have an insignificant impact on the Fire PRA CDF, LERF, ACDF and ALERF.

Attachment to L-20 14-071 Page 82 of 101 Table Error! No text of specified style in document.-1: Issue Characterization for Turkey Point-Specific Sources of Model Uncertainty (QU-F4 and LE-F3)

Topic Discussion of Issue Part of Model Plant-Specific Assumptions Made Impact on Model Charcterization Eval with respect to (tomeet (to meet QLJ-EI)

QU-E1)  : Affected

.(to Approach Taken ( meet meet QU-E2) ( mee l(to E meet QU-E4) Assessment . ppotential Fire PRA Model 4 Impact

1. Initiating Events - Human-induced errors Support system Human-induced errors Human-induced errors Support system A small amount of non- Pre-initiator human error Human Induced Errors can be contributors to initiating event are not explicitly are assumed to be only a initiators do not conservatism may exist in the events have a very small for loss of support various support system sequences included in the support minor contributor to the include any results for the les modeled by contribution to the Fire PRA system initiating events initiating events system initiator fault support system initiating contribution from fault trees (Loss of Vital 125 given that they are not tree modeling event frequency and are human-induced VDC, Loss of 120V Vital impacted by a fire since they hence neglected failures. Instrument AC, Losses of occur prior to the fire.

4.16kV Buses Therefore, their impact on 3AI3BI3C/4AN4B/4C, Loss of the Fire PRA is expected to Instrument Air, Loss of Intake be very small, smaller than Cooling Water, and Loss of on the internal events model.

Component Cooling Water).

The impact on the base PRA should be small; however, any risk-informed applications that are sensitive to the impacts of support systems initiators may need to consider these human error impacts further.

2. Initiating Events - Initiating events Support system Common cause Electrical common cause The PRA does not This assumption adds a small The impact of common Common cause failures resulting from common initiating event failures are considered failures are assumed to consider the risk amount of non-conservatism cause failures of electrical of electrical panels and cause failures of sequences in the mechanical not be significant risk impacts of common to the base PRA results. The panels and buses on non-fire buses multiple trains of support system contributors, cause losses of probability of common cause initiating event frequencies support systems could initiator fault trees; electrical panels or electrical failures of panels will not impact the Fire PRA be significant risk however, common buses as initiating and buses is small. However, quantification. The impact of contributors cause failures of events, risk-informed applications in these common cause electrical support which electrical bus/panel failures is expected to have systems (e.g., losses losses may have a risk impact minimal impact on the Fire of buses and electrical may need to further evaluate PRA given that fire losses panels) are not the impact of common cause are the primary contributor to considered. failures. the fire risk.

Attachment to L-2014-071 Page 83 of 101 Table Error! No text of specified style in document.-I: Issue Characterization for Turkey Point-Specific Sources of Model Uncertainty (QU-F4 and LE-F3)

Topic Discussion of Issue Part of Model Plant.Slieciflc Assumptions Made " "mpaco" Model Characterization E With respect to Affected Approach Taken Assessment .potential Fire PRA Model (to meet QU-E1). (to meet QU-E2) . (to meet QU-E4) Im pact

3. Initiating Events - The frequency of an ISLOCA Accident Generic data is used, Generic data is assumed ISLOCA initiating The use of generic data in the No pipe ruptures are Use of Generic Pipe ISLOCA event is a key Sequences based on selection of to be applicable to the event frequencies PTN PSA is consistent with assumed in conjunction with Failure Frequencies to driver to the overall most representative specific ISLOCA are established industry practice. However, fire events, therefore, no Determine ISLOCA likelihood of an ISLOCA generic data that is scenarios that are based on the for ISLOCA, the probability of impact on fire PRA Frequency core damage event and available. evaluated, generic pipe rupture failure of piping given an quantification.

resulting containment data for the various overpressure is a key factor in bypass. A relatively scenarios that are the ISLOCA frequency and its small amount of data evaluated in the contribution to LERF (in exists concerning low model. particular). This assumption pressure piping failure may need to be reevaluated when exposed to higher for risk-informed applications RCS pressures. The in which ISLOCA impacts are use of conservative an important contributor to the failure estimates will risk-informed decision directly impact the CDF process.

contribution from ISLOCA.

4. Accident Sequence MAAP 4 modeling is Accident sequences MAAP is used to MAAP is assumed to MAAP is used to The use of MAAP is an The impact on the fire PRA Analysis/Success used for the PTN PRA. that rely on MAAP- evaluate plant appropriately model plant evaluate plant industry-accepted consensus is the same as that on the Criteria - Use of Although use of MAAP derived success response for all phenomena pertaining to response to various approach. Use of other T/H internal events model. Use Appropriate T/H Codes 4 is a consensus criteria accident scenarios for the accident sequences initiating events, codes might provide different of other codes is not approach, the NRC has which EPRI has it was used to analyze. results in some cases, expected to have a posed potential issues indicated the code is However, the overall results of significant impact on the Fire regarding the code's acceptable for use. other codes should be similar PRA.

appropriate application, to those predicted by MAAP.

5. Accident Sequence Adequate post-LOCA Large LOCA Design basis success It is assumed that 2 The design basis This assumption is expected Impact is limited to Large Analysis - Post-LOCA cooling is required to accident sequences criteria for post-LOCA HHSI pumps injecting to success criteria is to have only a small impact on LOCA accident sequences Recirculation Cooling prevent core damage. recirculation cooling two RCS injection lines is used, which may be the base case PRA results, which are not postulated in for Large LOCA However, the success are used. required. somewhat However, this assumption conjunction with a fire and criteria should not be conservative. may need to be re-considered cannot be caused by a fire.

excessively for risk-informed applications, conservative. especially those in which

- situations whereone HHSI pump is out of service are beingconsidered.

Attachment to L-2014-071 Page 84 of 101 Table Error! No text of specified style in document.-I: Issue Characterization for Turkey Point-Specific Sources of Model Uncertainty (QU-F4 and LE-F3)

TT c 4 Discussion of Issue Part of Model Affected Plant-Specific Approach Taken' Assumptions Made Impact on Model*' Charabtrlzation Assessment Eal with Fire potential respect to PRAModel (to meet QU-EI) (to meet QU-E2). (to meet QUE4) . Impact.: . .

6. Systems Analysis - In some systems, System fault tree In general, passive It is assumed that for Passive failures of This assumption adds a small No pipe ruptures are Ruptures and other failures of piping, tanks, modeling failures are not most systems that modeled systems amount of non-conservatism assumed in conjunction with passive failure modes or other passive included in most passive failures add a are generally not to the base PRA results. fire events, therefore, no components could be system fault tree negligible amount of considered. Risk-informed applications for impact on fire PRA non-minor contributors models, unreliability as compared specific systems may wish to quantification. Any passive to system unreliability, to the contributions from re-examine this assumption, failures assumed are active failures particularly if the system expected to have limited contains common piping that impact on the Fire PRA could disable multiple trains of given their low relative risk equipment. with respect to fire induced failures.

7. Systems Analysis - Normally-operating Accident sequences Most PTN system CCW assumes that the A CCW modeling This assumption is acceptable The current model includes System Operating systems that have that require CCW models include the pump is running, with considers only one for the base PRA. However, flags for a 1/3 split fraction Alignments additional standby success possibility of multiple pumps B & C in standby. operating CCW component importance associated with each pump trains need to consider alignments. However, alignment, measures will be impacted as running at the time of that various systems a single assumed a result (i.e., equivalent quantification. Therefore, operating alignments alignment is modeled components will have differing this issue is addressed in the exist that could result in for the CCW system. importance, etc.). As other current model.

different risk results. CCW pumps could be running during the plant's operating cycle, the impacts of this assumption on the results should be considered when specific risk-informed applications are performed.

8. Systems Analysis - Fouling of the intake Accident sequences Intake structure fouling The likelihood of intake ICW modeling does This assumption may Intake structure fouling is not ICW failure due to structure during an that credit operation is considered as an structure fouling during not consider intake introduce a small non- assumed to occur in intake structure fouling accident could disrupt of the ICW for initiator, but is not the post-accident structure fouling, conservatism into the PRA conjunction with a fire. Its the transfer to heat to decay heat removal considered during the mission time is negligibly results. The likelihood of a impact would be minimal if the ultimate heat sink. and cooling of plant post-accident mission small when compared to blockage event during post- considered given the low equipment. time for other initiating other ICW failure accident response is relatively probability when compare to events, models. small; however, such an event the probability of a fire could disable both system initiating event.

trains. This assumption may need to be considered further for risk-informed applications for which ICW operability is a key risk contributor.

Attachment to L-2014-071 Page 85 of 101 Table Error! No text of specified style in document.-1: Issue Characterization for Turkey Point-Specific Sources of Model Uncertainty (QU-F4 and LE-F3)

Topic Discussion of Issue Part of Model Plant-Specific Assumptions Made Impact on Model z; Characterization Eval with respect to (to meet QU,*E1) Affected Approach Taken (to meet QUE2)" (tomeet QUE4)! Assessment potential Fire PRA Model

QU'.EI) Impact

9. Systems Analysis - Operation of the PPC Accident sequences Failures of the HPSI It is assumed that PPC modeling This assumption, if The assumption is supported Backup Nitrogen Gas following loss of that credit PPC header valves to move sufficient backup assumes adequate determined to be inaccurate, by the basis for the Nitrogen Capacity for PPC instrument air would be operation when to their proper position nitrogen gas is available air is available from would introduce non- gas backup pressure basis dependent on having instrument air is are assumed to result to allow the PPC to the backup nitrogen conservatism in scenarios in specified in the Design Basis sufficient backup unavailable, in failure of HPSI flow. perform its functions in system to allow for which PORV operation is Document.

nitrogen supply. the event of a loss of system operation. needed and instrument air is instrument air. not available. Sensitivity studies may be appropriate to determine the impact of this assumption on the PRA results.

10. Human Reliability - Cognitive failures are All accident The higher cognitive It is assumed that Computed HEPs This approach may result in The impact of fire on HEPs is Cognitive Failure important contributors sequences involving failure probability selecting the higher of use the maximum an under or over estimation of addressed in accordance Probabilities to the overall HEPs. human errors obtained from the the calculated value computed by these errors. Alternative with the guidance of Ensuring an accurate HCR/ORE and cause- probabilities better either method for approaches summ.. ing together NUREG-1921 estimate of the error based models is used accounts for the the cognitive the HCR/ORE and CBDTM likelihood contributes to for each HEP. dominant processes. contribution, failure probabilities to reflect a more realistic PRA processes inherent in each risk result. approach (HCR/ORE focuses more on time-dependent concerns and CBDTM focuses more on process-related and decision-related concerns), or the use of other approaches to quantify cognitive errors (e.g., ASEP, SPAR-H).

Since alternative approaches for human -reliability analysis exist, PRA applications should include consideration of potential uncertainties that arise from the estimation of HEP values and the methodologies used.

Attachment to L-2014-071 Page 86 of 101 Table 27.e.01-1: Level 2 Internal Events PRA Model Assumptions and Sensitivity Analysis Review for Potential Model Uncertainty Impact on the Turkey Point Fire PRA Assumption/Sensitivity Discussion of Issue Eval. with respect to potential Fire Analysis PRA Model Impact 3.1 The steam generator ADV or SRV will stick open once it passes saturated Conservative assumption water and superheated steam with high-temperature fission products, thereby providing a direct path to the atmosphere. Such conditions exceed the design basis of the relief valve, so there is a high probability that the valve will no longer be able to close properly.

3.2 Core damage sequences that continue on the high pressure branch of the Conservative assumption Containment Event Tree (CET) are assumed to be at or near the primary PORV/SRV setpoint. Without a small LOCA, medium LOCA, large LOCA, or steam generator cooling, the primary cooling system will remain at high pressure. The PORVs/SRVs will be the only means to relieve energy from the system. Small-small LOCAs or other minor leakage are unable to remove sufficient energy to prevent the pressure from rising to the PORV/SRV setpoint.

3.3 For Level 2 sequences with no explicit containment failures, source term Realistic/slightly conservative calculations assume normal plant leakage to determine offsite assumption consequences. Fission products may still escape containment through normal plant leakage since the containment is not 100% leak tight.

3.4 The loop seal and core barrel are not expected to dear during high Realistic assumption pressure core damage sequences. The loop seals may clear through one of two mechanisms. A large seal LOCA has the potential to clear a loop seal, but MAAP calculations for Turkey Point show that the seal is not expected to clear for any size seal LOCA prior to the time of an induced tube rupture or hot leg failure. Secondly, the loop seals could clear if the operators "bump the pump" to inject additional water into the reactor vessel, but this action is warned against in the Turkey Point severe accident management guidelines.

3.5 No credit is given for recovery of offsite power after. core damage but Slightly conservative assumption before a radioactive release. The ability for offsite power recovery prior to core damage is addressed by the Level 1 PRA. Given that power recovery has not occurred prior to core damage, there is a small, but non-zero chance of power recovery in the period between core damage and radioactive release. This time window will vary for different scenarios, and therefore the slightly conservative assumption of no power recovery during this window is taken.

3.6 Because basemat meltthrough takes many days to erode the thick Realistic/slightly conservative basemat, if at all, containment failure calculations are terminated at 96 hrs. assumption A All Containment Isolation (CI) failures go to LERF Conservative assumption. Less than a 25% increase in total LERF.

B Pessimistic values for TI-SGTR Less than a 10% increase in total LERF. Since these types of failures cannot be induced by a fire, the impact on the Fire PRA will be significantly less than the impact on the internal events Level 2 quantification.

C Optimistic values for TI-SGTR Less than a 10% decrease in total LERF. Since these types of failures cannot be induced by a fire, the impact on the Fire PRA will be significantly less than the impact on the internal events Level 2 quantification.

Attachment to L-2014-071 Page 87 of 101 Table 27.e.01 -1: Level 2 Internal Events PRA Model Assumptions and Sensitivity Analysis Review for Potential Model Uncertainty Impact on the Turkey Point Fire PRA Assumption/Sensitivity Discussion of Issue Eval. with respect to potential Fire Analysis PRA Model Impact D1 Increase operator failure to depressurize RCS Results show no change or only changes associated with induced SGTR scenarios (< 8% increase) and early containment failures (<1%

increase). Since these types of failures cannot be induced by a fire, the impact on the Fire PRA will be significantly less than the impact on the internal events Level 2 quantification. Total LERF is increased by less than 5%.

D2 Decrease operator failure to depressurize RCS Less than 1% decrease in induced SGTR, all other Level 2 contributors see no measurable change -

Decrease operator failure to keep a ruptured SG full during SGTR Less than 1% decrease in SGTR, all other Level 2 contributors see no measurable change F Credit for anticipatory declaration of a general emergency during a slow Less than 1% decrease in induced SGTR scenario SGTR, all other Level 2 contributors see.no measurable change G Include 10% possibility that a loop seal clears due to a large seal LOCA or Less than a 15% increase in Total operator action and leads to a TI-SGTR LERF. Since these types of failures cannot be induced by a fire, the impact on the Fire PRA will be significantly less than the impact on the internal events Level 2 quantification.

Attachment to L-2014-071 Page 88 of 101 RAI PRA 28 The responses to Fire Modeling RAI 01 .e from letter dated March 18, 2013 (ADAMS Accession No. ML13099A441) and Fire Modeling RAI 06 from letter dated April 16, 2013 (ADAMS Accession No. ML13109A008) state that the sensitivity analysis provided in Appendix B of the MCR abandonment calculation will be updated to demonstrate that modeling assumptions and parameter selections are either conservative or do not have a significant effect on the overall probability of abandonment. The responses further note that the baseline parameter selections used to support the FPRA MCR abandonment calculation will be revised if the parameters are shown to have a significant non-conservative effect on the probability of abandonment. Describe the results of this revised MCR sensitivity analysis, and if baseline parameter selections have changed as a result, provide updated risk results as part of the aggregate change-in-risk analysis requested in PRA RAI 29.

RESPONSE

The time to MCR abandonment from the updated calculation (based on revisions resulting from RAIs FMOD 01.e (issued via letter dated March 18, 2103, Letter Number L-2013-086) and FMOD-06 (issued via letter dated April 16, 2013, Letter Number L-2013-107) will be incorporated into the post-second-round-RAI quantification.

Based upon a clarification call with the NRC on February 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initial review and feedback on this response.

RAI PRA 29 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. RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," identifies NUREG/CR-6850, NEI 04-02, Revision 2, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," and the ongoing FAQ process as documenting acceptable to the staff for adopting a fire protection program consistent with NFPA 805.

The NRC staff identified several methods and weaknesses that were 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 complete (e.g., credit for control power transformers is not supported by experiments).

Unacceptable methods and weaknesses:

" PRA RAI 0l.k regarding the use of a minimum joint HEP probability below 1.OE-05

" PRA RAI 01.o regarding the removal of screening criteria in the multi-compartment analysis

Attachment to L-2014-071 Page 89 of 101

" PRA RAI 0 i.p regarding the delay to damage or ignition of targets impact by HEAF scenarios

" PRA RAIs 01.r and 08 regarding the removal of the FAQ 08-0050, "Manual Non-Suppression Probability," credit for incipient detection in the MCR

" PRA RAI 01 .t regarding the elimination of conditional probabilities for propagation of fire from electrical cabinets (i.e., panel factors)

" PRA RAI 011.u regarding removal of credit given to Thermolag for preventing cable damage in HEAF scenarios

" PRA RAI 01 .x and Fire Modeling RAI 04 regarding resolution to limitations in the application of the Generic Fire Modeling Treatments

" PRA RAI 01.z.i (as clarified in RAI 01.z.i-O1) regarding excluding the risk impact of transient scenarios that would impact only one piece of equipment

" PRA RAI 08 (as clarified in RAI 08-01) regarding the cable loading of MCR panels, the availability of the MCR HVAC control system, and the use of.NSP values less than 0.001 in the MCR abandonment analysis

• PRA RAI 12 regarding elimination of CPT credit

" PRA RAI 13 (as clarified in RAI 13-02) regarding use of non-l.0 failure probabilities for fire-affected VFDR components

" The following Fire Modeling RAIs appear to have caused changes that may impact the fire-affected components for a variety of fires. The aggregate change in risk evaluation should include the potential impact of changes in:

o Fire Modeling RAI O1 .c regarding growth times for transient fires in the MCR o Fire Modeling RAI Ol.f regarding MCR panel vent locations o Fire Modeling RAI 0l.g regarding secondary combustibles in the MCR o Fire Modeling RAI 01 .k regarding evaluation of impact of fire propagation o Fire Modeling RAI O1.p regarding wall and comer effects on panel fires o Fire Modeling RAI 01 .q regarding the growth rate of transient fires involving secondary combustibles o Fire Modeling RAI O1..u regarding comer and wall effects on cable spreading room scenarios The following methods and weaknesses have been identified, but the NRC Staff review is continuing with additional RAIs and further supporting information has been requested.

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

Methods and weaknesses still under review:

Attachment to L-2014-071 Page 90 of 101

" PRA RAIs 01.a (as clarified by PRA RAI 01.a-01) regarding revisions to transient influencing factors to address FAQ 12-0064 guidance

" PRA RAI 01.i-01 regarding the use of circuit failure probabilities not substantiated by circuit analysis

" PRA RAI 01 .j (as clarified by PRA RAI 01 .j-01) regarding providing mean values reflecting propagation of parametric uncertainty, accounting for the state-of-knowledge correlation

" PRA RAI 01.1-01 (a) regarding HEP screening and/or seeping values

" PRA RAI 01 .m-01 regarding assumptions that may produce non-conservative ACDF and ALERF results

" PRA RAI 01.m-02 regarding dependencies between multiplier-adjusted and screening-based HEPs

" PRA RAI 01.q-01 regarding use of fire protection system reliabilities and availabilities representative of plant-specific operating experience

" PRA RAI 0l.r-01 regarding the application of the Bin 4 MCB frequency

" PRA RAI 01 .r-02 regarding (a) removal of the 0.19 NSP credit for those panels that are not fully enclosed; (b) detector availability and reliability; (c) propagation between electrical (non-MCB) panels within the MCR; and (d) removal of credit given to smoke detectors to limit the extent of internal fire damage within electrical (non-MCB) panels they monitor

" PRA RAI 01 .t-01 regarding replacement of panel factors with a new fire scenario development methodology instead of acceptable methods

" PRA RAI 01 .v-01 regarding time to actuation of automatic suppression systems

" PRA RAI 01.y-01 and Fire Modeling RAI 01 .j regarding the impact of secondary combustibles on the zone of influence of ignition sources

" PRA RAI 01.z.i-01 regarding including the risk impact of transient scenarios that would impact only one piece of equipment

" PRA RAI 01.z.ii-0l regarding placement of transient fires behind all open-back MCR panels

" PRA RAI 01.dd-01 regarding hot gas layer impacts on ZOIs

" PRA RAI 07-01 regarding counting and treatment of Bin 15 electrical cabinets

" PRA RAI 11-01 regarding CCDPs and compliant plant risk associated with MCR abandonment

" PRA RAI 13-01 regarding change in risk calculations

" PRA RAI 28 regarding changes to baseline parameter selections in the MCR abandonment calculation

Attachment to L-2014-071 Page 91 of 101

a. For each method (i.e., each bullet) above, 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. In addition, provide confidence (e.g., with a proposed implementation item) 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. Note that continued use of unacceptable methods may prohibit the staff from completing its review for self-approval.

RESPONSE

a. The following Table addresses each of the issues identified above with respect to part a of the above question.

Attachment to L-2014-071 Page 92 of 101 Issue Final Composite Analysis Disposition Post Transition PRA Disposition Unacceptable methods and weaknesses:

- PRA RAI 01 .k regarding the use of a Sensitivity analysis (to be documented with Credit for joint HEPs below 1.OE-05 will be minimum 05 (May 15,joint HEPL-2013-164) 2013, probability below 1.0E- post 2nd round RAI quantification) retained with adequate justification per the PRA Standard

- PRA RAI 01.o regarding the removal of Eliminated screening criteria from analysis. The post transition fire PRA model will be the screening criteria in the multicompartment All MCA scenarios will be quantified. final composite analysis model with all required analysis (May 15, 2013, L-2013-164) changes/updates as committed in Attachment S.

• PRA RAI 01.p regarding the delay to No delay to damage or ignition of targets for The post transition fire PRA model will be the damage or ignition of targets impact by HEAF scenarios will be credited, final composite analysis model with all required HEAF scenarios (May 15, 2013, L-2013-164) changes/updates as committed in Attachment S

- PRA RAis 01 .r and 08 regarding the No credit is taken for detection in the MCR. The post transition fire PRA model will be the removal of the FAQ 08-0050, "Manual Non-' See PRA RAI 01 .r.01 response (revised Att. final composite analysis model with all required Suppression Probability," credit for incipient S provided to eliminate detection changes/updates as committed in Attachment S detection in the MCR (May 15, 2013, L- modification, provided with response to RAI 2013-164) PRA 01 .r.02)

- PRA RAI 01 .t regarding the elimination of Credit for electrical cabinet panel factors have The post transition fire PRA model will be the conditional probabilities for.propagation of been removed from the Fire PRA. final composite analysis model with all required fire from electrical cabinets (i.e., panel changes/updates as committed in Attachment S factors) (May 15, 2013, L-2013-164)

- PRA RAI 01 .u regarding removal of credit Credit to be retained in accordance with the The post transition fire PRA model will be the given to Thermolag for preventing cable response submitted for RAI PRA 01 .u. final composite analysis model with all required damage in HEAF scenarios (April 16, 2013, changes/updates as committed in Attachment S L-2013-107)

- PRA RAI 01 .x and Fire Modeling RAI 04 Analysis to be updated in conformance with The post transition fire PRA model will be the regarding resolution to limitations in the the response to PRA RAI 01 .x and FM 04. final composite analysis model with all required application of the Generic Fire Modeling changes/updates as committed in Attachment S Treatments (April 16, 2013, L-2013-107)

- PRA RAI 01 .z.i (as clarified in RAI 01 .z.i- See clarification provided in PRA RAI The post transition fire PRA model will be the

01) regarding excluding the risk impact of 01.z.i.01 response. Did not postulate final composite analysis model with all required

• transient scenarios that would impact only transient at each fixed ignition source if that changes/updates as committed in Attachment S one piece of equipment ignition source was the only component impacted by the transient scenario.

Attachment to L-2014-071 Page 93 of 101 Issue Final Composite Analysis Disposition Post Transition PRA Disposition

- PRA RAI 08 (as clarified in RAI 08-01) MCR analysis revision is based on multiple The post transition fire PRA model will be the regarding the cable loading of MCR panels, cable bundle (unless confirmed as single final composite analysis model with all required the availability of the MCR HVAC control cable bundle via walkdown); HVAC is changes/updates as committed in Attachment S system, and the use of NSP values less than assumed to be in the most conservative 0.001 in the MCR abandonment analysis configuration (unless confirmation of HVAC availability is confirmed by review of functions impacted on a panel basis). NSP less than 0.001 only utilized ifthe HRR bin cannot cause abandonment at any time.

PRA RAI 12 regarding elimination of CPT Credit for the reduction in spurious operation The post transition fire PRA model will be the credit (May 15, 2013, L-2013-164) probability from the CPTs will be removed, final composite analysis model with all required Interim Technical Guidance values will be changes/updates as committed in Attachment S used. The model may also be updated with the latest guidance contained in NUREG/CR 7150 Volume 2 when published PRA RAI 13 (as clarified in RAI 13-02) See PRA RAI 13.02 response clarification. The post transition fire PRA model will be the regarding use of non-1.0 failure probabilities Fire impact is assumed to result in a i.0 final composite analysis model with all required for fire-affected VFDR. components failure probability of impacted, cables unless a changes/updates as committed in Attachment S.

hot short probability is applied.

The following Fire Modeling RAIs appear to have caused changes that may impact the fire-affected components for a variety of fires. The aggregate change in risk evaluation should include the potential impact of changes in:

Fire Modeling RAI 01.c regarding growth Resolution incorporated. Transient fire HRR The post transition fire PRA model will be the times for transient fires in the MCR (March vs time is modeled per NUREG/CR-6850, final composite analysis model with all required 18, 20.13, L-2013-086) Supp. 1. changes/updates as committed in Attachment S Fire Modeling RAI 01.f regarding MCR panel Resolution incorporated. Panel fires with wall The post transition fire PRA model will be the vent locations (March 18, 2013, L-2013-086) and corner impact were added to MCR final composite analysis model with all required abandonment calculation and PRA changes/updates as committed in Attachment S incorporated the bounding evaluation.

Attachment to L-2014-071 Page 94 of 101 Issue Final Composite Analysis Disposition Post Transition PRA Disposition Fire Modeling RAI 01.g regarding secondary The updated MCR calculation includes a two The post transition fire PRA model will be the combustibles in the MCR (March 18, 2013, cable tray stack in combination with a multiple final composite analysis model with all required L-2013-086) cable bundle panel fire as a baseline fire changes/updates as committed in Attachment S scenario. This scenario was incorporated into the Fire PRA abandonment frequency evaluation.

Fire Modeling RAI 01.k regarding evaluation Resolution incorporated. New ZOI and hot The post transition fire PRA model will be the of impact of fire propagation (May 15, 2013, gas layer time analysis is used in HGL final composite analysis model with all required L-2013-164) evaluation. changes/updates as committed in Attachment S Fire Modeling RAI 01.p regarding wall and Resolution incorporated. Wall/corner effects The post transition fire PRA model will be the corner effects on panel fires (April 16, 2013, are applied based on validation of location of final composite analysis model with all required L-2013-107) ignition sources with respect to walls/corner. changes/updates as committed in Attachment S Fire Modeling RAI 01.q regarding the growth Resolution incorporated. Transient specific The post transition fire PRA model will be the rate of transient fires involving secondary secondary combustible ZOI and HGL timing final composite analysis model with all required combustibles (April 16, 2013, L-2013-107) analysis have been performed. changes/updates as committed in Attachment S Fire Modeiing RAI 01.u regarding corner and Resolution incorporated. , Analysis to The post transition fire PRA model will be the wall effects on cable'spreading room address wall/corner effects for the specific final composite analysis model with all required scenarios (April 16, 2013, L-2013-107) cable spreading room scenarios have been changes/updates as committed in Attachment S updated.

Methods and weaknesses still under review:

- PRA RAIs 01 .a (as clarified by PRA RAI Transient frequencies in accordance with The post transition fire PRA model will be the Ol.a-01) regarding revisions to transient FAQ 12-0064 will be included in the Fire final composite analysis model with all required influencing factors to address FAQ 12-0064 PRA. changes/updates as committed in Attachment S guidance

• PRA RAI 01 .i-01 regarding the use of Incorporated circuit analysis basis for all The post transition fire PRA model will be the circuit failure probabilities not substantiated circuits crediting hot short probabilities, final composite analysis model with all required by circuit analysis changes/updates as committed in Attachment S

- PRA RAI 01 .j (as clarified by PRA RAI 01 .j- Confirmation of uncertainty analysis mean Point value will continued to be used based on

01) regarding providing mean values value consistent with the point value used. the point value being consistent with the mean reflecting propagation of parametric value uncertainty, accounting for the state of knowledge correlation

Attachment to L-2014-071 Page 95 of 101 Issue Final Composite Analysis Disposition Post Transition PRA Disposition

- PRA RAI 01.1-01 (a) regarding HEP Updated HRA to eliminate the use of The post transition fire PRA model will be the screening and/or scoping values scoping/screening HEPs. final composite analysis model with all required changes/updates as committed in Attachment S

- PRA RAI 01 .m-01 regarding assumptions HRA no longer uses scoping/screening The post transition fire PRA model will be the that may produce non-conservative ACDF HEPs. Updated HRA will use NUREG-1921 final composite analysis model with all required and ALERF results approaches and HRA calculator. changes/updates as committed in Attachment S

- PRA RAI 01 .m-02 regarding dependencies HRA no longer uses scoping/screening The post transition fire PRA model will be the between multiplier-adjusted and screening- HEPs. Updated HRA uses NUREG-1921 final composite analysis model with all required based HEPs approaches and HRA calculator. changes/updates as committed in Attachment S

- PRA RAI 01 .q-01 regarding use of fire Evaluation performed on all readily available The monitoring program will track protection system reliabilities and data. reliability/availability of risk significant SSCs availabilities representative of plant-specific operating experience

• PRA RAI 01 .r-01 regarding the application MCB NUREG/CR-6850, Appendix L analysis The post transition fire PRA model will be the of the Bin 4 MCB frequency was updated to reflect total Bin 4 frequency. final composite analysis model with all required changes/updates as committed in Attachment S.

• PRA RAI 01 .r-02 regarding (a) removal of Credit for in-panel detection in main control The post transition fire PRA model will be the the 0.19 NSP credit for those panels that are room is eliminated, final composite analysis model with all required not fully enclosed; (b) detector availability changes/updates as committed in Attachment S and reliability; (c) propagation between electrical (non-MCB) panels within the MCR; and (d) removal of credit given to smoke detectors to limit the extent of internal fire damage within electrical (non-MCB) panels they monitor

- PRA RAI 01 .t-01 regarding replacement of Eliminated panel factors and incorporated The post transition fire PRA model will be the panel factors with a new fire scenario use of NUREG/CR-6850, Appendix H. final'composite analysis model with all required development methodology instead of changes/updates as committed in Attachment S acceptable methods

- PRA RAI 01 .v-01 regarding time to Automatic suppression system actuation is The post transition fire PRA model will be the actuation of automatic suppression systems based on method of actuation and final composite analysis model with all required comparison to associated hot gas layer- changes/updates as committed in Attachment S scenario credit.

Attachment to L-2014-071 Page 96 of 101 Issue Final Composite Analysis Disposition Post Transition PRA Disposition

- PRA RAI 01.y-01 and Fire Modeling RAI Incorporated new ZOI where cable tray The post transition fire PRA model will be the 01 .j regarding the impact of secondary secondary combustibles are present and final composite analysis model with all required combustibles on the zone of influence of using updated time to Hot Gas Layer analysis changes/updates as committed in Attachment S ignition sources which incorporates secondary combustibles.

- PRA RAI 01 .z.i-01 regarding including the See clarification provided in PRA RAI The post transition fire PRA model will be the risk impact of transient scenarios that would 01 .z.i.01 response. Did not postulate final composite analysis model with all required impact only one piece of equipment transient at each fixed ignition source if that changes/updates as committed in Attachment S ignition source was the only component impacted by the transient scenario.

- PRA RAI 01 .z.ii-01 regarding placement of Incorporated transient scenarios behind all The post transition fire PRA model will be the transient fires behind all open-back MCR open .MCR back panels. final composite analysis model with all required panels changes/updates as committed in Attachment S.

- PRA RAI 01 dd-01 regarding hot gas layer Clarified approach for incorporation of HGL The post transition fire PRA model will be the impacts on ZOls impact. final composite analysis model with all required changes/updates as committed in Attachment S

- PRA RAI 07-01 regarding counting and Incorporated guidance for exclusion of The post transition fire PRA model will be the treatment of Bin 15 electrical cabinets counting of sealed panels. Sealed 480V final composite analysis model with all required MCCs were not analyzed as propagating changes/updates as committed in Attachment S fires.

- PRA RAI 11-01 regarding CCDPs. and Clarified MCR CCDP analysis approach. The compliant CCDP and risk is only associated compliant plant risk associated with MCR with the final composite analysis model and will abandonment not be retained in the post transition model.

- PRA RAI 13-01 regarding change in risk Clarified methodology for delta risk Change in risk calculations will be used to calculations calculations. support change evaluations and will be done in accordance with procedure EN-AA-202-1004 which conforms with FAQ 12-0061

- PRA RAI 28 regarding changes to baseline Updated MCR abandonment frequency The post transition fire PRA model will be the parameter selections in the MCR based on new abandonment time analysis. final composite analysis model with all required abandonment calculation changes/updates as committed in Attachment S RAI 29 (Seal Modification) The post transition fire PRA model will be the final composite analysis model with all required changes/updates as committed in Attachment S

Attachmnent to L-2014-071 Page 97 of 101 An implementation item will be added to LAR Attachment S, Table S-3, Implementation Items, to state that a review of all changes to the PRA model 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.

A markup of Attachment S, Table S-3 is provided with this response RAI PRA 29.b

b. Describe any new modifications or operator actions with respect to. those identified in the original LAR (e.g., the FlowServe reactor coolant pump seal package), and summarize how these modifications have been modeled in the PRA, how they will be included in the final change in risk estimates in support of the LAR, and if an implementation item confirming the as-built change in risk estimates will also include conifirmation of these models. Indicate the expected reduction in risk from the modifications. Discuss the associated impacts to the fire protection program.

RESPONSE

b. The modification to incorporate the Flo.wServe RCP seal package will be incorporated into the updated LAR quantification. The following changes to the fault tree and recovery file will be implemented in the model to provide a bounding quantification of the performance of the FlowServe seals:

Probability of a seal LOCA when pumps are tripped is significantly reduced (was a total of 0.21 and is now reduced to 8.99E-06 with the new seals.

Human Error Probability for an operator action to trip the RCPs to reduce the probability of a seal LOCA is reduced from 4.00E-02 (with a time,window of 13 minutes) to 2.90E-04 (with a time window of 60 minutes).

An implementation item associated with the installation of the FlowServe seals will be added to LAR Attachment S, Table S-2, Plant Modifications Committed. This item will include confirmation of the above logic against the NRC approved FlowServe Topical Report. The expected reduction in risk is expected. to offset the increase in risk due to the resolution of various RAIs. An Attachment S modification to protect RCP related cables in the cable spreading room, will be removed given the incorporation of the new seal design. A markup of Attachment S to address these changes to the LAR is provided with this RAI response.

The resulting total risk/delta risk will beconfirmed to be within the limits specified in RG 1.174. Other changes to NFPA 805 recovery actions are expected due to the model change.

Based upon a clarification call with the NRC onFebruary 5, 2014, it was agreed that the response to the 2 nd round RAIs would be submitted without risk numbers reflecting Fire PRA requantification. The updated risk numbers that incorporate changes associated with this RAI response will be provided following NRC initialreview and feedback on this response.

Attachment to L-2014-071 Page 98 of 101 Security-Related Information -Withhold Under 10 CFR 2.390 Florida Power & Light Attachment S - Plant Modifications and Items to be Completed Table S-2 Plant Modifications Committed Item Ranke Unit Problem Statement Proposed Modification In Camp Risk-Informed Characterization FPRA Measure 22 Low. 3, 4 Plant fire detection system has Upgrade the fire detection system Yes Yes Ensures detection available where numerous minor deviations from the per the Code of Record. required Code of Record Compensatory Measure:

Compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in place until this modification is complete.

23 Medium 3, 4 Thermal detector is required for early Linear thermal detecting wire in Yes Yes Provides early detection of transient detection and suppression of cable tray risers in Fire Area CC fire for earlier suppression, transient fire in fire area CC. placed low to the ground (risers preventing fire spread to overhead 3JCT10, 3JDT10, 3JETIO, 3JFT10, Cable trays. Provides additional risk 3JGT1O, 3JHTIO, 3JJT10, 4JDT10, benefits to both COF and. LERF.

4JET10, 4JFT10, 4JGT10, 4JH.T10, Compensatory Measure:

and 4JJT10). Compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in place until this modification is complete.

24 High 3, 4 Credit protection of cables Protect cables in Fire Area CC and Yes Yes Allows the RCPs to be tripped from associated with the'following HH associated with RCPs to ensure the MCR to prevent a seal LOCA.

breakers to allow tripping of all three control room trip capabilities of all Provides additional risk benefits for RCPs for both units from the control RCPs: CDF and LERF.

room: 3AAO1-FTO, 3AB01-FTO, 3AAO1/3AA0113C03io02 Compensatory Measure:

3A806-FTO, 4AA01-FTO, 4AB01- 3AB01/3AB01/3C03f002 Compensatory measures will be FTO and 4AB06-FTO. 3AB06/3AB0613C031002 established when the NFPA 805 fire 4AAO1/4AA01/4C031002 protection program becomes 4AAO1/4AB06/4B071001 effective and remain in place until 4AI01/4AB01/4C03/002 this modification is complete.

4AB06/4Ao06/4C03/002 IDelete Revision 0 Page S-15

Attachment to L-2014-071 Page 99 of 101 Security-Related Information - Withhold Under 10 CFR 2.390 Florida Power & Light Attachment S - Plant Modifications and Items to be Completed Table S-2 Plant Modifications Committed In Comp Risk-Informed Characterization Item Ranks Unit Problem Statement Proposed Modification FPRA Measure 31 Medium 4 Fire induced fault on the RWST to Fire Area R - Reroute cable Y Y Reroute RWST to charging header charging header supply valve control 4V1 15B14C264/4C03/00H out of Fire supply valve control cable to prevent cable can isolate all sources of Area R. loss of make-up water to charging make-up water to the charging pumps and potential subsequent pumps which could result in a loss of RCP seal LOCA.

RCP seal cooling and subsequent Compensatory Measures seal LOCA. Compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in place until this modification is complete.

32 Low 3, 4 Additional defense-in-depth detection Fire Areas U, V,W, and X - Add N N No credit for area-wide incipient inthe 4kV Vital switchgear rooms is area-wide incipient detection to 4kV detection taken in NSCA or FPRA.

required. Vital S ;;itchear Rooms. This modification is being done for defense-in-depth only.

New High 3, 4 High contribution from fire Replace RCP seals with Y Y Replacement of seals reduces induced loss of RCP seal FlowServe seals risk contribution of RCP seal cooling LOCA scenarios C..mpensatory Measure:

Compensatory measures will be established when the NFPA 805 fire protection program becomes effective and remain in effect until this modification is complete.

Page S-18 Revision 0 Page S-1 8

Attachment to L-2014-071 Page 100 of 101 Security-Related Information -Withhold Under 10 CFR 2.390 Florida Power & Light Attachment S - Plant Modifications and Items to be Completed Table S-3 Implementation Items Item Unit Description LAR Section I Source 15 3,4 Implemen UJpdat the Fire PR: A ,M, de...aJ _ate all" -- "" ti--" a..z

. " ..... -"-- an are mnd Attachment E

. I. rz pkte and as built and all implmzmntatien items affe.ting . the fire PRA results eemplett. Review the c;Juts of the fir PRtA eoniparcd to the~ final updated veso

..... been....d ... d .. .

A.s .have r.......ll .. . A.. Revised by RA I greatcr than 1E 07yr LER g tr th... 1F. 08/Y.hl generat . a

• ..t....

.... o to dete . e the caue the rkdifthat the PRA 16.

..n.lusi.n in the [AR j Superseded by RAI PRA 1601 Add an appendix that identifies where run-off has the potential to route to a storm drain

. Modify the fire brigade training program to include enhanced radioactive release objectives

. Develop a standard operating procedure to support actions to mitigate a radioactive release

• Develop administrative controls to support compliance with NFPA 805.

Radioactive Release Criteria

• Stage materials and equipment to assist in preventing potentially contaminated run-oft from entering the storm drain system 16 3.4 UpdatE fire protection program documents and provide training as necessary- 4.7 This in plementation item is to address general program documents not already Revised addres ed in items 1-15 and 17 This includes tire protection design basis i docum nt, post-transition change process (including Fire PRA updates), and PPRA 16.01 qualific lion training.

17 3.4 Updatl plant procedures based upon detemlination of required fire protection 4.8 and Table C-2 syste ms and features. I 18 3, 4 gpdat.*hte Fi.r PR* Model, as necessary,. after all ,mod.firal.onsare comp ete 4.8.2 antd as-buill \

19 3, 4 Create response procedures for the incipient detection systems. Attachment S. Table S-2 Items 3, 25, and 32.

Update the Fire PRA Model after all moctifications and procedural changes are complete and as-built and all e nfinal iSt omplementation items affecting the fire PRA results are complete. Review the results of the fire PRA compared to the updated version in the LAR* after all RAIs have been responded and accepted. This will be treated as a change aditiona terns evaluation and evaluated in accordance with procedure EN-AA-202- 1004 Revision 0 Page S-21

Attachment to L-2014-071 Page 101 of 101 Security-Related Information - Withhold Under 10 CFR 2.390 Florida Power & Light Attachment S - Plant Modifications and Items to be Completed Table S-3 Implementation Items Item Unit Description LAR Section I Source New 3, 4 Update the PTN Transient Combustible and Flammable Substance Program Section 4.1.2; Attachment A, administrative procedure to include all NFPA 805 Power Block structures/areas in 3.3.1.2(6) - Added by RAI the Transient Combustible Area Boundary Map 02.c.01 (FPL Letter L-2014-003)

New 3, 4 Revise procedures for Control of Transient Combustibles, and Control of Ignition PRA RAI 18.01 Sources, to restrict storage of transient combustibles/ignition sources in specific locations in the following fire areas/zones in accordance with the credit taken in the Fire PRA:

Fire Zone 058/Fire Area F - Unit 3 and 4 Auxiliary Building 18' Elevation Hallway Fire Zone 079A/Fire Area CC - Unit 3 and 4 Aux Building North-South Breezeway Fire Zone 098/Fire Area HH - Unit 3 and 4 Cable Spreading Room New 3, 4 Confirm consistency between the PRA model logic used for modeling the PRA RAI 29.b FlowServe seals with NRC approved Topical Report for these seals when the Topical Report is approved New 3, 4 A review of changes will be performed to determine if a focused scope peer PRA RAI 29.a review will be required. A focused scope peer review is required if changes in the Fire PRA are identified which constitute an upgrade as defined in the ANS/ASME Standard. Any findings from a focused scope peer review, should one be required, will be resolved prior to self-approval of post-transition changes Page S-New Revision 00 Revision Page S-New