L-2014-056, Response to Request for Additional Information License Amendment Request for Transition to 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 for Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

ML14070A097
Person / Time
Site:	Saint Lucie
Issue date:	02/24/2014
From:	Jensen J Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2014-056
Download: ML14070A097 (77)

Text

0February 24, 2014 FPL.

L-2014-056 10 CFR 50.90 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Re:

St. Lucie Units 1 and 2 Docket Nos. 50-335 and 50-389 Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

References:

1. FPL Letter L-2013-099 dated March 22, 2013, Transition to 10 CFR 50.48(c) -NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)..

2. Email from Siva Lingam, NRC, to Ken Frehafer, FPL, dated June 7, 2013, St. Lucie NFPA-805 LAR Acceptance Review Clarification Questions.

3. FPL Letter L-2013-193 dated June 14, 2013 Transition to 10 CFR 50.48(c) -NFPA 805 Performnance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Editions) Acceptance Review Clarification Response.

4. St. Lucie Plant Units I and 2 Request for Additional Information on License Amendment Request to Adopt National Fire Protection Association Standard 805 Performance-Based Standard for Fire Protection (TAC Nos. MF 1373 and MF 1374) dated December 26, 2013.

Per Reference 1 above, Florida Power and Light Company (FPL) requested an amendment to the Renewed Facility Operating License (RFOL) for St. Lucie Units 1 and 2. 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.

Per Reference 3 FPL responded to NRC LIC-109 acceptance review questions received by FPL via Reference 2 to clarify aspects of the LAR submittal.

Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390. to this letter contains security-related information. Upon removal of Enclosure 2, this letter is uncontrolled.

Ax Florida Power & Light Company 6501 S. Ocean Drive, Jensen Beach, FL 34957

L-2014-056 10 CFR 50.90 By letter dated December 26, 2013 (Reference 4) NRC Staff requested additional information regarding the LAR. Based on discussions with the NRC Staff, the additional information requested was prioritized and the response to the request for additional information will be provided in three separate submittals. The attachments to this letter provide the 60-day response to the request for additional information.

The information provided in this submittal does not impact the 10 CFR 50.92 evaluation of"No Significant Hazards Consideration" previously provided in FPL letter L-2013-099.

FPL requests that Enclosure 2 to this letter, which contains sensitive security-related information, be withheld from public disclosure in accordance with 10 CFR 2.390.

This letter makes new commitments and changes existing commitments. The commitment revisions are included in Enclosure 2 as mark-ups to Attachment S, Table S-2, Implementation Items.

Should you have any questions regarding this application, please contact Mr. Eric Katzman, Licensing Manager, at 772-467-7734.

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

Executed on February,'I, 2014.

Respectful submitted, cc President St. Lucie Plant JJ/rcs

Enclosures:

1. FPL's St Lucie Units 1 and 2 NFPA 805 LAR 60-Day RAI

Response

2. FPL's St Lucie Units 1 and 2 NFPA 805 LAR 60-Day RAI Response - Withheld from Public Disclosure cc:

Regional Administrator, Region II, USNRC Senior Resident Inspector, USNRC, St. Lucie Plant USNRC Project Manager for St. Lucie Plant Ms. Cindy Becker, Florida Department of Health Security-Related Information - Withhold From Public Disclosure Under 10 CFR 2.390. to this letter contains security-related information. Upon removal of Enclosure 2, this letter is uncontrolled.

Enclosure I to L-2014-056 Page 1 of 75 and 2 NFPA 805 LAR 60-Day RAI Response FPL's St Lucie Units 1 PSL FM RAI 02a PSL FM RAI 02b PSL FM RAI 02c PSL FM RAI 02d PSL FM RAI 02e PSL FM RAI 05a PSL FM RAI 05b PSL FM RAI 05c PSL FPE RAI 02 PSL FPE RAI 03 PSL FPE RAI 04 PSL FPE RAI 05 PSL FPE RAI 06 PSL FPE RAI 07 PSL FPE RAI 08 PSL FPE RAI 09 PSL FPE RAI 10 PSL FPE RAI 11 PSL PRA RAI 04b PSL PRA RAI 04c PSL PRA RAI 07 PSL PRA RAI IOa PSL PRA RAI 10b PSL PRA RAI IOd PSL PRA RAI IOe PSL PRA RAI IOf PSL PRA RAI 12 PSL PRA RAI 13 PSL PRA RAI 15a PSL PRA RAI 15b PSL PRA RAI 15c PSL PROG RAI 01 PSL PROG RAI 02 PSL SSA RAI 01 PSL SSA RAI 02 PSL SSA RAI 03 PSL SSA RAI 05 PSL SSA RAI 06 PSL SSA RAI 07 PSL SSA RAI 08 PSL SSA RAI 09 PSL SSA RAI 10 PSL SSA RAI 11 PSL SSA RAI 12 PSL SSA RAI 13 PSL SSA RAI 14

Enclosure I to L-2014-056 Page 2 of 75 PSL FM RAI 02a American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-S-2008, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications.", Part 4, requires damage thresholds be established to support the FPRA. Thermal impact(s) must be considered in determining the potential for thermal damage of Structure, Systems, and Components (SSCs). Appropriate temperature and critical heat flux criteria must be used in the analysis.

Provide the following information:

a. Describe how the installed cabling in the Units 1 and 2 power block was characterized, specifically with regard to the critical damage threshold temperatures and critical heat flux for thermoset and thermoplastic cables as described in NUREG/CR-6850.

RESPONSE

The cabling in both Unit 1 and Unit 2 is characterized as thermoplastic as described in Section 6.1 of Report 0493060006.104 ("St. Lucie Units 1 and 2 Fire Probabilistic Risk Assessment Fire Scenario Report NUREG/CR-6850 Tasks 8 and 11"). The generic damage thresholds for thermoplastic cable as described in NUREG/CR-6850, Appendix H, Table H-I are used in the Fire PRA to determine cable damage. Specifically, the temperature damage threshold that is applied to cable targets is 400'F for hot gas layer and thermal plume exposures. The heat flux damage threshold that is applied to cable targets is 0.5 Btu/s-ft2. The damage threshold for exposure conditions involving an elevated temperature and an incident heat flux consider the combined effects of both the temperature and heat flux components.

PSL FM RAI 02b American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-S-2008, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications.", Part 4, requires damage thresholds be established to support the FPRA. Thermal impact(s) must be considered in determining the potential for thermal damage of Structure, Systems, and Components (SSCs). Appropriate temperature and critical heat flux criteria must be used in the analysis.

Provide the following information:

b. The technical documentation supporting the LAR that describes the fire modeling that was performed seems to imply that IEEE [Institute of Electrical and Electronics Engineers]-383 qualified cables are assumed to be equivalent in terms of damage thresholds to "thermoset" cables as defined in Table 8-2 of NUREG/CR-6850. In addition, non-IEEE-383 qualified cables are assumed to be equivalent to "thermoplastic" cables as defined in Table 8-2 of NUREG/CR-6850.

These assumptions may or may not be correct. An IEEE-383 qualified cable may or may not meet the criteria for a "thermoset cable" as defined in NUREG/CR-6850. It is also possible that a non-IEEE-383 qualified cable actually meets the NUREG/CR-6850 criteria for a "thermoset" cable.

Provide clarification on the assumptions that were made in terms of damage thresholds of cables.

Enclosure I to L-2014-056 Page 3 of 75

RESPONSE

The generic damage thresholds for thermoplastic cable as described in NUREG/CR-6850, Appendix H, Table H-I were used to characterize cable targets in the fire modeling analyses supporting the Fire PRA for both Unit I and Unit 2 (see Section 6.1 of Report 0493060006.104).

These damage thresholds are lower than the damage thresholds for thermoset cables and do not factor in the IEEE-383 qualification status. Refer to the response to PSL RAI FM 02a, "Cable Characterization" for additional details on the cable damage thresholds that are used in the Fire PRA.

PSL FM RAI 02c

c. Describe how cable tray covers and conduits affect the damage thresholds that were used in the fire modeling analyses.

RESPONSE

Neither cable tray covers nor conduits were used as a reason to preclude damage and were subjected to the same zone of influence (ZOI) as all other targets in a given scenario. The same damage threshold was applied for cables routed in covered trays or in conduits as was applied for cables in trays which were not covered. The damage criteria for non-IEEE thermoplastic cables was applied for all cables. For PSL Unit 2 the majority of the cables used were Thermoset Kerite cables. Due to concerns regarding lower damage temperatures for Kerite cables, the thermoplastic damage temperature criteria was used for these cables also. The only credit for covered trays and conduits is in terms of secondary combustibles. Neither fire propagation nor additional HRR for the HGL calculations was postulated for covered trays or conduit.

PSL FM RAI 02d American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-S-2008, "Standard for Level I/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications.", Part 4, requires damage thresholds be established to support the FPRA. Thermal impact(s) must be considered in determining the potential for thermal damage of Structure, Systems, and Components (SSCs). Appropriate temperature and critical heat flux criteria must be used in the analysis.

Provide the following information:

d. Explain what damage thresholds were used in the fire modeling analyses for cables coated with Flamemastic 77.

RESPONSE

Flamemastic was not credited to limit damage or fire propagation for secondary combustibles in the PSL FPRA. The damage criteria for thermoplastic non-IEEE 383 cable was applied to all cables, whether coated or not coated with Flamemastic. PSL Unit 2 uses thermoset, Kerite cables, for which the thermoplastic non-IEEE 383 cable damage temperature criteria was also applied due to concerns regarding Kerite cables qualification.

Enclosure I to L-2014-056 Page 4 of 75 PSL FM RAI 02e American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-S-2008, "Standard for levell/large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications," Part 4, requires damage thresholds be established to support the FPRA. Thermal impact(s) must be considered in determining the potential for thermal damage of structures, systems, and components (SSCs). Appropriate temperature and critical heat flux criteria must be used in the analysis. Provide the following information:

e. Describe the methodology that was used to convert damage times for thermoplastic cables in Appendix H of NUREG/CR-6850 to percent damage as a function of heat flux, and discuss the assumptions and technical basis for this methodology. In addition, explain how this methodology was applied to determine damage to targets within the horizontal ZOI.

RESPONSE

The NUREG/CR-6850 Appendix H Table H-8 data provides times to target damage for thermoplastic cables for a set of constant incident heat flux values. This provides a time delay for target damage beyond the damage heat flux of 5.7 kW/mA2. For instance, Table H-8 provides a 19 minute time to damage delay for a thermoplastic cable with a constant incident heat flux of 6 kW/mA2.

In order to apply the NUREG/CR-6850 Appendix H data to a fire with a t^2 growth rate, a methodology of damage accrual is applied. 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 enviromriental qualification (EQ) of components such as cables in a containment accident environment, to determine the time to damage of equipment and cables. A discussion of an NRC internal evaluation of the Arrhenius methodology for environmental qualification is provided in a February 24, 2000 NRR Memo from Samuel J. Collins to Ashok Thadani (ML003701987).

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 1

kW/m^2 incident heat flux in Table H-8 is converted to a 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 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 will be used 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 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

Enclosure I to L-2014-056 Page 5 of 75 conservative, bounding analysis.

profile that models this approach.

Figure 1 below shows a plot of the damage rate - heat flux Damage Rate vs. Heat Flux C

SE SDR(q) 4)

cc 20 q

Heat Flux [kw/m2]

Figure 1. Damage rate - heat flux profile.

Note: The above curve uses a profile more conservative than that specified in NUREG/CR-6850, Appendix H. Appendix H assumes no damage below the critical heat flux (the above curve assumes the rate of damage below critical heat flux is the same as at the critical heat flux).

Appendix H also assumes a damage rate of 1/minute at a heat flux greater than 16 kW/m^2 (the above curve allows for application of higher damage rates above 16 kW/m^2) as defined by the curve derived from the Appendix H, Table H-8 data.

In order to calculate a time to damage, the t^2 heat flux - time profile is then correlated to the damage rate - heat flux profile to produce a damage rate - time function. The damage rate - time function is calculated by performing point by point multiplication over the heat flux - time and the damage rate - heat flux profiles. Figure 2 provides the plots of the damage rate - time function for several example HRR bins.

Enclosure I to L-2014-056 Page 6 of 75 Damage Rate vs. Time at a Given Distance "z" SJ I,

J Sv (2(.,.0,))

I 0.6-SD'p(q=(z.t.5))

S DR(q=(z.t -)))

/

1q z/-t-1 1 0, D.;L(qt(z.t. 14))

/

/

0 I

0 5

10 15 Time, t (mintres)

Figure 2. Damage rate - time function for several bins.

In order to calculate the time to target damage, the damage rate - time function is integrated, resulting in target accrued damage over time. The time at which the cumulative accrued damage for each bin equals 1.0 is the resulting time to target damage. Attachment 1 provides a simplified example calculation evaluating the time to target damage for a thermoplastic target located 3 vertical feet from a NUREG/CR-6850 Appendix E Case 3 Bin 3 fire.

The approach described above is focused on evaluating the time to damage using a vertical ZOI distance. This approach is conservative because vertical targets see both direct fire heat flux and plume temperature, where plume temperature is frequently driving the damage times. In situations where the most critical target is in the horizontal direction, the horizontal ZOI distance must be transformed to an equivalent vertical ZOI distance. This correlation between vertical and horizontal distance is performed using the vertical and horizontal ZOI distances provided in the Generic Fire Modeling Treatments (GFMTs). For instance, for a Case 1 medium sized electrical cabinet, the GFMTs provide horizontal and vertical distances in Table 5-10 and Table 5-11, respectively, for the 15 NUREG/CR-6850 Appendix E bins. A linear regression was performed on this data to generate a function that correlates horizontal distance to vertical distance. Applying this approach to horizontal targets is conservative because horizontal targets will not be subject to plume temperature effects.

Enclosure I to L-2014-056 Page 7 of 75 PSL RAI FM 02e - Attachment 1 PSL RAI FM 02e: Example Case 3 Bin 3 calculation This calculation performs a simplified evaluation for the time to damage of a target using the damage accrual methodology.

This example will evaluate a NUREG/CR 6850 Appendix E Case 3, single cable bundle thermoplastic, bin 3 ignition source fire impacting a target at a vertical distance of 3 feet. The calculation is organized into 4 sections: input data, heat flux profile damage rate profile, and correlation of the heat flux and damage rate profiles.

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

HRR - the HRR distribution row for a Case 3 bin 3 fire from NUREG/CR 6850 Appendix E HFdistance - The Generic Fire Modeling Treatments Table 5-11 provides vertical damage distances for an Appendix E Case 3 bin 3 fire impacting thermoplastic (5.7 kW/mA2), class A combustible (9 kW/mA2), and thermoset (11 kW/mA2) targets. These three heat flux - distance points are used for the first three columns. The Generic Fire modeling Treatments Table 5-4 provides 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 kW/mA2 is incident. The max heat flux - distance point for a medium sized cabinet is used as the fourth column in the matrix.

distance - the vertical distance from the fire to the target q_data - tabulates the four heat flux points 5.7, 9, 11, and 120 kW/mA2 damagedata - tabulates the damage time from NUREG/CR 6850 Appendix H Table H-8 HRR=

"Lower Bound" "Upper Bound" "Point Value" "SF" H53 79 65 0.192)

HF distance :=

"5.7 kW/m^2" "9 kW/m^.2" "I1 kW/m^22" "120 kW/MA^2"

6.4 5.4 4.9 1.15 distance := 3 5.7) q_data: [11.4/

ý, 120 damage data :

(6 8

10 11 14

,16 19) 10 6

4 2

1 Page 1 of 6

Enclosure I to L-2014-056 Page 8 of 75 PSL RAI FM 02e - Attachment 1 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 NUREG/CR 6850 Appendix G, replacing peak HRR with peak heat flux.

(HI-wdistance T) (1) =

6.4 5.4 4.9 1.15)

Extract from input data the relevant distance data.

(1) 50 153.551 '

qft=pwrfi (HIFdistanceT)1 qgata{,j1 L-17.304 1f(

-3

-7.964 Fflux(d,A,B,C) := A-dB + C Perform a power regression of the heat flux and distance data to generate a heat flux -

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

Generate generic power function that accepts A, B, and C coefficients.

Apply the power regression coefficients to the generic power function, using a maximum heat flux of 120 kW/mA2.

qz(d) :=

120 if Fflux(d, qfit qfit0,qfilft)

> 120 Fflux(d,qfit0,qfitl,qfit 2) otherwise Heat Flux vs. Vertical Distance for Select HRR Bins qIId 0

2 4

6 8

d Distance [ft]

10 Page 2 of 6

Enclosure I to L-2014-056 Page 9 of 75 PSL RAI FM 02e - Attachment 1 q_max 3 := qz(3) = 28.682 Maximum 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 will increase proportionally to the HRR increase of the ignition source. Therefore, to determine the heat flux as a function of time, the equation from NUREG/CR-6850 Appendix G was used. The peak HRR was replaced with the peak heat flux, q3 ' as determined by the function qz(d).

q3(t):=

ma 1.10-4,mi 120,q_max 3.+/-Jj if t <12 niax(l.10-4, min(120,q_max3 ))

otherwise Incident Heat Flux vs. Time at a Distance of 3 feet 01 q3(t) 0 2

4 6

8 Time, t [minutes]

10 Page 3 of 6

Enclosure I to L-2014-056 Page 10 of 75 PSL RAI FM 02e - Attachment 1 Damage Rate Profile damage_q := damage-data"0) damage -rate.- damage time 6

8 10 11 14 16) damagetime:= damage-data(I) = 6~

41 (191 161 (0.053 0.1 0.167 0.25 0.5 1

1 4.702 x03

=

expfidanaaeq,damagejate, lI, 3 4 Jue q

0.332 DRcurve(q) :=C

0. exp(q.C 1) + C 2 Perform an exponential regression of the heat flux and distance data to generate a heat flux -

damage rate function coefficients A, B and C, for the Case 3 bin 3.

Generate an exponential function that uses the A, B, and C coefficients.

Ensure a minimum value of 1/19 damage rate is applied (1

N DR(q) :=

if q <5.7 19 I

if q > 5.7 A DR curve(q) < -

19 DR curve(q) if q > 5.7 A DR curve(q) 19 The plot below shows the damage rate vs. heat flux function produced from the exponential regression. As the plot shows, a damage rate equal to the critical damage heat flux damage rate of 1/19 per minute is applied to heat fluxes up to and including the critical heat fluz of 5.7 kW/mA2.

No maximum damage rate boundary is applied.

Page 4 of 6

Enclosure I to L-2014-056 Page II of 75 PSL RAI FM 02e - Attachment 1 Damage Rate vs. Heat Flux S

0.

I-

~

D~jq) 4 0

0 4-.

S 0

5 10 15 q

Heat Flux [kw/m2]

20 Page 5 of 6

Enclosure I to L-2014-056 Page 12 of 75 PSL RAI FM 02e - Attachment I 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 over time, this function needs to be integrated.

Damage Rate vs. Time at a Distance of 3 feet tw 03 4-Id 4-to03 CE DR(q 3(t))

0 5

10 t

Time, t [minutes]

15 The function below integrates the damamge rate vs. time function using a time step of 0.1. The integration terminates when the accrued damage equals 1.0. The time step at which this occurs is the time to target damage.

tdam:=

tstep <-- 01.

Ding - 0 tdmg <-- 0 while Drng < I tdmng <- tdmng + tstep Ding -- Ding + DR(q3(tdmg))'tstep I tdain - tdmg tdam = 8.6 Page 6 of 6

Enclosure I to L-2014-056 Page 13 of 75 PSL FM RAI 05a NFPA 805, Section 2.7.3.4, "Qualification of Users," states: "Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the fire PRA development (NFPA 805 Section 4.2.4.2). This requires that qualified fire modeling and PRA personnel work together. Furthermore, LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states:

"Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.

During the transition to 10 CFR 5 0.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g. fire modeling) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805, Section 2.7.3.4.

Post-transition, for personnel performing fire modeling or fire PRA development and evaluation, Florida Power & Light Company will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805, Section 2.7.3.4, to perform assigned work. See Implementation Item 15 in Table S-2 of Attachment S (see LAR Attachment S)."

Regarding qualifications of users of engineering analyses and numerical models (i.e., fire modeling techniques):

a. Describe the requirements to qualify personnel for performing fire modeling calculations in the NFPA 805 transition.

RESPONSE

The governing requirement for qualifying personnel performing fire modeling calculations during the NFPA 805 transition was NFPA 805, Section 2.7.3.4. The qualifications for fire modeling personnel were based on the education and background for those individuals performing the fire modeling tasks. Specific fire modeling tasks (development of the "Generic Fire Modeling Treatments" report and its associated supplements and the preparation of the Main Control Room abandonment calculation) was carried out by fire protection engineers that meet the qualification standards described in R.G. 1.189, Section 1.6.1.a.The qualifications that are required for the staff and consulting engineers that use and apply these technologies depend in part on their specific assigned role on the project. In general, the qualification requirements for those who are technical leads in the preparation of technical tasks are consistent with and often exceed those articulated in NEI 07-12 for qualification of Peer Reviewers and thus meet the qualification requirements of NFPA 805, Section 2.7.3.4. Section 2.2 of NEI 07-12 describes the desired experience requirements for Peer Reviewers. Similar qualification was expected of technical leads involved in the performance of the PSL fire PRA. Given the magnitude of the technical activity that was

Enclosure I to L-2014-056 Page 14 of 75 performed, the technical leads were assisted by support staff. There are no specific qualifications for those in a support role as the assigned technical lead would retain overall technical responsibility for the entire body of work.

PSL FM RAI 05b NFPA 805, Section 2.7.3.4, "Qualification of Users," states: "Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the fire PRA development (NFPA 805 Section 4.2.4.2). This requires that qualified fire modeling and PRA personnel work together. Furthermore, LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states:

"Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.

During the transition to 10 CFR 50.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g. fire modeling) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805, Section 2.7.3.4.

Post-transition, for personnel performing fire modeling or fire PRA development and evaluation, Florida Power & Light Company will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805, Section 2.7.3.4, to perform assigned work. See Implementation Item 15 in Table S-2 of Attachment S (see LAR Attachment S)."

Regarding qualifications of users of engineering analyses and numerical models (i.e., fire modeling techniques):

b. Describe the process for ensuring that the fire modeling personnel meet those qualifications, not only before the transition but also during and following the transition.

RESPONSE

The fire modeling calculations performed during the transition were conducted by fire protection contractors having the appropriate qualifications as described in the response to PSL RAI FM 05a "Fire Modeling Qualifications." As stated in the response to PSL RAI FM-05a personnel performing fire modeling activities during the time supporting the LAR and during transition met the requirements in RG 1.189, Section 1.6.1.a. The qualification of these personnel often exceed the minimum requirements in the RG and have significant experience in fire modeling specifically supporting NFPA 805 transitions.

Personnel who will perform fire modeling calculations following transition will be qualified to a specific qualification mentoring guide for fire modeling. Based on the usage of the generic fire modeling treatments at St. Lucie it is expected that little detail fire modeling will be required.

Enclosure I to L-2014-056 Page 15 of 75 However, the understanding of the generic fire modeling treatments and their limitations is part of the qualification to perform fire PRA analysis. Refer to the response to PSL RAI PROG 02 "Qualification of Users," regarding training of engineering personnel to qualify them to perform fire modeling post transition.

PSL FM RAI 05c NFPA 805, Section 2.7.3.4, "Qualification of Users," states: "Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the fire PRA development (NFPA 805 Section 4.2.4.2). This requires that qualified fire modeling and PRA personnel work together. Furthermore, LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states:

"Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.

During the transition to 10 CFR 5 0.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g. fire modeling) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805 Section 2.7.3.4.

Post-transition, for personnel performing fire modeling or fire PRA development and evaluation, Florida Power & Light Company will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805, Section 2.7.3.4, to perform assigned work. See Implementation Item 15 in Table S-2 of Attachment S (see LAR Attachment S)."

Regarding qualifications of users of engineering analyses and numerical models (i.e., fire modeling techniques):

c. When fire modeling is performed in support of FPRA, describe how proper communication between the fire modeling and FPRA personnel is ensured.

RESPONSE

Throughout the Fire Probabilistic Risk Assessment (FPRA) process, the Fire Protection Engineers (FPE) who conducted the fire modeling and the Fire PRA personnel maintained frequent communications. Fire modeling calculations were initiated with a description of the fire modeling objective, a visual inspection of the area by the FPE engineers, and a design input document transfer. FPRA team members reviewed the documentation prior to its incorporation into the FPRA model. The fire modeling results that are used by the FPRA are contained in calculations which are reviewed in accordance with the appropriate Quality Assurance (QA) program. These calculations were reviewed under the contractors' QA program with individuals familiar with the technical aspects of the calculation. Direct communication was also provided through the

Enclosure I to L-2014-056 Page 16 of 75 calculation review process. The fire modeling calculations were reviewed by FPRA and FPE contractors and FPL personnel, and comments were provided and addressed by the FPE staff. During the preparation of the LAR, meetings were held between the FPRA and FPE staff to review the necessary fire models and to ensure the results accurately reflected the needs of the FPRA model.

The post transition process (currently being implemented to support Duane Arnold) implements the process described in FAQ 07-0061. This process will have a screening done by the responsible engineer. If that screening indicates that there is any potential to impact the fire protection program the screening will have to be reviewed by a qualified fire protection engineer. Any actual impacts to the fire protection program will be reviewed by both a qualified fire protection engineer and a qualified FPRA engineer working together. This ensures proper communication between fire protection and FPRA.

PSL FPE RAI 02 National Fire Protection Association Standard 805, (NFPA 805) "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," Section 3.3.1.2(6) requires the storage and use of flammable gases to be "in accordance with applicable NFPA standards".

However, LAR Attachment A, Section 3.3.1.2(6) indicates compliance with something other than NFPA standards. Describe the requirements used and provide justification for their use in lieu of NFPA standards.

RESPONSE

The PSL Response to Section 3.3.1.2(6) states:

3.3.1.2 (6)

• Controls on use and storage of flammable gases shall be in accordance with applicable NFPA standards.

Compliance Statement:

Complies with Clarification Compliance Basis:

PSL is not committed to any flammable gas standards, with the exception of NFPA 50A for bulk hydrogen, and as such are not part of the current license basis. Per FAQ 06-0020, the following guidance applies as to which NFPA standards referenced in Chapter 3 are applicable.-

"Where used in NFPA 805, Chapter 3, the term, "applicable NFPA Standards" is considered to be equivalent to those NFPA standards identified in the current license basis (CLB) for procedures and systems in the Fire Protection Program that are transitioning to NFPA 805."

Flammable gases are controlled per AP-0010434.

PSL does not have any NFPA standards explicitly in its current license basis identified for use and storage of flammable gas other than NFPA 50A for bulk hydrogen storage. PSL controls flammable gases per a plant administrative procedure, AP-0010434.

AP-0010434, Sections 8.2.12 and 8.8 contain direction on the use of flammable gases (note that the LAR did not cite section 8.2.12).

Enclosure I to L-2014-056 Page 17 of 75 Section 8.2.12 discusses control of combustibles, specifically bulk compressed or cryogenic flammable gas storage. Specific controls include:

" Storage is not permitted inside structures house systems, equipment, or components important to nuclear safety

" Storage of flammable gas shall be located outdoors, or in separate detached buildings

" NFPA 50A, Standard for Gaseous Hydrogen Systems at Consumer Sites, shall be used for hydrogen storage

" Outdoor high-pressure flanimmable gas storage containers shall be located so that the long axis is not pointed at buildings

" Flammable gas storage cylinders not required for normal operation shall be isolated from the system Section 8.8 discusses requirements for handling flammable gases. Specific controls include:

" When handling flammable gases, release to the air should be avoided in that it may provide a means for a combustion explosion, fire hazard or both

" In areas where flammable gases are handled, adequate fire extinguishing capability shall be provided

" When handling flammable gases, some form of ventilation shall be provided. In a closed storage area, exhaust ventilation shall be provided

" When transferring flammable gases, a ground strap will be connected between the containers involved whenever the possibility of electrical sparking exists These controls have been developed over years of operating experience with industry guidance.

No specific NFPA standards other than NFPA 50A are cited in the administrative procedure. This is acceptable as defined in FAQ 06-0020.

PSL FPE RAI 03 LAR Attachment A, Section 3.3.1.3.1 Control of Ignition Sources, identifies both "complies via Engineering Evaluation", and "Complies" as the compliance statement. Provide a description to clarify the difference, and describe how each compliance strategy is applied.

RESPONSE

There is no difference in the compliance strategies utilized for LAR Attachment A, Section 3.3.1.3.1. This NFPA 805 Chapter 3 element requires a hot work safety program. The PSL hot work program is implemented in AP-0010434 hence the "Complies" Compliance Statement.

Section 3.3.1.3.1 also states the program shall be developed in accordance with NFPA 51 B. PSL performed a code compliance evaluation for NFPA 51 B as documented in PSL-FPER-09-052 which demonstrates compliance of the PSL hot work program. Because the compliance was documented in an engineering evaluation the "Complies via Engineering Evaluation" Compliance Statement was also used. PSL will revise the compliance statements in the NFPA 805 LAR as shown on the attached LAR mark-up page for clarity. The "Complies via Engineering Evaluation" Compliance Statement will be struck and the "Complies" Compliance Basis enhanced to clarify that "Compliance with NFPA 51B is documented in PSL-FPER-09-052 and confirms AP-0010434 implements a compliant hot work program."

PSL FPE RAI 03 - Mark-up Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure I to L-2014-056 Page 18 of 75 Compliance NFPA 805 Ch. 3 Reference Requirements / Guidance Statement Compliance Basis 3.3.1.2 Control of Combustible 3.3.1.2 (5)

• Controls on use and storage of flammable and combustible Complies via See the NFPA 30 Code Compliance Evaluation for evaluation of Materials (5) liquids shall be in accordance with NFPA 30, Flammable and Combustible Engineering use and storage of flammable and combustible liquids. No other Liquids Code, or other applicable NFPA standards.

Evaluation NFPA standards were determined to be applicable based on FAQ 06-0020.

References Document ID PSL-FPER-1 1-005 Rev. 0 - NFPA Code Compliance Evaluation for NFPA 30, 1973 Edition, Flammable and Combustible Liquids Code Complies Additional requirements for handling flammable liquids are provided in AP-0010434.

References Document ID AP-0010434 Rev. 44 [Section 8.2.12, 8.7] - Fire Protection Guidelines 3.3.1.2 Control of Combustible 3.3.1.2 (6)

• Controls on use and storage of flammable gases shall be in Complies with PSL is not committed to any flammable gas standards, with the Materials (6) accordance with applicable NFPA standards.

Clarification exception of NFPA 50A for bulk hydrogen, and as such are not part of the current license basis. Per FAQ 06-0020, the following guidance applies as to which NFPA standards referenced in Chapter 3 are applicable:

"Where used in NFPA 805, Chapter 3, the term, "applicable NFPA Standards" is considered to be equivalent to those NFPA standards identified in the current license basis (CLB) for procedures and systems in the Fire Protection Program that are transitioning to NFPA 805."

Flammable gases a Enhance Compliance Basis as References Document ID follows:

"Hot work is controlled by AP-001 0434. Compliance with AP-0010434 Rev. 44 [Section 8.8] - Fire Protection Guidelines NFPA 51B is documented in PSL-FPER-09-052 and confirms that AP-0010434 implements a compliant hot 3.3.1.3 Control of Ignition 3.3.1.3 Control of Ignition Sources NIA Section Title work program."

.0.) I.Q. I L'V ;U VI ;~I ICILIUII

-f UC f

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SourcesI Sources Code Requirements]

Rafarances A hot work safety procedure shall be developed, implemented, and periodically updated as necessary in accordance with NFPA 51 B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work, and NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations.

Doc,-mnt ID Engnee~ing F;0 I.uatman ing and Welding Pr12Ocess Hot work is controlled by AP-0010434.

PSL -FPFII-Q9-Q5 U2 R~ev. U NI-PA. Gode Gompliance hvalwation for NI-.A. 51R, 1976 Eddt*Qn, G6't#

Ndd Reference -

ISL-FPER-09-052 Rev. 0 - NFPA

ode Compliance Evaluation Document ID or NFPA 51B, 1976 Edition, P-0010434 Rev. 44 [Section 8.3, 8.4] - Fire Protection Guidelines

utting and Welding Processes Complies Fire Safety Analysis Data Manager (4.129)

FPL - St. Lucie Run: 03/18/2013 01:10 Page: 9 of 61

Enclosure I to L-2014-056 Page 19 of 75 PSL FPE RAI 04 The compliance basis for LAR Attachment A, Section 3.3.3 Interior Finishes states that "walls, floors and ceilings are of reinforced concrete or concrete block construction and use of interior finish materials is limited to non-combustible material or those with a flame spread, smoke and fuel contribution of 50 or less (ASTM E-84) whenever practicable." However, NFPA 805 Section 3.3.3 requires interior wall or ceiling finish classification to be in accordance with NFPA 101, Life Safety Code requirements for Class A materials. NFPA 101 Section 10.2.3.4 defines Class A as "Flame spread index, 0-25". Provide a justification for the apparent finish classification discrepancy.

RESPONSE

PSL complies with NFPA 805 Section 3.3.3 in the following manner:

1) The control rooms, which contain the majority of interior finish materials at PSL, are evaluated for compliance via engineering evaluations. Engineering evaluations PSL-FPER-05-047, PSL-FPER-05-048, and PSL-FPER-10-029 determined that the interior finishes of the control rooms comply with the requirements of NFPA 805. PSL-FPER-10-029 also determined that Unit 2 overall is compliant with the requirements of NFPA 101, 1973 edition. The control rooms and Unit 2 overall meet the requirements of NFPA 805 Section 3.3.3 as documented in engineering evaluations.

2) PSL does utilize epoxy floor finishes. The request for NRC approval of epoxy floor finishes is documented in LAR Attachment L, Approval Request 2. Epoxy floor coatings will be acceptable upon approval documentation in the NFPA 805 Safety Evaluation.

3) The balance of the buildings and rooms within the PSL Power Block are constructed primarily of steel and concrete, which may or may not be painted. Floors are either unfinished concrete, painted concrete, or vinyl floor tiles over concrete. The PSL UFSAR Appendix 9.5A, Table 2.4, Section D. 1(d) states that walls and structural materials are noncombustible. This statement is in response to Appendix A of BTP 9.5-1 guidelines that requires interior finishes to have a flame spread rating of 25 or less. In addition, Regulatory Guide (RG) 1.189, Fire Protection for Nuclear Plants, Section 4.1.1.1 states that interior finishes should be noncombustible and further lists materials acceptable for use as interior finish without evidence of test and listing by a recognized testing laboratory. These materials include (in part):

" plaster, acoustic plaster, and gypsum plasterboard (gypsum wallboard), either plain, wallpapered, or painted with oil-or water-base paint,

" brick, stone, and concrete blocks, plain or painted,

" steel and aluminum panels, plain, painted, or enameled, and

" vinyl tile, vinyl-asbestos tile, linoleum, or asphalt tile on concrete floors PSL interior finish materials are noncombustible and consistent with the guidance in Appendix A to BTP 9.5-1 Guidelines Section D,I(d) and Regulatory Guide 1.189, Section 4.1.1.1 and subsequently the requirements of NFPA 805 Section 3.3.3.

The current UFSAR Appendix 9.5A wording, which is inconsistent between Section 2.1 and Table 2.4, Section D.1(d), will be revised as part of implementation. UFSAR Section 9.5A is currently the Fire Protection Program Report. This information will be revised during the creation of the Fire Protection DBD. Implementation Item #15 creates a fire protection

Enclosure I to L-2014-056 Page 20 of 75 design basis document, Implementation Item #3 updates station documentation to include the applicable interior finish requirements from NFPA 101, and Implementation Item #9 will ensure the details supporting compliance bases in Table B-1 (Attachment A) that originate from the UFSAR are carried forward as a nuclear record. These implementation items will resolve the discrepancy between current program documentation and the requirements of NFPA 805.

PSL will revise the compliance statements in the NFPA 805 LAR Attachment A, Section 3.3.3 as shown on the attached LAR markup page for clarity.

PSL FPE RAI 04 - Mark-up Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Enclosure I to L-2014-056 Page 21 of 75 Compliance NFPA 805 Ch. 3 Reference Requirements / Guidance Statement Compliance Basis Unit 2 UFSAR, Appendix 9.5A Rev. 20 [Section 2.1]- Fire Protection Program Report 3.3.3 Interior Finishes 3.3.3 Interior Finishes.

Complies via For Unit 1 Control Room compliance is documented in PSL-References Interior wall or ceiling finish classification shall be in accordance with Engineering FPER-05-047.

NFPA 101, Life Safety Code, requirements for Class A materials.

Evaluation Interior floor finishes shall be in accordance with NFPA 101 requirements For Unit 2 Control Room comp for Class I interior floor finishes.

FPER-05-048 and PSL-FPER-Document ID PSL-FPER-05-047 Rev. 3 - Fire Protection Evaluation for PSL Unit 1 Control Room Modification (PCM 04115)

PSL-FPER-05-048 Rev. 4 - Fire Protection Evaluation for PSL Unit 2 Control Room Modification (PCM 04115)

PSL-FPER-10-029 Rev. 0 - NFPA Code Compliance Evaluation for NFPA 101 (Interior Finishes), 1973 Edition, Life Safety Code liance is documented in PSL-10-029.

Submit for NRC Epoxy floor coatings were reviewed under CR 2007-25587 and Approval determined to meet the requirements contained in GL 86-10 (no more than 1/8" thick with flame spread of 50 or less). This is not the same as the requirements from NFPA 805.

See Attachment L of the License Amendment Request for further details on the request for NRC approval for evaluation of epoxy floor coatings interior finish requirements.

Document ID CR 2007-25587 Rev. 0 - OE Review of IN 07-26 Combustibility of Epoxy Floor Coatings at Commercial Nuclear Power Plants References Insert:

The balance of the buildings and rooms within the PSL Power Block are constructed primarily of steel and concrete, which may or may not be painted. Floors are either unfinished concrete, painted concrete, or vinyl floor tiles over concrete.

These materials are noncombustible and consistent with guidance in Appendix A to BTP 9.5-1 Guidelines Section D. 1 (d) and Regulatory Guide 1.189, Section 4.1.1.1.

Complies The 'Unit I and Unit 2 Fire Protection Prooram Rero.rt. Section Non crombustible 2nd fire resistant mtatrials are used throughout the St. Lucie plant: walls, floors and ceiling aFre of reinforcsd concrete or concrete block constru-ction and use of interior finish materia's is limited to non combu-tible material or those with a flameO spread, smoke and fuel contriblution of 50 or less (AST-M F 8 4) whenever practicablee.

Implementation Item: The applicable station documentation will be updated as required to include the interior finish requirements per NFPA 805.

See Implementation Item 3 in Table S-2 of Attachment S.

Insert:

Table 2.4, Section D. I (d)

I IJ References Document ID Unit 1 UFSAR, Appendix 9.5A Rev. 25 [Section 2.1] - Fire Protection Program Report Unit 2 UFSAR, Appendix 9.5A Rev. 20 [SeGt#e 24.] - Fire Protection Program Report lIns Insert:

Table 2.4, Section D. l (d)

Reg Pov ert new

Reference:

gulatory Guide 1.189, Fire Protection for Nuclear ver Plants, Revision 2, Section 4.1. 1.1 Fire Safety Analysis Data Manager (4.129)

FPL - St. Lucie I'*lul1. UJ/ I OI/U IO U 1. 1U I Page: 11 of 61

Enclosure I to L-2014-056 Page 22 of 75 PSL FPE RAI 05 LAR Attachment A lists the requirement for NFPA 805 Section 3.4.4, "Firefighting Equipment." However, radiation monitoring, personal dosimeters, and fire extinguishers are not discussed in the LAR. Provide additional details for these elements, listed in the NFPA 805 Section 3.4.4 requirement.

RESPONSE

The NFPA 805 Chapter 3 requirement for Section 3.4.4 states firefighting equipment shall be provided for the fire brigade which includes protective clothing, respiratory protective equipment, radiation monitoring equipment, personal dosimeters, and fire suppression equipment such as hoses, nozzles, fire extinguishers, and other needed equipment. The Compliance Basis provided in the LAR only specifically identified fire brigade turnout gear and breathing apparatus. The NFPA 805 Section 3.4.4 response will be enhanced to address the other elements listed in the section requirement.

" A Radiation Team is responsible to verify proper levels and precautions for radiation and contamination concerns as identified in the Fire Fighting Strategies (1(2)-1800023). The fire brigade does not respond with radiation monitoring equipment. This equipment is brought, as needed, by the Radiation Team.

" Operations personnel that also serve on the fire brigade are trained for access to radiological areas (Radiological Worker Training). In order to access radiological areas, personal dosimetry (TLD) and other dosimetry are issued to each individual.

" Fire extinguishers and hose stations are installed throughout the plant for fire brigade use as documented in the NFPA 10 (fire extinguishers) code compliance evaluation (PSL-FPER-10-00 1), in the NFPA 14 (hose stations) code compliance evaluation (PSL-FPER 007), and in the Fire Fighting Strategies (1(2)-1800023), and drawings 8770-G-060 through

-077 and 2998-G-060 through -077.

" Additional fire suppression equipment provided for the fire brigade includes hydrants, hose houses (hose, nozzles, wrenches, gate valves, spanners), and dedicated equipment storage cages/locations (foam, fans, SCBA). Hydrant locations are documented in the NFPA 24 code conformance evaluation (PSL-FPER-08-070) and drawings 8770-G-170 & -

174 and 2998-G-170 & -174. Contents of hose houses and fire cages are documented in surveillance procedures.

PSL FPE RAI 06 LAR Attachment A, Section 3.3.5.1 identifies the need for 10 CFR 50.48(c)(2)(vii) approval for use of non-plenum rated cables above the suspended ceiling of the control room (CR). LAR Attachment L, Approval Request #3 indicates that "the wiring/cable may be small amounts of video /communications cable that is not listed for plenum use as required by this section of the code." Provide a more definitive description of the "limited amount" of unqualified cabling materials and quantities (either inside or outside the raceways). Include a more detailed justification for the conclusion that it "does not present a significant fire hazard." Describe whether the areas above the suspended ceilings are provided with credited fire detection and how it is identified in LAR Attachment C, Table C-2.

Enclosure I to L-2014-056 Page 23 of 75

RESPONSE

LAR Attachment L, Approval Request #3 was for the use of non-plenum cables above suspended ceilings in multiple areas. The specific request is for control cable trays above the Unit 1 Control Room Complex and limited amount of video/communication cable above the suspended ceilings.

LAR Attachment L states: "This approval request is specifically for the cable trays above the ceiling in the Unit I Control Room Complex and limited amounts of video/communication cable above all suspended ceilings." This response will address the cable trays above the Unit 1 Control Room Complex separately from the more generic 'limited amounts of video/communication cable" Limited Amount Outside Raceways A more definitive description of the "limited amount" of unqualified cabling materials and quantities (either inside or outside the raceways) has been requested.

As identified in PSL-FPER-12-001 the cables that are not tracked within the existing cable and raceway system and may not be routed in conduit are those cables associated with communications. Communications can be more clearly broken down into the categories of standard telephone, PAX, Public Address (PA), Sound Powered Phones and other computer cables required for interface with equipment in the Control Room. Controlled drawings identify that the cable associated with PAX, PA, and Sound Powered Phones are routed in conduit. (8770-G-333 Sh 1 and Sh 2) The potential for concern regarding computer cable and standard telephone cables is what remains.

PSL-FPER-12-001 did not identify any obvious areas where there might be unscheduled/unrated cable e.g. no cables were identified as hanging down from the overhead or penetrating the suspended ceiling. However, the walkdowns and reviews done for PSL-FPER 001 cannot preclude that unscheduled/unrated cable exist because the suspended ceilings were not removed and the entire area above the ceiling was not visually observed.

Computer Cables : Modifications made to the Control Room (EC 235249 and EC 235438) related to the installation of raised floor, and workstations (digital controls and digital displays) identified that communications cable is considered "L" (instrumentation). The modifications identified that cables associated with equipment relocated or installed by the modification would be routed in the raised floor area of the Control Room. In general, it would not have been considered efficient to have routed computer cables in a manner that would have included a routing through the overhead. In a similar manner, it is unlikely that computer cable in other areas with suspended ceilings would have been run overhead. While visual examination of these areas shows that little, if any, cable was run up into the ceiling, the presence of such cables above the ceiling cannot be excluded.

Telephone Cable: Telephone cable is not tracked or controlled by St. Lucie. Similar to computer cable described above, the locations of telephones currently present in the Control Room Complex of Unit 1 and Unit 2 do not afford themselves to having been routed within the overhead of the Control Room, predominantly due to their location within the rooms involved as well as the modifications performed to implement the Control Room Digital Upgrades. As described with computer cables above, a visual examination of these areas shows little potential for these types of cables to above suspended ceilings but this cannot be excluded.

Enclosure I to L-2014-056 Page 24 of 75

==

Conclusion:==

While it is unlikely that any unscheduled/unrated cable exists above the suspended ceilings in the St. Lucie Power Block, it cannot be explicitly excluded. With respect to 'limited amount' as stated in the Attachment L approval request a limited amount means that there may be a single cable or a small group of cables above the suspended ceilings that are unscheduled (e.g.

not tracked in the cable and raceway system) and potentially not plenum rated. There was no plan to route such cables above the suspended ceilings and that type of cable routing is not used.

However, there may have been some cables routed above the suspended ceilings to support things such as additional telephone or communication equipment to support outage work or test/computer cables to support special testing that may still remain (likely no longer connected) as a single cable or isolated sections of a small group of cables. While it has not been the practice to route cables above the suspended ceilings at St. Lucie (except for cable trays in the control rooms [see below])

the approval request for 'limited amounts' is meant to cover the fact that these cables may exist.

Non-Plenum Cables in Raceways (applies to Unit I Control Room Only)

As stated in the LAR Attachment L Approval Request 3 there are cables routed above the suspended ceilings in the Unit 1 Control Room Complex that may not be plenum rated and in ventilated trays. This does not meet the requirements of NFPA 805 Section 3.3.5.1 which would require these trays to be solid bottom with solid covers. The cable trays in question are 'C' type trays. There are only three such cable trays located above the suspended ceiling. The cables in these trays are either IEEE-383 qualified or covered with Flamemastic 77. While this is not equivalent to being plenum rated it does provide some assurance that these cables are difficult to ignite and limit the spread of combustion along the cable length. Two of these trays (C5 and C38) are at the bottom of the tray stack of 3 trays, where the other trays are "L" type with solid bottoms and covers. But one tray (C63) is located on the top of a stack with only one "L" tray beneath which is solid bottom and cover. There are no ignition sources in the vicinity of the three 'C' trays (C5, C38 and C63) and even if the cables in these trays were ignited the fire would be limited to these "C" trays and the 'L' trays above (C5 and C38) might be damaged they would not contribute any combustibles to the fire because the cables are completely enclosed.

Include a more detailed justification for the conclusion that it "does not present a significant fire hazard."

As stated above unscheduled/unrated cables are not expected to be present but cannot be excluded.

There are no ignition sources above the suspended ceilings and it is expected that the majority of these cables may not be energized. Even if energized, these cables would be low voltage communication/computer cables with little energy present that could cause these cables to fail and ignite. In addition, any cables if present would be scattered in location and would not provide any continuity of combustibles allowing any fire to spread.

Specifically for the Unit 1 the cables of concern are in a limited number of cable trays that are not in the vicinity of any ignition sources and even if they were to ignite there are no additional combustibles near these trays that would allow the fire to grow beyond the combustibles that are contained within the tray.

Tray Locations: The cable trays that are located above the Unit 1 Control Room suspended ceiling are routed in a manner such that they are closer to the perimeter of the Control Room and not densely packed in the overheads of the Unit 1 Control Room (8770-G-410 Sh 6 and 7).

Enclosure I to L-2014-056 Page 25 of 75 Unit 1 tray details: The "L" designated (instrumentation) cable trays in the Unit 1 Control Room are identified to be solid bottom with covers (8770-B-328 Sh. 5 and 9A). There are a total often (10) cable trays over the Control Room at various lengths. Seven (7) of these trays are designated as "L" trays and are therefore solid bottom with tray covers in place. Three of the trays are "C" trays and do not have covers and are ventilated trays. The actual percent cable fill in all the trays in the overhead is less than 20% at a majority (>85%) of all plant points associated with the Control Room (8770-B-328).

==

Conclusion:==

Based on the following points described in more detail above

" location of trays (limited in number and not densely packed over the entire Unit 1 Control Room),

" tray style (solid bottom with tray covers where identified - meets the NFPA 805 requirements),

" the quantity of trays not meeting the NFPA 805 requirements (3 for Unit 1),

" the actual percent fill (<20% for the majority of locations for all trays)

" Documented cables in Unit 1 is either IEEE 383 rated or non-rated and coated with Flamemastic (both difficult to ignite and difficult to have sustained combustion along their length)

Describe whether the areas above the suspended ceilings are provided with credited fire detection and how it is identified in LAR Attachment C, Table C-2.

Ionization smoke detection is provided over the cable trays in both Units Control Room Complexes as documented on plant drawings 8770-G-413, Sh. 11, 2998-G-413, Sh. 7. The smoke detection installed over the cable trays is identified in Table C-2 of the St. Lucie NFPA 805 LAR in Fire Area IF for zones 1-70 and 1-73 as Detection System, Unit 1 Zone 8B and for Fire Area 2F for zones 2-421 as Detection System, Unit 2 Zone 8A and Detection System, Unit 2 Zone 8B.

PSL FPE RAI 07 LAR Attachment C, Tables C-I and C-2 identify "enhanced transient controls" for certain areas (for example electrical penetration rooms, cable spread room, cable loft, and reactor auxiliary building corridor). Describe what those controls are, where in the plant these controls are established, and how this is managed by plant procedures. Describe how this is different than "Transient Exclusion Zones" listed in LAR Section 4.6.2.

RESPONSE

"Enhanced transient controls" are those controls that are beyond the current plant guidelines.

These enhanced transient controls may include:

" Additional / tighter limits on the amount and type of transient combustible materials in an area.

" Identification of transient exclusion or zero transient zones. This will result in no transient combustibles or ignition sources in specific locations. Currently these enhanced transient controls are in the following fire areas/zones where credit is taken in the Fire PRA. A fire watch is not required for this enhanced transient control.

Enclosure I to L-2014-056 Page 26 of 75 AreaZone 1BI157 2A2_51X 2B2_52 2H25 1E 212_51W The "transient exclusion zones" are not different than "enhanced transient controls" but a subset of the "enhanced transient controls". Note that at this time, under the current licensing basis, there are no transient exclusion zones at PSL.

For areas in the LAR where "enhanced transient controls" are identified (see list above which correlates to those areas listed in RAI PRA 4.a), storage of transient combustibles will not require a fire watch. This is documented in Section 1.2.8.1 of the Fire Risk Evaluation report (PSL-FPER-11-001) for the identified areas.

The enhanced transient controls will be further defined and refined during the implementation period concurrent with implementation of the Monitoring Program and will include development of a procedure which will manage the transient controls.

PSL FPE RAI 08 LAR Attachment K, identifies previously approved exemptions and deviations from 10 CFR 50, Appendix R that require transition to the NFPA 805 program. Provide clarification to the following:

a. LAR Attachment K, identifies radiant energy shields and 18 accessible fire extinguishers credited in Fire Area 1K as an element of the exemption 1-LA-03-19850221 for Unit 1 reactor coolant pump (RCP) oil collection system. LAR Attachment C, Table C-2 identifies radiant energy shields required for risk reduction only. Justify why the radiant energy shields and fire extinguishers shouldn't also be required for "E-EEEE/LA Criteria" as indicated in LAR Section 4.8.1.

b. LAR Attachment K, identifies fire extinguishers and hose stations for Fire Areas A, J, and L as one element of the basis for the exemption 1-LA-08-19850221 for the lack of fire rated dampers.

However, LAR Attachment C, Table C-2 does not identify either extinguishers or hose stations as fire protection features for the exemptions being transitioned. Justify why the fire extinguishers and hose stations shouldn't also be required for "E-EEEE/LA Criteria" as indicated in LAR Section 4.8.1.

c. LAR Attachment K, identifies fire extinguishers and standpipe system as part of the evaluation for exemption 1-LA-13-19870305, the lack of 3-hour rated penetration seals. However, these features are not identified in Attachment C, Table C-2. Justify why these features shouldn't also be required for "E-EEEE/LA Criteria" as indicated in LAR Section 4.8.1.

d. LAR Attachment K, identifies fire extinguishers and hose stations as elements of the basis for exemption 1-LA-14-19870305 for lack of fire rated dampers. Justify why the fire extinguishers and hose stations shouldn't also be required for "E-EEEE/LA Criteria" as indicated in LAR Section 4.8.1.

Enclosure I to L-2014-056 Page 27 of 75

e. Provide a review of the fire protection features being relied upon for any of the transitioned exemptions and deviations of LAR Attachment K. Provide justifications for any fire protection features not identified in LAR Attachment C, Table C-2.

RESPONSE

a) Licensing Action 1-LA-03-19850221 is the exemption for the Ul RCP oil collection system.

The basis for the exemption does not include radiant energy shields or fire extinguishers as credited fire protection features. The radiant energy shields and fire extinguishers discussed in the summary of letter dated September 16, 1983 which was a Supplement to the Fire Hazards Analysis for St. Lucie Unit 1. The discussion on radiant energy shields and fire extinguishers were not associated with any discussion of the K.2 Exemption and are incorrectly included in the K.2 Exemption review within Attachment K of the LAR. The September 16, 1983 letter is cited as a reference to the licensing action because the Fire Hazard Analysis for Fire Area K Section 6a and a brief statement in Section 6b of the letter addresses the details of the oil collection system (Section 6a) and a summary statement which identified that the RCP Oil Collection System design was amended by exemption request K.2 (Section 6b). The radiant energy shields discussed in Section 6b and the fire extinguishers mentioned in Section 6c of the letter are related to the overall Fire Hazard Analysis discussion for Fire Area K and were not identified with any of the discussion associated with the RCP Oil Collection System Exemption K.2. The NRC approval of the RCP oil collection system in the SER dated February 21, 1985 (also summarized in Attachment K) does not cite fire extinguishers or radiant energy shields as the basis for approval. Because the basis for the approval request, per the NRC SER, does not cite fire extinguishers or radiant energy shields, the features are not required for "E-EEEE/LA Criteria" as indicated in LAR Section 4.8.1 A markup of LAR Attachment K is attached to this response.

b) The installation of portable extinguishers and/or hose stations are required by NFPA 805, Chapter 3, Sections 3.4.4 (firefighting equipment including hoses, nozzles, extinguishers), 3.6.1 (hose stations and standpipes) and 3.7 (fire extinguishers). Therefore, although licensing actions may credit fire extinguishers or hose stations as the basis for approval/acceptability, the features are not explicitly called out in Attachment C, Table C-2 as fire extinguishers and hose stations are already required by NFPA 805 Chapter 3.

c) The installation of portable extinguishers and/or hose stations are required by NFPA 805, Chapter 3, Sections 3.4.4 (firefighting equipment including hoses, nozzles, extinguishers), 3.6.1 (hose stations and standpipes) and 3.7 (fire extinguishers). Therefore, although licensing actions may credit fire extinguishers or hose stations as the basis for approval/acceptability, the features are not explicitly called out in Attachment C, Table C-2 as fire extinguishers and hose stations are already required by NFPA 805 Chapter 3.

d) The installation of portable extinguishers and/or hose stations are required by NFPA 805, Chapter 3, Sections 3.4.4 (firefighting equipment including hoses, nozzles, extinguishers), 3.6.1 (hose stations and standpipes) and 3.7 (fire extinguishers). Therefore, although licensing actions may credit fire extinguishers or hose stations as the basis for approval/acceptability, the features are not explicitly called out in Attachment C, Table C-2 as fire extinguishers and hose stations are already required by NFPA 805 Chapter 3.

Enclosure I to L-2014-056 Page 28 of 75 e) Fire Protection systems and features required by licensing actions are included in Attachment C, Table C-2. Table C-2 of the LAR is a summary table of the overall system and feature review. The NFPA 805 Required Fire Protection Systems and Features are detailed in a separate evaluation (PSL-FPER-13-004). PSL-FPER-13-004 documents the required systems and features that are presented in LAR Attachment C, Table C-2 that are required to support various source documents. Source documents include the licensing action review (PSL-FPER-1 1-006), the existing engineering evaluation review (PSL-FPER-1 1-007), the fire risk evaluations (PSL-FPER-11-001), and other transition documents. The evaluation PSL-FPER-13-004 includes a detailed table in Attachment A which identifies the required systems and features.

The detailed table identifies the specific reference from which each feature is credited. Therefore, when reading PSL-FPER-13-004 Attachment A, each feature associated with PSL-FPER-11-006 is a system or feature required by a licensing action. In addition, the basis for each licensing action is validated in Attachment 2 of the Transition Review of Existing Licensing Actions (PSL-FPER-11-006). Attachment 2 is the field validation and controlled documents which confirm the bases for acceptability. The attachment identifies each licensing basis and provides the reference that validates the basis. All Fire Protection systems and features identified as bases for acceptance of the licensing actions are captured in LAR Attachment C, Table C-2.

PSL FPE RAI 08 - Mark-up Enclosure I to L-2014-056 Page 29 of 75 Attachment K Existing Licensing Action Transition Licensing Action 1-LA-03-19850221, Unit 1 Appendix R Exemption K.2 Because the Reactor Coolant Pump Oil Collection System is Not Capable of Collecting Lube Oil From All Four of the Reactor Coolant Pump Lube Oil Systems (111.0 Criteria)

Evaluation oil from an entire Reactor Coolant Pump Lube Oil System (190 gallons) and a generous allowance for reasonable leakage from the remaining systems. This is the basis of Exemption Request K2 from Section 111.0 of Appendix R to 10CFR50.

Shutdown equipment in Fire Area K required in the event of a fire consists of Safety A and B train instrumentation for the following functions: Pressurizer pressure and level, Steam Generator pressure and level and Reactor Coolant System temperature; Safety A and B train system valve; related cables and controls.

IDelete Fi, al, approxi

.00Btu/Sq ft, which is considered a low combustible load. The mao obustible load is attributable to cable nga h eerto ra, elevation 23.00' between or afeshudon fom fre n te enetration Area. The cables are

- 3 974 qulifed o ar prvidd with a fire retardant protective when ignite and burn until completely consumed, has no adverse effect on redundant cable or equipment required for safe shutdow t of a fire.

e other concentrated combustibles are evaluated individually:

Items 1-4 (Reactor Coolant Pump Lube Oil Systems)

Delete An oil collection system has been provided in accordance with Appendix R to 10CFR50, Section 111-0, as amended by exemption request K-2 submitted April 25, 1983 (L-83-261) and August 24, 1983 (L-83-453)."

I~c.

C erating:

Fir AraKF n

e ctor e

and therýmadetýector wcal an'd ýin the ýControl Room. Th~ereare 1 8 a ccess ible fire Iextinguis he rs in the fire area*

emp*

oy te tra ined fi re brigade,' utilizi ng the po rtablýe fire ex nlimiting the consequences of any I *r~-efe. Stereacto shutdown capabiltieswl not be adversel affected in the evento a fr e.

Document ID 1985-02-21-lb [SER Sections 15.0 thru 15.4] - NRC-FPL, Exemption Requests for St Lucie Plant, Unit No. 1, 10 CFR Part 50, Appendix R Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, dated February 21, 1985 Evaluation This NRC SER granted the exemption for the oil collection system. It stated:

"15.0 Oil Collection System for Reactor Coolant Pumps 15.1 Exemption Requested An exemption is requested from Section 111.0 to the extent it requires an oil collection tank sized to hold the lube oil inventory of all three Reactor Coolant Pump (RCP) motors.

15.2 Discussion The unit has four reactor coolant pumps with an oil collection system that drains to a vented closed collection tank. The quantity of lubricating oil in each pump is 190 gallons. The capacity of the oil collection tank is 225 gallons. The components have been designed so that they are capable of withstanding a safe shutdown earthquake (SSE).

Fire Safety Analysis Data Manager (4.129)

FPL - St. Lucie Run: 03/1812013 01:20 Page: 6 of 97

Enclosure I to L-2014-056 Page 30 of 75 PSL FPE RAI 09 The compliance statement for LAR Attachment A, Table B-I, Section 3.4.1 (c) is "complies".

NFPA 805, Section 3.4.1 (c) specifically requires the fire brigade leader and two brigade members to have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria. The compliance basis for this element states that the brigade leader and at least two brigade members have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance; but does not specify the details of the training and knowledge of these members. Describe how the requirements of NFPA 805 Section 3.4.1 (c) are met with regard to training and knowledge of the leader and at least two members of each fire brigade.

An approach acceptable to the staff for meeting this training and knowledge requirement is provided in Regulatory Guide (RG) 1.189, "Fire Protection for Nuclear Power Plants," Rev. 2, Section 1.6.4.1, Qualifications:

"The brigade leader and at least two brigade members should have sufficient training in or knowledge of plant systems to understand the effects of fire and fire suppressants on safe-shutdown capability. The brigade leader should be competent to assess the potential safety consequences of a fire and advise control room personnel. Such competence by the brigade leader may be evidenced by possession of an operator's license or equivalent knowledge of plant systems."

RESPONSE

PSL requires the brigade leader and at least two brigade members have sufficient training in or knowledge of plant safety related systems to understand the effects of fire and fire suppressants on safe shutdown capability. This is documented in 1800022, Section 8.7.6.B:

"Composition and Responsibility - A Shift Fire Brigade of at least five members shall be maintained on site at PSL at all times. The brigade leader and at least two brigade members shall have sufficient training in or knowledge of plant safety related systems to understand the effects of fire and fire suppressants on safe shutdown capability."

"The fire brigade shift shall not include the Shift Manager, nor the three other members of the minimum shift crew necessary for safety shutdown of the unit and any personnel required for essential functions during a fire emergency. Each shift will have a Fire Brigade Leader assigned. A list of all qualified Fire Brigade members is maintained in a current, up-to-date condition in LMS." LMS is St. Lucie's Learning Management System.

1800022, Section 8.7.5.C.1.c states:

"The Fire Brigade Leader course, as outlined in the Fire Protection Training Guide, is required for all Fire Brigade Leaders. The Fire Brigade Leader is required to have sufficient training or knowledge of plant safety related systems. This knowledge may be demonstrated by having an operator license or having completed Senior Nuclear Plant Operator Training. The Brigade Leader is trained and qualified in fire fighting practice to understand the effects of fire and fire suppressants on safe shutdown capability and is competent to assess the potential safety consequences of a fire, so as to advise Control Room personnel. Fire Brigade Leader training shall be completed prior to assignnment as Fire Brigade Leader."

Enclosure I to L-2014-056 Page 31 of 75 In addition, 1-OSP-100.27, Attachments 1 and 2, document the Operations Surveillances performed for the midnight and day shift. These surveillances include a specific note that the Fire Brigade Leader and at least two Fire Brigade Members have sufficient training in or knowledge of plant safety related systems to understand the effects of fire and fire suppressants on safe shutdown capability.

All Fire Brigade Members are Nuclear System Operators and receive full Systems (TR-AA-104 and TPD-PSL-E0I) and Fire Protection (0005729) training. They are required to attend requalification classes throughout the year on operation of the systems (TR-AA-104, TPD-PSL-E01, and 0005729) and fire event impacts to plant operations. Fire Brigade Members participate, as a minimum, quarterly in fire drill exercises (1800022 and 00005729). Fire Brigade Leader and two other members have knowledge of the effects of suppressants on safe shutdown equipment. This is achieved by being, at a minimum, qualified Senior Nuclear Plant Operators (SNPOs) that have received training on plant safe shutdown equipment as a part of accredited non-licensed training. The other brigade members are non-licensed operators but are not required to be qualified as a SNPO.

PSL FPE RAI 10 LAR Attachment S, Table S-1 and LAR Attachment C, Table C-I state that incipient detection is to be installed in the cable spreading room (CSR). Provide more details regarding system design features: NFPA code(s) of record, installation, acceptance testing, set-point control, alarm response procedures and training, and routine inspection, testing, and maintenance that will be implemented to credit this new incipient detection system. Describe whether the installation and credit that will be taken will be in accordance with the method elements, limitations, and criteria of NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," Chapter 13 and provide justification for any deviations.

RESPONSE

The PSL description in LAR Attachment S, Table S-1 and Attachment C, Table C-I identifies that a modification to install incipient detection to detect a fire well in advance of significant damage.

The VEWFD system is to be installed in the Annunciator Logic Cabinet in the Unit 2 Cable Spread Room.

Provide more details regarding system design features:

The design for the Incipient Detection VEWFD system to be installed in the Unit 2 Cable Spread Room is in the conceptual phase at this time. The current NFPA codes that are applicable to the design of the system(s) are NFPA 76, 2012 ed., Standard for the Fire Protection of Telecomimunications Facilities, and NFPA 72, 2013 ed., National Fire Alarm and Signaling Code.

The system design will be to install incipient detection inside the cabinet to detect fire well before significant damage occurs and prior to the occurrence of visible fire outside the cabinet. In this way damage to the cables located above (not related to the function of the cabinet) can be minimized. The design of the system and credit for the system is not to prevent fire in the cabinet or prevent fire damage to the contents of the cabinet.

Specifications to be issued for the system equipment and design procurement will identify the design features required, as well a need for the manufacturer's recommended acceptance testing requirements as well as routine inspection, testing and

Enclosure I to L-2014-056 Page 32 of 75 maintenance requirements. The system design, procurement, installation, acceptance test performance, and turnover to Operations will also include the development of procedures for inspection, testing, maintenance, alarm response, and control of set point controls. The development of procedures will include a review for identification of required training for Operations and Maintenance personnel.

The design will meet the guidance of FAQ 08-0046 as well as the NFPA Codes of Record and will be developed following the St. Lucie Engineering controls for modification development, implementation and turnover to the plant.

Where conflicts between the documents arise the more restrictive requirement will apply.

" Installation methods will be aligned with the NFPA Code of Record, and Manufacturer's Recommended Practices, which will include any parameters that may be associated with the specific requirements associated with the Underwriter's Laboratories Listing of Factory Mutual Approval, as applicable.

" Acceptance testing for the system will be developed as part of the modification package and will incorporate those requirements documented in NFPA 72 Code of Record and the Manufacturer's Recommended Practices.

" Set-point control will be administratively controlled by procedures to be written as a result of the Engineering process for development of modification packages.

" Alarm response procedures for Operations Personnel will be developed as part of the modification package process and a review for training needs will be performed as part of the normal process for development of procedures and modifications.

" Routine inspection, testing, and maintenance will be controlled and performed by procedures that will be implemented to credit the new incipient detection system. The procedures will utilize the manufacturer's criteria and NFPA 72, which will be developed as a result of the modification package.

The installation and credit that will be taken will be in accordance with the method elements, limitations, and criteria of NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," Chapter 13. No deviations from this guidance were taken. The system design will be to install incipient detection inside the cabinet to detect fire before visible fire occurs such that the potential for damage to cables located above and not related to the function of the cabinet can be minimized.

PSL FPE RAI 11 LAR Attachment S, Table S-1 identifies several proposed modifications to "protect" cables or signals (i.e., lA-MR-l, 1B-MR-01, 1C-MR-01 through 05, 2A-VM-01, and 21-MV-02). Provide a more definitive description of this protection. If barriers or rated cabling is to be used, describe the type of rated configurations that will be installed. If cable reroutes will be used, describe what type of new cable will be used and how that will be incorporated in the design.

RESPONSE

The above modifications all protect specific cables or eliminate them as secondary combustibles in specific fire zones. Modification I C-MR-0 I involves a modification to a ventilation fan motor to facilitate venting a fire zone to the outside atmosphere to reduce risk by mitigating the formation of a hot gas layer.

Enclosure I to L-2014-056 Page 33 of 75 Cables routed in conduit, before or after the modification, will be protected as necessary with a Thermo-Lag fire barrier system of the appropriate rating commensurate with the suppression and detection present in the affected fire zone. The conduit will receive the Thermo-Lag wrap in accordance with applicable requirements utilizing the standard design in use at St. Lucie for protected cables. Cables routed in cable trays will be spliced outside the zone and routed into a conduit in the affected zone and splice at the other end outside the zone. The existing cable that was in the cable tray will be abandoned in place. The routing for the spliced cable will be similar to the cable tray route and not a reroute to eliminate fire scenarios. There are no proposed modifications which involve the application of a rated fire barrier system to cable trays or fire rated cable. The new portion of the spliced cable route will utilize IEEE-383 rated cable.

LAR Attachment S Table S-2 Item 18 requires that after all modifications have been implemented and as-built, that the fire risk will be evaluated and compared to the baseline risk. Action will be taken if the results exceed the limits stated in the implementation item.

PSL PRA RAI 04b LAR Attachment V,Section V.3 indicates that a reduction in 98th-percentile HRR of NUREG/CR-6850 for transient fires is credited (i.e., from 317 kilowatts (kW) to 69 kW) for certain fire zones; however, no basis is provided to support this reduction. Address the following:

b. For these PAUs, address the location-specific attributes and considerations, plant administrative controls, the results of a review of records related to violations of the transient combustible and hot work controls, and any other key factors used to demonstrate the adequacy of a reduced HRR per the guidance endorsed by the NRC Letter from Joseph Giitter to Biff Bradley, Nuclear Energy Institute, "Recent Fire PRA Methods Review Panel Decisions And EPRI 1022993, Evaluation Of Peak Heat Release Rates In Electrical Cabinet Fires," June 21, 2012 (ADAMS Accession No. ML12171A583). 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 peak HRR greater than 69 kW.

RESPONSE

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 combustible limit 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 program to address future adherence to the requirements. A review of existing transient control experience is provided in the response to RAI PRA 4c. The results of post transition monitoring with respect to these controls will be the basis for implementation of appropriate corrective actions should violations occur. This review will also assess the configuration and potential size of any such violations and determine if the use of the 69 kW HRR criteria remains appropriate.

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

PSL is implementing additional administrative controls such as a fire watch (or other compensatory measures) for conditions in which transients are stored in these areas. These fire zones will be subject to strict combustible controls (designated as "No Storage") and so paper, cardboard, scrap wood, rags and other trash will not be allowed to accumulate in the

Enclosure I to L-2014-056 Page 34 of 75 area. An implementation item specifying these additional controls will be added to LAR Attachment S.

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

" Large combustible liquid fires are not expected in these fire zones because activities in these areas do not include maintenance of equipment containing large quantities of oil.

" 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 PSL 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 HRR 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 Btu/s) 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 because this configuration would provide both the ignition source (energized temporary cabling) and combustible (cable insulation). The motor configuration would resemble such a transient fire. A transient fire in an area of strict combustible controls, where only small amounts of contained trash are considered possible, is judged to be no larger than the 69 kW fire. Because only small quantities of trash in temporary containers can be expected, a 69k W peak heat release rate was determined to be appropriate to represent this quantity of combustibles. The 69kW heat release rate bounds the small trash can fires reported in NUREG/CR-6850 Appendix G.

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 PSL and industry wide with respect to the use of the nominal 317 kW (300 Btu/s) peak heat release rate transient fire.

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 PSL 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

Enclosure I to L-2014-056 Page 35 of 75 and/or restrictions." The use of the lower HRR in the areas with significantly increased transient controls is considered to be consistent with this guidance.

PSL PRA RAI 04c

c. Perform and document a review of past transient fire experience as well as a review of records 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), and discuss how this review informs the development of administrative controls credited, in part, to justify a HRR lower than 317 kW.

RESPONSE

A review of plant records over the past five years has been performed to identify violations of existing transient controls. The review was based on plant records/databases which would include tracking items associated with NRC inspections. This review has identified no instances of transient related fires or fires involving transient combustibles. Only two instances were identified of issues related to control of transient combustibles which potentially violated the transient combustible control limits in place. Based on this review, the St. Lucie plants have a good track record with respect to compliance with existing transient controls. The limited review findings, and the increased level of controls to be implemented, support the proposed use of a lower transient combustible HRR.

Zones crediting the reduced heat release rate, 69kW, transient fire are being subjected to a new and stricter set of transient controls than those previously in place at PSL. These zones are subjected to transient free controls. Fire zones crediting these administrative controls will be monitored going forward to ensure there are no violations. If any such violation is recorded corrective actions will be put in place to ensure that the transient free administrative controls are appropriately maintained. The use of such controls to credit a reduced transient HRR is in accordance with the Joseph Giitter letter to Biff Bradley (see response to RAI PRA 4b).

The review of plant records indicates a good compliance record with regards to existing transient controls. The imposition of these new controls is expected to see a similar level of compliance.

PSL PRA RAI 07 The ASME/ANS RA-Sa-2009 standard states that an ignition frequency greater than zero is to be assigned to every plant physical analysis unit (e.g., manholes). Considering guidance for the exclusion of transients provided in FAQ 12-0064 (e.g., manholes are welded shut, space too small to allow personnel access under any conditions, etc.), identify locations where transient fire scenarios are not postulated, and provide justification for their exclusion.

RESPONSE

Transient fires are postulated to occur in all fire zones containing Fire PRA equipment except manholes. Manhole covers are of substantial build that require special equipment to open. The manholes are considered to be a confined space that requires special procedures for entry, and are not subject to entry without a work order. Space within the manholes is also very limited.

Additionally, adding transient fires to the manhole zones would serve only to dilute the frequency from other fire zones. The co-location of a transient combustible and an ignition source within these manholes is not considered credible.

Enclosure I to L-2014-056 Page 36 of 75 PSL PRA RAI 10a Attachment V indicates that "bounding" abandonment conditional core damage probabilities (CCDPs) of 0.1, 0.2 and 1.0 were utilized rather than detailed human error analyses and that abandonment conditional large early release probabilities (CLERPs) were obtained by multiplying a scenario's corresponding abandonment CCDP by the ratio of the FRANC-calculated CLERP to CCDP; however, there appears to be no justifiable basis to assume that these values are bounding.

Provide the following:

1. An identification of the fire areas/compartments that credit MCR abandonment due to loss of habitability and/or loss of control or function.

RESPONSE

The following Fire Areas credit MCR abandonment:

1B (Fire Zone 1_57, CSR) - Unit 1 cable spreading room for HGL scenarios (Due to loss of control or function only, CCDP of 1.0 is assumed) 1 F (Fire Zone 1 _70, MCR) - Unit 1 control room for a percentage of all fires in zone (Due to loss of habitability only, for the portion of fires that are not suppressed prior to loss of habitability) 2B (Fire Zone 2_52, CSR) - Unit 2 cable spreading room for HGL scenarios (Due to loss of control or function only, CCDP of 1.0 is assumed) 2F (Fire Zone 2421, MCR) - Unit 2 control room for a percentage of all fires in zone (Due to loss of habitability only, for the portion of fires that are not suppressed prior to loss of habitability)

PSL PRA RAI 10b Attachment V indicates that "bounding" abandonment conditional core damage probabilities (CCDPs) of 0.1, 0.2 and 1.0 were utilized rather than detailed human error analyses and that abandonment conditional large early release probabilities (CLERPs) were obtained by multiplying a scenario's corresponding abandonment CCDP by the ratio of the FRANC-calculated CLERP to CCDP; however, there appears to be no justifiable basis to assume that these values are bounding.

Provide the following:

b. For those areas in which abandonment is credited on loss of control or function, a description of the criteria used for making the determination to abandon, and clarify how these criteria will be addressed by the going-forward fire procedures. In particular, justify the assumption that MCR abandonment due to fires in the cable spreading room is only assumed to occur for HGL scenarios

RESPONSE

The St. Lucie Fire PRA does not credit/require control room abandonment due to loss of control or function. MCR abandonment is assumed/required only for fires in the MCR for both Unit 1 and Unit 2 which cause habitability criteria to be exceeded. For all fires that do not impact habitability the limited fire damage and the ability to maintain command and control from the main control room will result in a lower risk than would MCR abandonment, therefore, no MCR abandonment is required for loss of control or function. The quantification of the cable spreading room fire

Enclosure I to L-2014-056 Page 37 of 75 scenarios is based on no credit for control from the hot shutdown control panel. The cable spreading room (CSR) HGL scenario, which impacts all cables in the CSR and thereby the majority of MCR functions is assumed to have a CCDP of 1.0. Although an attempt will likely be made to shut down from the hot shutdown control panel for such a fire, the Fire PRA conservatively assumes such a fire will result in core damage. The only reason to abandon the MCR for fires outside of the MCR would be a loss of functionality. There are no individual scenarios that would lead to a plant state that would make it more advantageous to attempt shutdown from the hot shutdown control panel than at the MCR. As such the operators are not expected to leave the MCR, where they have the greatest capability to reach safe shutdown.

It should be noted that the analysis for the CSR was done by systematically creating scenarios throughout the CSR. For each of these scenarios the percentage of fires that create a HGL is calculated and the sum of the HGL contribution for all scenarios is the factor that is applied to the HGL scenario. Using this methodology inherently means that every scenario contributes to MCR abandonment, however for the percentage of each individual scenario that does not lead to a HGL, enough functionality would remain from the MCR that it would not warrant abandonment. The remaining functionality is defined by the CCDP of the fire scenario which is based on the extent of damage of the fire scenarios that do not result in a HGL.

PSL PRA RAI 10d Attachment V indicates that "bounding" abandonment conditional core damage probabilities (CCDPs) of 0.1, 0.2 and 1.0 were utilized rather than detailed human error analyses and that abandonment conditional large early release probabilities (CLERPs) were obtained by multiplying a scenario's corresponding abandonment CCDP by the ratio of the FRANC-calculated CLERP to CCDP; however, there appears to be no justifiable basis to assume that these values are bounding.

Provide the following:

d. Explanation of the timing considerations (i.e., total time available, time until cues are reached, manipulation time, and time for decision-making) made to characterize scenarios in Part (c).

Include in the explanation the basis for any assumptions made about timing.

RESPONSE

MCR actions are assumed to fail at time zero such that no credit is given for these actions regardless of how much time is available prior to abandonment. This is a very conservative approach as the operators would not exit the control room at the start of a fire. No special timing considerations were made for actions external to the MCR which would be taken regardless of whether or not the operators exited the MCR. These actions are subject to the timing as outlined by the HRA analysis or the safe shutdown analysis manual action feasibility evaluation. All screening actions that are currently used as surrogates in the Fire PRA will either be removed or replaced with a detailed HFE in accordance with NUREG-1921. Timing considerations due to MCR abandonment will be taken into account as appropriate. See responses to PRA RAIs 01.d, 01.1 and 01.o (to be submitted with the 120 day RAIs).

Enclosure I to L-2014-056 Page 38 of 75 PSL PRA RAI 10e Attachment V indicates that "bounding" abandonment conditional core damage probabilities (CCDPs) of 0.1, 0.2 and 1.0 were utilized rather than detailed human error analyses and that abandonment conditional large early release probabilities (CLERPs) were obtained by multiplying a scenario's corresponding abandonment CCDP by the ratio of the FRANC-calculated CLERP to CCDP; however, there appears to be no justifiable basis to assume that these values are bounding.

Provide the following:

e. Discussion of how the probability associated with failure to transfer control to the alternate shutdown panel is taken into account in Part (c).

RESPONSE

The St Lucie Fire PRA analysis calculates a CCDP and CLERP for MCR abandonment based on components affected by a particular MCR fire and assumed failure of all in control room actions.

The analysis then conservatively increases these values to account for additional risk associated with the transfer of control to the hot shutdown control panel. This risk is appropriately distributed to scenarios based on how severe those scenarios are (as measured by their quantified CCDP). The minimum risk has been set to a CCDP of 0.1, for scenarios with a calculated CCDP of< 1E-03.

For potentially challenging fires a 0.2 has been used, and for the most severe fires core damage is assumed (CCDP = 1.0). Additional detail is provided in the response to part c of this question.

PSL PRA RAI 10f

f. Description of how the feasibility of the operator actions supporting the alternate shutdown pathway was considered by the scenario characterization performed in Part (c).

RESPONSE

For MCR abandonment, ex-control room actions are credited the same as if control was maintained at the MCR. Actions taken at the hot shutdown control panel (HSCP) are subject to the time limitations of their associated components/in-control room HEPs. All screening HFEs will be either removed from the analysis or replaced with a detailed HFE in accordance with NUREG-1921, to be addressed in response to RAIs PRA 01.d, 01.1 and 01.o. Feasibility of actions at the ASP is consistent with the feasibility of the associated in-control room operator actions because the timing associated with control room abandonment is not expected to have a significant impact on the operator response once control is transferred to the ASP. As such, the additional risk associated with the transfer of control is bound by the increase in the CCDPs and CLERPs from their nominal in-control room value to a higher control room abandonment value as specified in Attachment V, page V-5, of the PSL NFPA 805 LAR submittal. Scenarios with a very low calculated CCDP (< 1 E-03) have a very high probability of success in transferring control to the ASP. The CCDP for these scenarios has been increased to 0.1. Actions with a medium level of risk

(< 0.1 and > 1E-03) have been assigned a value twice as great, 0.2, as those with a low risk. The most risk significant scenarios (calculated CCDP > 0.1) are conservatively assumed to have a CCDP of 1.0. Thus for the most complicated abandonment scenarios no credit is given for actions taken at the ASP.

In the process of responding to other RAIs, PRA 01.d, 01.1 and 01.o (to be submitted with the 120 day RAIs)., related to issues associated with Fire HRA, it is anticipated that the approach described above may be altered with respect to the level of credit taken for actions at the ASP and the need to develop specific HEPs for such actions.

Enclosure I to L-2014-056 Page 39 of 75 PSL PRA RAI 12 A review of LAR Attachment W, Table W-7 indicates that for Fire Areas 2J and 2M, a value of "O.OOE+00" is reported for both ACDF and ALERF. The risk summaries provided for Fire Areas 2J and 2M in LAR Attachment C, Table C-I state that "[a] negative delta risk results from the summation of cutsets for the system/components made available in the compliant case [that]

exceed the risk associated with the failure of the system," further noting that "this is typically due to conservative estimates of risk associated with operator action dependencies". It is further stated that delta risk values for Fire Areas 2J and 2M are set to zero to preclude the resulting negative delta risk from offsetting the risk associated with other fire areas or scenarios. The risk summary provided for Fire Area IC contains a similar discussion; however, non-zero values are reported to give credit to an identified risk-reduction modification. It is unclear how negative delta risk values could be obtained, particularly noting that Fire Areas 2J and 2M do not appear to credit risk-reduction modifications, and to what extent the ACDF and ALERF values for other fire areas reported in Tables W-6 and W-7 are systematically underestimated. Explain this modeling anomaly, and evaluate its impact on risk results.

RESPONSE

This is not a modeling anomaly. A negative delta risk can be obtained when risk associated with a Compliant case exceeds the risk of the respective Variant case due to conservative estimates of associated human failure events (HFEs) and associated joint human error probabilities (HEPs) that could differ due to availability of equipment in the Compliant case (and associated HFEs) that is not available in the Variant case with a high level of dependency between the associated HEPs. In order to preclude such a negative delta risk from offsetting positive delta risk in other fire areas, the negative delta risk is set to zero.

Typically the conservative estimates of the risk associated with operator action dependencies could result in the failure to credit the Compliant system due to being totally dependent on another HEP related to a system which is failed. This dependency alone results in equal Compliant and Variant risk with the additional risk associated with non-operator action failures of the Compliant system further increasing the risk of the Compliant case and resulting in a negative delta. HRA dependency analysis relies on generic treatment of human failures in terms of the order of events and the level of dependencies between them that are considered conservative at best, and could lead to numerical results such as that described above.

The following simplified example illustrates this condition. Assume a condition whose cutsets include the following operator actions; A (1E-02), B (2E-03) and C (3E-02).

Compliant Cutset:

INIT COMP A

B C

Zi Variant Cutset:

INIT COMP A

C Z2 Assuming B and C have a high dependency on A in the Compliant case, the respective joint human failure (combination event) of A,B,C is ZI which is equal to (1E-02 x high dependency probably for B given A is successful of 0.5 x high dependency probability of C given A and B are successful of 1.0 = 5E-03). For the same condition (Variant case) in which hardware failure negates the need to have operator action B, then the associated cutsets will include only operator actions A and C, i.e., dependency on other operator actions in the scenario is either reduced or eliminated, thus the joint human failure (Z2) in this case is equal to (1 E-02 x medium dependency of C given A is successful of 1.5E-02 = 1.5E-04). Because the only difference in the cutsets is the difference in

Enclosure I to L-2014-056 Page 40 of 75 joint human failures (ZI of 5E-03 in compliant case vs. Z2 of 1.5E-03 in Variant case), a comparison of cutsets would indicate that the Compliant case would have higher risk than Variant case, which would produce a negative delta. In these instances, the difference was set to zero.

An example of a scenario in Fire Area 2M with a negative delta, 2_161 - F03 PTB "Variant" vs.

FRE "Compliant" is given below. For this example the first 50 cutsets for the Variant and Compliant cases were compared. These cutsets accounted for 75% of the total scenario risk. The increase in risk for the Compliant case is driven by the HFE JHFPSDCS - which is a failure to implement shutdown cooling (S LOCA). This is not available in the Variant case due to fire failures in this area but is in the Compliant case. This in itself does not increase the risk but it causes a more conservative HEP combination value to be injected into the cutset which leads to an overall higher risk. Of the 50 cutsets compared, a total of three differed between the Variant and Compliant case. These are as shown below:

CUT SET CASE CCDP INITIATOR TERM 1 TERM 2 TERM 3 TERM 4 TERM 5 TERM 6 TERM 7 CHFPRCP CTM2CCW NHFPMAN NLCD2AM COMBINA 1

Variant 1.76E-07 %ZZFIREU2 TRP HXB UALR 525 TION_1437 CHFPRCP CTM2CCW JHFPSDC NHFPMAN NLCD2AM COMBINA Compliant 3.96E-07 %ZZFIREU2 TRP HXB S

UALR 525 TION_1418 CHFPRCP NHFPMAN NLCD2AM QHFPICW QTM2BHD COMBINA 2

Variant 1.55E-07 %ZZFIREU2 TRP UALR 525 CCW R

TION_1439 CHFPRCP JHFPSDC NHFPMAN NLCD2AM QHFPICW QTM2BHD COMBINA Compliant 1.55E-07 %ZZFIREU2 TRP S

UALR 525 CCW R

TION_1420 3

Variant 9.11E-08 %ZZFIREU2 CHFPRCP CTM2CCW GMPF2PA TRP HXB REC CHFPRCP CTM2CCW GMPF2PA JHFPSDC COMBINA Compliant 9.10E-08 %ZZFIREU2 TRP HXB REC S

TION_1411 As seen above cutset 1 sees an increase to the CCDP of 2.2E-07, cutset 2 remains unchanged and in cutset 3 there is a slight reduction to risk. The conservatism in the HRA combinations can lead to both an increase or decrease in risk which will likely offset each other in most cases.

Regardless, the overall delta CCDP for the example scenario is -2.20E-07, however, once this number is multiplied by the ignition frequency the scenario delta CDF is -3E-1 1 (based on a Variant case CDF of 3.26E-09 and an Compliant case CDF of 3.29E-09). The table below gives each of the basic events from the above cutsets, a brief description and there probabilities.

Enclosure I to L-2014-056 Page 41 of 75 BE DESCRIPTION PROBABILITY

%ZZFIREU2 FIRE RELATED INITIATING EVENT 1.OOE+00 CHFPRCPTRP OPERATOR FAILS TO TRIP RCPS LOSS OF CCW 1.42E-02 COMBINATION_1411 HEP dependency factor for CHFPRCPTRP,JHFPSDCS 2.38E+05 COMBINATION_1418 HEP dependency factor for CHFPRCPTRP,JHFPSDCS,NHFPMANUALR 6.80E+05 COMBINATION_1420 HEP dependency factor for CO_1 CHFPRCPTRP,QHFPICWCCW,JHFPSDCS,NHFPMANUALR 1.26E+10 COMBINATION_1437 HEP dependency factor for CHFPRCPTRP,NHFPMANUALR 1.27E+00 COMBINATION_1439 HEP dependency factor for CHFPRCPTRP,QHFPICWCCW, NHFPMANUALR 529E+04 CTM2CCWHXB CCW HX 2B IN TEST OR MAINTENANCE 5.58E-03 GMPF2PAREC HPSI PUMP 2A FAILS TO RUN DURING RECIRCULATION 1.17E-03 JHFPSDCS OPERATOR FAILS TO ESTABLISH SDC FOLLOWING SML LOCA 4.20E-06 NHFPMANUALR OPERATOR FAILS TO INITIATE SUMP RECIRC AFTER LOCA & AUTO 3.50E-01 SWITCHOVER FAILS NLCD2AM525 LOGIC CIRCUIT AM525 FAILS TO GENERATE SIGNAL 5.07E-03 QHFPICWCCW OPERATOR FAILS TO RESTORE ICW TO CCW HX WHEN LOST 5.40E-05 QTM2BHDR 2B ICW HDR IN TEST OR MAINTENANCE 2.18E-03 No credit was taken for the negative delta risk in Fire Areas 2J or 2M to offset a positive delta risk in another area. While it is possible that this is an issue that shows up in other fire areas and is being masked by a positive delta, no impact on the total delta would be expected due to the exceedingly small values involved. Negative deltas calculated are less than one percent of the total risk for the fire areas where they have been identified and typically are 3 to 4 orders of magnitude less on a scenario level. As such, no impact to the total plant delta CDF or delta PSL PRA RAI 13 The seismic CDF (9.19E-8/yr) reported in LAR Attachment W Table W-1 and used to estimate the total plant CDF appears to be low compared to the seismic CDF estimate (4.6E-5/yr) presented for in a memorandum from NRC staff dated September 2010 providing updated results for Generic Issue 199 (memo titled: "Safety/Risk Assessment Results for Generic Issue 199, Implication for Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United states on Existing Plants"). The total CDF, if higher CDF values for seismic events were used, exceeds RG 1.174 risk acceptance guidelines. As a result, provide the basis and corresponding technical justification for the seismic events CDF and LERF presented in the LAR. Additionally, considering deficiencies identified by F&O SF-A 1-01, provide further justification that seismic/fire interactions are adequately evaluated in light of the new seismic hazard data from the United States Geological Survey 2008.

Enclosure I to L-2014-056 Page 42 of 75

RESPONSE

The seismic CDF (9.19E-8/yr) reported in LAR Attachment W Table W-1 and used to estimate the total plant CDF is consistent with the same value that was presented in an earlier submittal associated with Extended Power Uprate (EPU) LAR (Letter L-2010-078 dated April 16, 2010 for Unit 1, Section 2.13, Table 2.13-14). Due to the low seismicity location of St. Lucie plant, the seismic hazard was analyzed as part of the Individual Plant Examination for External Events (IPEEE) using the "scaled-back" option that was offered by the NRC to FPL which included Structure, Systems, and Components (SSCs) screening based on seismic plant walkdown. The Unit 1 CDF estimated value reported in the IPEEE was lower than the value presented in EPU LAR. Further estimates of seismic CDF based on the EPU plant configuration indicated a slightly higher value that was reported as presented above. The following table summarizes the seismic risk impact as presented in EPU LARs (Letter L-2010-078 dated April 16, 2010, and Letter L-2011-021, dated February 25, 2011, for Unit 1 and Unit 2, respectively).

EPU Seismic Risk CDF, per year LERF, per year Unit 1 9.19E-08 1.33E-08 Unit 2 4.50E-09 6.50E-1 0 Regarding the CDF reference related to Generic Issue 199 (GI-199), FPL believes that the NRC referenced value is considered a conservative bounding value, based on the simplifying assumption used in the GI-199 calculations that the plant-level High Confidence of a Low Probability of Failure (HCLPF) value was equal to the Safe Shutdown Earthquake (SSE) (0.lg). That is in contrast to the best-estimate plant-specific value expected to be evaluated for the site, given the Florida region low seismicity. Per the Near Term Task Force (NTTF) Initiative 2.1, FPL plans to follow its recommendations which includes reevaluation of seismic hazard and comparison of plant-specific Ground Motion Response Spectra (GMRS), being developed by EPRI for St. Lucie (currently in progress and not issued yet), against the site's design basis SSE spectra, then revise the seismic hazard analysis if warranted when the plant-specific GMRS is greater than SSE.

The current seismic-induced fire impact, as presented in the LAR, is valid and requires no change. Should an update to seismic-induced fire impact be warranted in response to NTFF Initiative 3.0, such update will be developed in accordance with the respective ruling.

PSL PRA RAI 15a Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting an FPP consistent with NFPA-805. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

Enclosure I to L-2014-056 Page 43 of 75 Identify any changes made to the IEPRA or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in ASME/ANS-RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency for Nuclear Power Plant Applications," as endorsed by RG 1.200. Also, address the following:

a. Clarify why the SRs in LAR Attachment V Table V-I were found to be not applicable (N/A) by the peer review team. Identify what the supporting requirements are not applicable to, for example, the type of reactor. If there are specific and multiple reasons that the supporting requirements are determined to be N/A, identify the reason for each SR, grouping them as appropriate. In general, SRs that are not deemed applicable to an application require application specific justification; so, all reasons provided should include a discussion of the NFPA 805 application characteristics.

RESPONSE

A review of Table V-I in the PSL LAR, as modified by the PSL LAR Supplement dated June 14, 2013, for all SRs listed as N/A by the peer review has been completed. The majority of all SRs will remain N/A. The SRs that are still N/A are listed below with a reason for their disposition as such:

SR Disposition Basis for N/A Disposition in Table V-I ES-B33 N/A No additional equipment credited that is not in the Fire Safe Shutdown analysis or in the Full Power Internal Events PRA No exclusions of fire-induced spurious operations of components were applied based on ES-B5 N/A the conditional probability of occurrence of the spurious operation condition subsequent to a fire CS-A7 N/A PSL does not use ungrounded three phase circuits. Therefore, this SR is not applicable QLS-A1 N/A PSL did not use qualitative screening QLS-A2 N/A PSL did not use qualitative screening QLS-A3 N/A PSL did not use qualitative screening QLS-A4 N/A PSL did not use qualitative screening QLS-B1 N/A PSL did not use qualitative screening QLS-B2 N/A PSL did not use qualitative screening QLS-B3 N/A PSL did not use qualitative screening No new initiating events were identified in PRM-B3 (referenced SR in PRM-B4 is erroneously identified as PRM-B2 in the ASME/ANS Standard RA-Sa-2009).

PRM-B6 N/A No new accident sequences were identified in PRM-B3 or B5 PRM-B8 N/A No new success criteria were identified in PRM B-7 No Fire specific PRM probability values were identified in PRM-B12 which require re-PRM-B13 N/A aayi analysis PRM-B15 N/A No new LERF accident progressions were identified in PRM-B14 3N/A No specific consideration of burnout (due to depletion of available fuel) was applied in the analysis

Enclosure I to L-2014-056 Page 44 of 75 SR Disposition Basis for N/A Disposition in Table V-1 The disposition of FSS-C7 is extensive and is being provided separately with the response to RAI PRA In (120 day RAI response)

FSS-C8 N/A Raceway fire wrap is not credited in the PSL Fire PRA FSS-E2 N/A PSL used generic fire modeling parameters for all scenarios No Fire PRA scenarios involving exposed structural steel required modeling per FSS-F1 requirements.

This SR is not applicable since no high hazard areas containing structural steel fire FSS-F3 N/A proofing (see SR FSS-F1) were identified. No Fire PRA scenarios involving exposed structural steel required modeling.

FSS-G5 N/A Active fire barriers were not credited in the multi-compartment analysis.

IGN-A2 N/A No non-nuclear power industry data utilized in the FPRA.

IGN-A3 N/A No engineering judgment based data was utilized in the FPRA.

QNS-Al N/A Quantitative screening not used at PSL QNS-B1 N/A Quantitative screening not used at PSL QNS-B2 N/A Quantitative screening not used at PSL QNS-C1 N/A Quantitative screening not used at PSL QNS-D1 N/A Quantitative screening not used at PSL QNS-D2 N/A Quantitative screening not used at PSL No "claim of nonapplicability of any of the referenced requirements in Part 2 beyond that FQ-F2 N/A already covered by the clarifications in this Part" were applied, so there was no need to provide documentation.

PSL PRA RAI 15b Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting an FPP consistent with NFPA-805. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

Identify any changes made to the IEPRA or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in ASME/ANS-RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency for Nuclear Power Plant Applications," as endorsed by RG 1.200. Also, address the following:

b. If any changes are characterized as a PRA upgrade, indicate if a focused-scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, as endorsed by Regulatory Guide 1.200, and describe any findings from that focused-scope peer

Enclosure I to L-2014-056 Page 45 of 75 review and the resolution of these findings. If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, describe what actions will be implemented to address this review deficiency.

RESPONSE

ASME/ANS-RA-Sa-2009 defines an upgrade to an existing PRA that would require a focused scope peer review as follows: "the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. This could include items such as new human error analysis methodology, new data update methods, new approaches to quantification or truncation, or new treatment of common cause failure." While there have been several updates to the St Lucie Fire PRA model, these updates are not considered to fall within the definition of an upgrade. All Fire PRA updates since the peer review have been built off of peer reviewed models (whether IEPRA or Fire PRA models). As such, they do not require a focused scope peer review per the definition given in ASME/ANS-RA-Sa-2009. For further details on specific updates see PSL RAI PRA 15c.

PSL PRA RAI 15c Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting an FPP consistent with NFPA-805. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

Identify any changes made to the IEPRA or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in ASME/ANS-RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency for Nuclear Power Plant Applications," as endorsed by RG 1.200. Also, address the following:

c. The NRC staff notes that, several "updates" made to the fire scenario selection and analysis after the peer review appear to qualify as PRA "upgrades", including, but not necessarily limited to:

" Replacement of a unreviewed analysis method (UAM), i.e., panel factors, with a target-distance-based and time-dependent methodology,

" Significant enhancements made to the HGL analysis (e.g., new fire modeling to address secondary combustibles),

" Incorporation of new cable spread methodology (i.e., FLASH-CAT),

" Significant enhancements made to the MCA, including transition from a two-tiered quantitative screening approach that screened all such scenarios as a very small fraction of the fire CDF at the time of the peer review to one that fully quantifies scenarios and identifies them as significant fire initiating event contributors to both CDF and LERF in LAR Attachment W, and

Enclosure I to L-2014-056 Page 46 of 75 Credit for automatic suppression systems when the original peer-reviewed analysis only credited manual suppression.

Provide justification for why each of these model updates are not considered PRA upgrades in accordance with the PRA Standard ASME/ANS-RA-Sa-2009, as endorsed by Regulatory Guide 1.200 (Rev. 2). If considered a PRA upgrade, provide the information requested by Part (b).

RESPONSE

The methodology utilized for the target-distance-based time to damage for severe and non-severe fires is simply an application of the generic treatments which has previously been peer reviewed. The data was further utilized by breaking down the heat release rate probability distribution (HRR) to each constituent bin, however, this data is taken directly from the generic treatments using the NUREG/CR-6850 HRR and bin definitions. The use of the time dependent damage accrual methodology is an extrapolation of the data provided in Appendix H of NUREG/CR-6850 using a formulation used for evaluation of cable degradation for Equipment Qualification analysis. As such this is not an upgrade to the PRA, but an update to a peer reviewed methodology already being applied at PSL and does not require a focused scope peer review.

The methodology for calculating HGL has not changed due to the addition of secondary combustibles. Additional heat sources have been added to scenarios where applicable resulting in a change in the total HRR for individual scenarios but not altering the methodology which is defined in the generic fire modeling treatments. The same CFAST methodology used to calculate time to HGL is utilized and the source data has remained unchanged. As such the addition of secondary combustibles is an update to the existing Fire PRA model and does not require a focused scope peer review.

The cable spread rate used at PSL comes directly from Appendix R in NUREG/CR-6850. The spread rate and PSL specific configurations were used as inputs into a FLASH-CAT model to determine the HRR for individual configurations. FLASH-CAT is an NRC validated model addressed in NUREG/CR-7010 which applies the NUREG/CR-6850 methodology along with insights from cable testing to define the means for which to combine HRRs from an ignition source with those from cable tray combustibles. The HRRs calculated in FLASH-CAT are fed into CFAST to calculate scenario specific times to HGL. The use of FLASH-CAT to get PSL specific data is an update to the existing methodologies used and applying an NRC validated data methodology using NUREG/CR-6850 criteria. As such a focused scope peer review is not necessary for this item.

MCA calculations are based on the HGL calculation for the initiating zone. As noted the HGL calculation comes from the generic treatments which is a peer reviewed methodology. The same methodology is still applied with the screening step excluded and all scenarios retained. As such this is an update to the PSL Fire PRA which addresses peer review F&Os and does not require a focused scope peer review.

Credits for both automatic suppression as well as manual suppression are used in accordance with NUREG/CR-6850 specified methodology. The method used was part of the peer reviewed Fire PRA methodology. F&Os relative to this issue were addressed in the methodology update.

Further discussion of the dependency of the automatic and manual suppression/detection systems will be provided in response to RAI PRA 01m.

Enclosure I to L-2014-056 Page 47 of 75 There are no other known updates to the PSL Fire PRA that would be a new methodology warranting a focused scope peer review.

PSL PROG RAI 01 Based on the NRC staff s review of the LAR and associated documentation, it was determined that the LAR did not provide the information needed for the NRC staff to evaluate what changes will be made to the FPP to incorporate the requirements of NFPA 805, Sections 2.7.3.4 and 3.3.1.1.

Describe the changes that are planned to the FPP as part of the NFPA 805 transition process that are specifically associated with training. In addition, describe the positions where such training would be necessary.

RESPONSE

All current fire protection program documents such as AP-1800022, Fire Protection Plan, will be updated and a new fire protection program design basis document will be issued that contains all the aspects of the fire protection program. This will result in multiple changes in implementing documents for the fire protection program. The documents to be revised will be determined by the process that revises the primary fire protection program documents and issues the design basis document. All new procedures or procedure revisions require an assessment of training impact and completion of training prior to issuance per procedure AD-AA-100-1004, Preparation Revision, Review and Approval of Site-Specific Procedures. The type of training and the recipients of such training are identified as part of that process. In addition to this, due to the complex nature of this fire protection licensing basis change, a change management plan has been developed to support the transition to NFPA 805. That change management plan consists of two parts, the first is a fleet level plan (AR 1623193) and the second is a site level plan (AR 1667266). These plans also require review of training requirements. Positions where such training is necessary are identified in the change management plans. Positions include Operations Personnel, Engineers (Fire Protection, Safe Shutdown, and PRA), Schedulers and Planners, and Fire Brigade Members.

Based on the current development of the transition process the following training is anticipated.

This process is currently being implemented at Duane Arnold through mostly fleet level procedures. The procedures, qualification cards, and training developed for Duane Arnold will be utilized to the extent possible at St. Lucie.

Enclosure I to L-2014-056 Page 48 of 75 Training Module Task 1

NFPA 805 Overview Awareness for All Licensee Personnel CBT 2

NSCAINPO Analyses Qualification Card 3

Fire PRA Qualification Card 4

NFPA 805 Process Changes for Site Personnel 5

Applicability Screening of Changes for OPS/MTC/WC etc.

6 Applicability Screening of Changes for ENG 7

Fire Protection Modification Review Engineering Qualification Card 8

Monitoring Program Data Input for OpsIWC/MTC/CAP 9

Monitoring Program Engineering Data Review Qualification Card 10 Change Evaluation Qualification Card Development PSL PROG RAI 02 NFPA 805, Section 2. 7.3.4, "Qualification of Users," states that cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.

Describe the qualification program that will support the NFPA 805 change evaluation process.

Include positions that will be trained and how the training will be implemented (e.g., classroom, computer-based, reading program).

RESPONSE

The response to PSL RAI Programmatic 01 outlined changes to the fire protection program and provided identification of how training is to be handled under the transition. The change evaluation process will incorporate the guidance in FAQ 12-0061 when it is approved. This guidance will be used to revise existing procedures and/or create new procedures for the change evaluation process.

As stated in the response to PSL RAI Programmatic 01, all new procedures or procedure revisions require an assessment of training impact and completion of training prior to issuance per procedure AD-AA-100-1004, Preparation Revision, Review and Approval of Site-Specific Procedures. The type of training and the recipients of such training are identified as part of that process. In addition to this, due to the complex nature of this fire protection licensing basis change, a change management plan has been developed to support the transition to NFPA 805. That change management plan consists of two parts, the first is a fleet level plan (AR 1623193) and the second is a site level plan (AR 1667266). These plans also require review of training requirements.

Based on the current development of the transition process, the following training is anticipated for the change evaluation process (screening is included because that is the start of the process).

Positions requiring training primarily include Engineering personnel (Fire Protection, Safe Shutdown, and PRA). The change management plan states that the Training Department will define a graded approach to be used and a resource plan will be written to identify who, how, and when training supporting procedure changes will be implemented. This process is currently being

Enclosure I to L-2014-056 Page 49 of 75 implemented at Duane Arnold through mostly fleet level procedures. The procedures, qualification cards, and training developed for Duane Arnold will be utilized to the extent possible at St. Lucie.

Module Applicability Screening of Changes for Operations/MaintenancelNork Control etc.

Applicability Screening of Changes for Engineering Fire Protection Modification Review Engineering Qualification Card Change Evaluation Qualification Card Development PSL SSA RAI 01 LAR Attachment G, Table G-1, "Recovery Actions and Activities Occurring at the Primary Control Station(s)," provides a list of primary control stations (PCSs) for Units 1 and 2. There are approximately 25 PCSs locations identified for Unit 1 and 32 for Unit 2. The LAR defines PCS actions that contain activities such as placing specified transfer switches in their "isolate" position, selected tripping of breakers, selected removal of power to provide independence, and electrical separation from potentially damaged circuits. Many of these appear to be operator actions in locations separate from the remote shutdown panels, which may not meet the definition of PCS in RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing light-Water Nuclear Power Plants," Revision 2. Justify each of these locations as having met the definition of PCS as defined in RG 1.205, Revision 2.

RESPONSE

PSL-FPER-1 1-013 Revision 0, Definition of Primaty Control Station for use with Transition to NFPA 805 provided the basis for defining the Primary Control Stations at St. Lucie in accordance with Regulatory Guide 1.205 Revision 1 December 2009 and FAQ 07-0030, Revision 5, Establishing Recovery Actions (ML103090602). It states in part that, in addition to the definition presented in the regulatory guide, actions taken in the process of abandoning a control room and transferring to a primary control station may meet the definition of a recovery action, but the additional risk of their use does not need to be evaluated to demonstrate compliance with NFPA 805 Section 4.2.4. In accordance with the information contained in FAQ-07-0030 and accepted by the NRC, actions taken to enable the PCS are considered equivalent to being performed at the PCS and are not considered recovery actions requiring the evaluation of additional risk. To meet this definition, the action must meet the following as stated in FAQ-07-0030:

Actions that are necessary to activate or switch over to a primary control station(s) may be considered as taking place at primary control station(s) under the following conditions:

The actions are limited to those necessary to activate, turn on, power up, transfer control or indication, or otherwise enable the primary control station(s) and make it capable of fulfilling its intended function following a fire. These actions must be related to the alternative/dedicated shutdown function and should take place in locations common to panels that perform the transfer of control. For example, switches that disable equipment in order to allow the alternative/dedicated shutdown location to function would be included as part of the primary control station. However, these actions must be in the same location(s) (panel or the local vicinity surrounding the panel) as the normal/isolation

Enclosure I to L-2014-056 Page 50 of 75 switches and may include de-energization of selected equipment and/or circuits (if such actions are similar to the use of isolation switches). This does not include additional actions in the plant that, while necessary to achieve the NSPC, are not part of enabling the primary control station(s) (e.g., controlling inventory by locally controlling valve(s)).

" The actions are feasible and take place in sufficient time to allow the primary control station(s) to be used to perform the intended functions. The intended functions are defined as the original design criteria for the alternative/dedicated shutdown location(s) as provided in Generic Letter 86-10, Enclosure 2, Question 5.3.10 and Section 5.4.1 of RG 1.189, Revision 2.

" The switches or other equipment being operated to transfer control to the primary control station(s) are free from fire damage and the operators are able to travel from the main control room to the transfer location(s) and on to the primary control station(s) without being impeded by the fire.

To determine which actions (and locations) meet this definition, a review of the Safe Shutdown Analysis (SSA) and Control Room Inaccessibility Procedure for each unit was performed to make a determination if the action was a transfer action to enable the Hot Shutdown Control Panel (HSCP), or a one-time simple action that disables equipment to allow the success of the HSCP by either a transfer switch or simple removal of power.

The following provides the justification that the PCS actions listed in LAR Attachment G under the definition of the primary control station aligns with the guidance in RG 1.205 Revision 1 and FAQ 07-0030. There is a separate table for Unit 1 and Unit 2 and the table is grouped by transfer panel location. However, each location may not represent a single panel. For example, the A Switchgear Transfer Panels consist of a transfer panel (instrumentation), switches located at the breaker (4 kV breakers), or switches located in a separate compartment (480V Load Center and MCC). This is consistent with RG 1.205 guidance, which refers to "panel(s)" which indicates that transfer switches in more than one panel is acceptable. The RAI discusses approximately 25 locations on Unit 1 and 32 locations on Unit 2. However, the data presented in LAR Attachment G Table G-l was listed by action and not location. St. Lucie uses a distributed transfer scheme with instrumentation in transfer panels and switchgear/load centers/MCCs with transfer switches at the switchgear/load center/MCC. This equipment is in close proximity to the instrument transfer panels. Each transfer scheme is train related with all the A train transfers in the A switchgear room and all the B transfers in the B switchgear room, etc. While there are 25 actions on Unit 1 and 32 actions on Unit 2, these actions are performed in only 7 locations on Unit 1 and 8 locations on Unit 2. The majority of the locations are in close proximity to each other. For example, for a control room fire the majority of the transfer action take place in the A switchgear room, B switchgear room, and the cable spread room. Each of these rooms is adjacent to one another.

Enclosure I to L-2014-056 Page 51 of 75 Unit 1:

PCS Description from Transfer of Control Location (all or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 1)

Sound Powered Phone Isolation Box RAB 1-2 located in the A switchgear room.

Transfer of Control A Switchgear room Normal/Isolate switches transfer panels 1A Isolation Panel located in the A A Switchgear room Transfer of Control wicearom Transfer Switch switchgear room. Normal/Isolate switches transfer panels 480V MCC 1A6 Normal/Isolate switches A Switchgear room Transfer of Control wicearom Transfer Switch located in the A switchgear room transfer panels Simple breaker action in the 480V MCC 1A6 located in the A switchgear De-energization to A Switchgear room vicinity of the transfer panel.

room breakers 1-41311, 1-41316, and 1-allow HSCP to transfer panels Therefore, meets the definition in 41321 to the off position function RG 1.205 Revision 1 and FAQ 07-0030 1A2 480V Load Center Normal/Isolate Transfer of Control A Switchgear room Transfer Switch switches located in the A switchgear room transfer panels Simple breaker action in the 1A2 480V Load Center located in the A De-energization to A Switchgear room vicinity of the transfer panel.

switchgear room breakers 1-40207 and 1-allow HSCP to Therefore, meets the definition in 40210 to the open position function transfer panels RG 1.205 Revision 1 and FAQ 07-0030 480V MCC 1A5 Normal/isolate switches A Switchgear room Transfer of Control wtcearom Transfer Switch located in the A switchgear room transfer panels Simple breaker action in the 480V MCC 1A5 located in the A switchgear De-energization to A Switchgear room vicinity of the transfer panel.

room breakers 1-41202, 1-41272, 1-41254, allow HSCP to tchger room Therefore, meets the definition in 1-41260 to the off position function transfer panels RG 1.205 Revision 1 and FAQ 07-0030 1A3 4160V Switchgear Normal/Isolate Transfer of Control A Switchgear room Transfer Switch switches located in the A switchgear room transfer panels Simple breaker action in the 1A3 4160V Switchgear trip of breakers 1-De-energization to A Switchgear room vicinity of the transfer panel.

20204 and 1-20212 located in the A allow HSCP to transfer panels Therefore, meets the definition in switchgear room function RG 1.205 Revision 1 and FAQ 07-0030 1 B Isolation Panel located in the B B Switchgear roomSwitch switchgear room. Normal/Isolate switches transfer panels 480V MCC 1B6 Normal/Isolate Switches B Switchgear room Transfer of Control wicgarom Transfer Switch located in the B switchgear room transfer panels Communications Isolation Panel B-1609 B

located in the B switchgear room.

Transfer of Control Switchgear room Transfer Switch Normal/Isolate switches transfer panels 1B3 4160V Switchgear Normal/Isolate Transfer of Control B Switchgear room Transfer Switch switches located in the B switchgear room transfer panels

Enclosure I to L-2014-056 Page 52 of 75 PCS Description from Transfer of Control Location (all or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 1) 1B3 4160V Switchgear located in the B Energize equipment Simple breaker action in the switchgear room, close breakers 1-20404 to allow for HSCP vicinity of the transfer panel.

and 1-20410, and close breaker 1-20403 control (de-energize B Switchgear room Therefore, meets the definition in (control room fire) or trip 1-20403 (cable for a cable spread transfer panels RG 1.205 Revision 1 and FAQ spread room fire) room fire) 07-0030 1 B2 480V Load Center Normal Isolate B Switchgear room Transfer of Control Sicgarom Transfer Switch switches located in the B switchgear transfer panels 1 B2 480V Load Center located in the B Energize and de-Simple breaker action in the switchgear room breakers 1-40503, 1-energize equipment B Switchgear room vicinity of the transfer panel.

40505, and 1-40506 to the closed position to allow for HSCP transfer panels Therefore, meets the definition in with breaker 1-40507 to the open position control 07-0030 1 B5 480V MCC Norma/Isolate switches Transfer of Control B Switchgear room Transfer Switch located in the B switchgear room transfer panels 1 B5 480V MCC located in the B switchgear Simple breaker action in the room breakers 1-42016, 1-42017, 1-42018, De-energize B Switchgear room vicinity of the transfer panel.

8 1 5 a equipment to allow tchger room Therefore, meets the definition in 1-42068, 1-42035, and 1-42040 to the off for HSCP control transfer panels RG 1.205 Revision 1 and FAQ position0703 07-0030 WA1 6.9KV Switchgear located in the Turbine Building Turbine Building Switchgear Room. Local Transfer of Control Transfer Transfer Switch switch for breaker 1-30102 Switches/Panel WA1 6.9KV Switchgear located in the Simple breaker action in the Turbine Building Switchgear Room.

De-energization to Turbine Building vicinity of the transfer panel (fuse Breaker trip for breakers 1-30103, 1-allow HSCP to Transfer blocks designed to be pulled 30104, and 1-30105 (including pulling fuse function Switches/Panel without tools). Therefore, meets blocks) the definition in RG 1.205 Revision 1 and FAQ 07-0030 1A2 4 KV Switchgear located in the Turbine Building Turbine Building Switchgear Room Transfer of Control Transfer Transfer Switch Normal/Isolate switches Switches/Panel 1 B2 4 KV Switchgear located in the Turbine Building Turbine Building Switchgear Room.

Transfer of Control Transfer Transfer Switch Normal/Isolate'switches Switches/Panel 1 B1 6.9 KV Switchgear located in the Turbine Building Turbine Building Switchgear Room. Local Transfer of Control Transfer Transfer Switch switch for breaker 1-30202 Switches/Panel 1 B1 6.9 KV Switchgear located in the Simple breaker action in the Turbine Building Switchgear Room.

De-energization to Turbine Building vicinity of the transfer panel (fuse Breaker trip for breakers 1-30203, 1-allow HSCP to Transfer blocks designed to be pulled 30204, and 1-30205 (including pulling fuse function Switches/Panel the definition in RG 1.205 blocks)

Revision 1 and FAQ 07-0030 lAB Isolation Panel (control room fires C

only) located in the cable spread room.

Transfer of Control able Spread room Transfer Switch Normal/Isolate switches transfer panels

Enclosure I to L-2014-056 Page 53 of 75 PCS Description from Transfer of Control Location (all or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 1) lAB 480V MCC Normal/Isolate switches (for control room fires only) located in the Transfer of Control able Spread room Transfer Switch cable spread room transfer panels 1 B3 Pressurizer heater 1 B3 480V load center Normal/Isolate switches (control Transfer of Control Cable Spread room Transfer Switch room fire only) located in the cable spread transfer panels room 1A3 Pressurizer heater 1A3 480V load center Normal/Isolate switches (control Transfer of Control Cable Spread room Transfer Switch room fire only) located in the cable spread transfer panels room B PORV and RCS Gas Vent Isolation B Penetration room switches in the 1 B electrical penetration Transfer of Control transfer panel Transfer Switch room A PORV and RCS Gas Vent Isolation A Penetration room switches in the 1 A electrical penetration Transfer of Control transfer panel Transfer Switch room 4160V Switchgear lAB Normal/Isolate A

switches located in the AB Switchgear Transfer of Control aB Switchgear Room Transfer Switch Transfer Panels room Simple breaker action in the 4160V Switchgear lAB located in the AB De-energize AB Switchgear Room vicinity of the transfer panel.

Switchgear room trip breakers 1-20502 equipment to allow Transfer Panels Therefore, meets the definition in and 1-20503 for HSCP control RG 1.205 Revision 1 and FAQ 07-0030 lAB 480V Load Center Normal/Isolate AR Switchgear Room switches located in the AB switchgear Transfer of Control Transfer Panels Transfer Switch room 1 B Diesel Generator Isolate switches B EDG Transfer panel Transfer of Control (part of the B EDG Transfer Switch located in the 1B Diesel Generator roomPanel)

Enclosure I to L-2014-056 Page 54 of 75 Unit 2:

PCS Description from Transfer of Control Location (All or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 2)

MCC 2A6 Normal/Isolate switches located A Switchgear Room Transfer of Control SwitchgrarwRoom in the A switchgear room Transfer Panels Simple breaker action in the MCC 2A6 position breakers 2-41310 and De-energize A

vicinity of the transfer panel.

2-41319 to off located in the A switchgear equipment to allow for Switchgear Room Therefore, meets the definition in room HSCP control RG 1.205 Revision 1 and FAQ 07-0030 2A2 480V Load Center Normal/Isolate Transfer of Control A Switchgear Room Transfer Switch switches located in the A switchgear room Transfer Panels Simple breaker action in the 2A2 480V Load Center position breaker 2-De-energize A Switchgear Room vicinity of the transfer panel.

40212 to open located in the A switchgear equipment to allow for Transfer Panels Therefore, meets the definition in room HSCP control RG 1.205 Revision 1 and FAQ 07-0030 MCC 2A5 Normal/Isolate switches located Transfer of Control A Switchgear Room in the A switchgear room Transfer Panels Simple breaker action in the MCC 2A5 position breaker 2-41202 to off De-energize A

vicinity of the transfer panel.

located in the A switchgear room equipment to allow for Switchgear Room Therefore, meets the definition in locatedtrol Transfer Panels Teeoe et h

eiiini HSCP control RG 1.205 Revision 1 and FAQ 07-0030 2A3 4160V Switchgear Normal/Isolate Transfer of Control A Switchgear Room switches located in the A switchgear room Transfer Panels 2A3 4160V Switchgear position breakers 2-Simple breaker action in the 20201, 2-20203, and 2-20205 to trip De-energize and vicinity of the transfer panel (fuse including pulling fuse blocks, and position energize equipment A Switchgear Room blocks designed to be pulled breakers 2-20204, 2-20210, 2-20213, 2-to allow for HSCP Transfer Panels without tools). Therefore, meets 20206, and 2-20207 to close located in the control the definition in RG 1.205 A switchgear room Revision 1 and FAQ 07-0030 2A Transfer Panel Normal/isolate switches A Switchgear Room Transfer of Control TwitchgearwRoom located in the A switchgear room Transfer Panels LT-9012 transfer switch located in the A A Switchgear Room wicgrromTransfer of Control TasePnls Transfer Switch switchgear room Transfer Panels 2B3 4160V Switchgear Normal/Isolate B Switchgear Room switches located in the B switchgear Transfer Panels 2B3 4160V Switchgear located in the B De-energize switchgear room trip breakers 2-20405, 2-epene r B Switchgear Room 20407, and 2-20408 including pulling fuse HSCP control Transfer Panels Transfer Switch blocks 2B transfer panel Normal/Isolate switches B Switchgear Room located in the B switchgear room Transfer Panels 2B2 480V Load Center Normal/Isolate Transfer of Control B Switchgear Room Transfer Switch switches located in the B switchgear room Transfer Panels 2B5 480V Load Center Normal/Isolate B Switchgear Room Transfer of Control Transfrser Room switch located in the B switchgear room Transfer Panels TrnfrSic

Enclosure I to L-2014-056 Page 55 of 75 PCS Description from Transfer of Control Location (All or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 2)

Simple breaker action in the 235 480V Load Center located in the B De-energize B Switchgear Room vicinity of the transfer panel.

285equipment to allow for Transfer Panels Therefore, meets the definition in switchgear room breaker 2-4051 HSCP control RG 1.205 Revision 1 and FAQ 07-0030 Simple switch operation similar to transfer switch. No dependence on communication or other 289 MCC fan cooler switches to fast B Switchgear Room actions. Equivalent to simple located in the B switchgear room See Remarks Transfer Panels breaker operation to energize equipment to allow HSCP function. Therefore, meets the definition in RG 1.205 Revision 1 and FAQ 07-0030 2A 125 VDC buss breakers to on position Simple breaker action in the for breakers 2-60166, 2-60123, and 2-Energize HSCP In the vicinity of the B vicinity of the transfer panel.

60139 located in the 2A battery charger control Switchgear Room Therefore, meets the definition in 60oocTransfer Panels RG 1.205 Revision 1 and FAQ 07-room 0030 2AB transfer panel Normal/Isolate switches In the vicinity of the B Transfer of Control Switchgear Room Transfer Switch located in the 2AB battery charger roomTrnfrPel Transfer Panels Simple switch operation similar to transfer switch. No dependence on communication or other MCC 2A9 MCC containment fan cooler B Switchgear Room aker Equivalent to simple switches to fast located in the B switchgear See Remarks TransferwicheaR breaker operation to energize room equipment to allow HSCP function. Therefore, meets the definition in RG 1.205 Revision 1 and FAQ 07-0030 2A5 480V Load Center Normal/Isolate B Switchgear Room switch located in the B switchgear room Transfer Panels Communication isolation switches located Transfer of Control B Switchgear Room in the B switchgear room Transfer Panels MCC 286 Normal/Isolate switches located B Switchgear Room Transfer of Control TwitchgearwRoom in the B switchgear room Transfer Panels Simple breaker action in the MCC 2B6 position breaker 2-42118 to off De-energize B Switchgear Room vicinity of the transfer panel.

located in the B switchgear room equipment to allow for Transfer Panels Therefore, meets the definition in oHSCP control RG 1.205 Revision 1 and FAQ 07-0030 MCC 285 Normal/Isolate switches located B Switchgear Room Transfer Switch in the B switchgear room Transfer Panels MCC 285 position breakers -42004, 2-Simple breaker action in the 42012, 2-42033, 2-42036, 2-42037, and 2-De-energize B

vicinity of the transfer panel.

equimen toallo fo Swtchgar oom Therefore, meets the definition in 42052 to off located in the B switchgear eqCpmcntrol Transfer Panels RGe1.205 Revs ion 1

F roomHSCP control RG 1.205 Revision 1 and FAQ 07-room 0030

Enclosure I to L-2014-056 Page 56 of 75 PCS Description from Transfer of Control Location (All or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 2) 2C AFW pump room Normal/Isolate A

switches located in the 2C AFW pump Transfer of Control aFW Pump Area Transfer Switch Transfer Panels room 2A and 2B AFW pump room Normal/Isolate AFW Pump Area switches located in the 2A and 2B AFW Transfer of Control fTransfer Switch Transfer Panels TrnfrSic pump room 2A1 6.9 KV Switchgear breakers to trip Simple breaker action in the position located in the Turbine Building De-energize Turbine Building vicinity of the transfer panel (fuse switchgear room for breakers 2-30105, equipment to allow for Switchgear Transfer cks designed to be pulled without tools). Therefore, meets 30104, and 30103 including pulling fuse HSCP control Panels the dno R1 blocks the definition in RG 1.205 Revision 1 and FAQ 07-0030 2A2 4160V Switchgear Normal/Isolate Turbine Building switches located in the Turbine Building Transfer of Control Switchgear Transfer Transfer Switch switchgear room Panels 2B2 4160V Switchgear Normal/Isolate Turbine Building switches located in the Turbine Building Transfer of Control Switchgear Transfer Transfer Switch switchgear room Panels 2B1 6.9 KV Switchgear position breakers Simple breaker action in the to trip position located in the Turbine De-energize Turbine Building vicinity of the transfer panel (fuse Building switchgear room for breakers 2-equipment to allow for Switchgear Transfer cks designed to be pulled without tools). Therefore, meets 30203, 2-30204, and 2-30205 including HSCP control Panels the dno R1 pulling fuse blocks the definition in RG 1.205 Revision 1 and FAQ 07-0030 MCC 2AB Normal/Isolate switches located C

in the cable spreading room (control room Transfer of Control able Spread RoomTransfer Switch fire only)

Transfer Panels Simple breaker action in the MCC 2AB position breaker 2-42406 to off De-energize C

vici located in the cable spreading room equipment to allow for Table Spread Room nity of the transfer panel.

Transer Pnels Therefore, meets the definition in (control room fire only)

HSCP control RG 1.205 Revision 1 and FAQ 07-0030 2A3 480V Load Center Normal/Isolate C

Trasfe Swtc switches located in the cable spread room Transfer of Control able Spread RoomTransfer Switch (control room fires only)

Transfer Panels Simple breaker action in the 2A3 480V Load Center position breaker 2-De-energize Cable Spread Room vicinity of the transfer panel.

40305 to off located in the cable spread equipment to allow for Table Prels Therefore, meets the definition in Transfer Panels Teeoe et h

eiiini room (control room fires only)

HSCP control RG 1.205 Revision 1 and FAQ 07-0030 2B3 480V Load Center Normal/Isolate switches located in the cable spread room Transfer of Control Transfer Panels R (control room fires only)

Simple breaker action in the 2B3 480V Load Center position breaker 2-De-energize vicinity of the transfer panel.

40602 to off located in the cable spread equipment to allow for Cable Spread Room Therefore, meets the definition in Transfer Panels Teeoe et h eiiini room (control room fires only)

HSCP control RG 1.205 Revision 1 and FAQ 07-0030

Enclosure I to L-2014-056 Page 57 of 75 PCS Description from Transfer of Control Location (All or Power (energize/

locations are on Remarks LAR Attachment G de-energize)

Unit 2)

B PORV and RCS gas vent isolation B Penetration Room located in the 2B electrical penetration Transfer of Control Transfer Panels Transfer Switch room A PORV and RCS gas vent isolation A Penetration Room located in the 2A electrical penetration Transfer of Control Transfer Panels Transfer Switch room 2AB 4160 Switchgear Normal/Isolate A

switches located in the AB switchgear Transfer of Control aB Switchgear Room Transfer Panels TrnfrSic room Simple breaker action in the 2AB 4160 Switchgear located in the AB Energize equipment nity of the tr swichgar oomcloe beakrs -2002 o alowforHS A AB Switchgea r Room vicint oftetansfer panel.

switchgear room close breakers 2-20502 to allow for HSCP Transfer Panels Therefore, meets the definition in and 20503 control RG 1.205 Revision 1 and FAQ 07-0030 2AB 480V Load Center Normal/Isolate AB Switchgear Room switches located in the AB switchgear Transfer of Control Transfer Panels Transfer Switch room Simple breaker action in the 2AB 480V Load Center located in the AB De-energize vicinity of the transfer panel.

switchgear room position breakers 2-equipment to allow for AB Switchgear Room Therefore, meets the definition in 40702, 2-40706, and 2-40707 to open HSCP control RG 1.205 Revision 1 and FAQ 07-0030 2A diesel generator Normal/Isolate A EDG Transfer switches located in the 2A diesel generator Transfer of Control Panel (part of A EDG Transfer Switch room control panel)

PSL SSA RAI 02 LAR Attachment B, Table B-2 Section 3.2.1.2 "Fire Damage to Mechanical Components" identifies that mechanical components subjected to a fire are not considered credible. However, the guidance also states that instrument tubing with brazed, soldered joints, or other heat sensitive materials should not be included in this non-failure assumption. Describe how the failure of brazed or soldered joints in a fire was considered in the NFPA 805 nuclear safety capability analysis.

RESPONSE

The systems relied upon to satisfy the post-fire NFPA 805 nuclear safety performance criteria do not contain brazed or soldered joints. The integrity of the instrument air system may have some vulnerability due to the presence of heat sensitive components associated with the end devices, but this system is not credited in the nuclear safety capability assessment (NSCA). This system is assumed to fail in the NSCA resulting in all instrument air powered valves failing to the loss of air position unless a fire induced cable failure could a cause a valve to fail in an adverse position. The NSCA then assumes the valve will move to the adverse position even if that would require instrument air to be available. Therefore, the failure of brazed or soldered joints does not impact the NSCA.

Enclosure I to L-2014-056 Page 58 of 75 PSL SSA RAI 03 LAR Attachment B, Table B-2 Section 3.2.1.2 "Fire Damage to Mechanical Components" states:

"...the SSA includes required actions to manually operate valves that are in the affected fire area."

This appears to be in conflict with the statement made in LAR Section 4.2.1.1, Subheading "Comparison to NEI 00-01, Revision 2" that states "There are no recovery actions on valves in the same area as the fire."

Discuss which statement is correct. If the safe shutdown analysis (SSA) includes required actions to manually operate valves that are in the affected fire area, provide an analysis that demonstrates the operability of the valves given potential fire damage.

RESPONSE

The statement made in LAR Section 4.2.1.1, Subheading "Comparison to NEI 00-01, Revision 2" that states "There are no recovery actions on valves in the same area as the fire" is correct. This statement specifically addresses rising stem manual valves as discussed in NEI 00-01 Revision 2 (coefficients of friction, handwheel sizes, rim pulls, etc.). PSL-FPER-11-002, NFPA 805 Recovery Action Feasibility Evaluation, concluded that all actions relied upon for risk, or defense-in-depth, are feasible. Some of the action statements include entry into the affected area, but these are all contingency actions and none of these actions involved the use of rising stem valves.

The statement in the B-2 Table was made during its initial development before the evaluation of recovery actions and defense in depth had taken place. At that time actions were included that are now not required based on a fire risk evaluation that evaluated VFDRs that resulted from these actions as being acceptable for risk, defense in depth, and safety margin with the action not being performed. Additionally, the actions taking place in an affected fire area are contingent upon accessibility while the credited actions are those that do not require access to the affected area.

This portion of the B-2 Table should have been revisited for update. The LAR Table B-2 will be revised. A markup is attached to this RAI response.

Enclosure I to L-2014-056 Page 59 of 75 PSL SSA RAI 03 - Mark-up Attachment B NEI 04-02 Table B-2, Nuclear Safety Capability Assessment - Methodology Review 2.4.2.1 Nuclear Safety Capability System and Equipment Selection NEI 00-01 Ref.

3.2.1.2 [Fire Damage to Mechanical Components (not electrically supervised)I NEI 00-01 Guidance 3.2.1.2 Assume that exposure fire damage to manual valves and piping does not adversely impact their ability to perform their pressure boundary or safe shutdown function (heat sensitive piping materials, including tubing wilh brazed or soldered joints. are not included in this assumption). Fire damage should be evaluated with respect to the ablillyto manually open or close the valve should this be necessary as a part of the post-fIre safe shutdown scenario.

Applicability Applicable Alignment Statement Alionment Basis Reference ALIGNS WITH INTENT Aithough the methodology documentation does not consider mechanical 2998-B-048 Rev. 21 - Unit 2 Appendix R Safe Shutdown Analysis damage to components as a result of fire, the SSA includes required actions to Report manually operate valves that are in the affected fire area. Evaluations were 8770-B-048 Rev. 31 - Unit 1 Appendix R Safe Shutdown Analysis performed to determine the effect of maloperation of equipment, and time Report limits were established as the bounds for mitigation, including manual actions.

PSL-ENG-SEMS-98-035 Rev. 4 [2.6.7,2.6.2.4.2.3.1] - St. Lucie A manual action feasibility study was conducted to ensure required actions Unit 1 Appendix R Validation Effort Safe Shutdown Analysis could be performed within the established time constraints.

PSL-ENG-SEMS-98-067 Rev. 4 [2.6.7, 2.6.2.4. 2.3.11 - St. Lucie Unit 2 Appendix R Validation Effort Safe Shutdown Analysis Ruantitative evaluations on the effects on mechanical attributes involved in the PSL-FPER-11-002 Rev. 1 - St. Lucie Nuclear Power Plant Units 1 Replace existing mitigation strategy were nol performed, but qualitative evaluation (i.e.,

and 2 NFPA 805 Recovery Action Feasibility Evaluation alignment basis response time and available alternatives) are included in the SSA.

istatement with A A

Athough the methodology documentation does not consider mechanical damage to components as a result of fire, the SSA Includes required actions to manually operate valves that are in the affected fire area. These actions are contingent upon accessibility while the credited actions are those that do not require access to the affected area.

Evaluations were performed to determine the effect of maloperation of equipment, and lime limits were established as the bounds for mitigation, including manual actions. A manual action feasibility study was conducted to ensure required actions could be performed within the established time constraints.

Fire Safety Analysis Data Manager (4.129)

FPL - St. Lucie Run: OY18/2013 01:15 Page: 36 of 102

Enclosure I to L-2014-056 Page 60 of 75 PSL SSA RAI 05 LAR Attachment B, Section 3.3.1.6 "ESFAS Initiation" identifies the compliance strategy as "Aligns with Intent". Provide an explanation of the portion of the guidance provided in NEI 00 01, "Guidance for Post Fire Safe Shutdown Circuit Analysis," Rev. 2, Section 3.3.1.6 that is not met

RESPONSE

LAR Attachment B, Section 3.3.1.6 "ESFAS Initiation" is from NEI 00-01 Revision 1. In NEI 00-01 Revision 2 this section is 3.3.1.1.4.1. The wording between the two sections in each revision is the same. Evaluation of multiple instrument cable failures was included in the SSA as potential generators of spurious safety system initiation signals. The evaluations were done at the affected component level, not wholesale at the ESFAS logic level, which the NEI guidance implies should be done. Although specific signal-by-signal analyses are not performed in all cases, all automatic signals that could affect safe shutdown components are included in the SSA.

As stated in Step 3 of FAQ 07-0039 "Since NEI 00-01 is a guidance document, portions of its text could be interpreted as 'good practice' or intended as an example of efficient means of performing the analysis. In some instances the commentary presents analytical preferences which can be performed in a number of different ways without impacting the validity of the results.

These sections of NEI 00-01 can be dispositioned without further reference." In this case NEI 00-01 states "If not protected from the effects of fire, the fire-induced failure of automatic initiation logic circuits should be considered for their potential to adversely affect any post-fire safe shutdown system function." Because St. Lucie does not analyze automatic initiation circuits directly but analyzes the inputs to the logic and the outputs from the logic as impacting individual components it was felt during the development of the LAR that this does not meet the exact statement in the guidance. However, the results do analyze the impact of spuriously generated automatic initiation signals for the adverse effects on components required for safe shutdown. Therefore, as stated in FAQ 07-0039 the analysis was done in a different manner without impacting the validity of the results. That is why the statement of 'aligns with intent' was chosen.

Enclosure I to L-2014-056 Page 61 of 75 PSL SSA RAI 06 LAR Attachment B, Section 3.3.1.7 "Circuit Coordination" states "St. Lucie did not follow the methodology in NEI 00-01 but did end up with the same result." Provide more explanation with regard to the scope of circuit coordination used to satisfy NFPA 805. Include in that discussion the types of resolutions for lack of coordination and if any of these resolutions included variance from deterministic requirements (VFDRs), RAs, modifications, electrical raceway fire barrier systems (ERFBS), or "risk evaluation with no further action required."

RESPONSE

The statement above is based on the revision of the methodology documents in affect at the time of the LAR preparation. The coordination analysis was included in the methodology documents (PSL-ENG-SEMS-98-035 and PSL-ENG-SEMS-98-067) in Attachment 4, Circuit Breaker/Fuse Coordination Study. This study was performed with a focus on safe shutdown load components (the first type of association described in the guidance). The power sources for the essential equipment were reviewed to assure that the circuit protective devices are selectively coordinated, such that the load breakers in the circuit will open to clear a downstream fault prior to the main supply breaker to the power source opening. Any circuits for redundant train equipment which were noted to be routed through the same fire area during the review were evaluated to assure that one train of equipment needed for safe shutdown remains available for a fire in any plant fire area. This evaluation justified the loss of a power supply by looking at the available equipment and power supplies to ensure that at least one train of equipment needed to attain safe shutdown for a fire in any plant area was available. These justifications may have credited redundant equipment but may have also credited existing operator manual actions in the pre-transition safe shutdown analysis (SSA). However, the associated cables were never input to the SSA. This does not follow the guidance in NEI 00-01 because the information was not added to the SSA.

The results were valid because circuit coordination was considered and circuits that resulted in additional failures were considered. This documentation was difficult to retrieve and static. This weakness was recognized during the transition project.

The electrical coordination studies were updated in calculations PSL-1FSE-09-001 and PSL-2FSE-08-001. These calculations identified previously unknown weaknesses in the electrical coordination. The issue was captured in Action Request (AR) 574316. This AR generated multiple additional ARs as well as engineering changes to update the SSA with cables that would affect circuit coordination that were previously justified within the electrical calculations. The information generated by the ARs associated with this issue was incorporated into the Fire PRA (by adding cables as associated cables) as well as commitments to resolve issues that require modifications to resolve (LAR Attachment S Table S-1 Items EC 278417, EC 278455, EC 278504, 278508, and EC 278510). The SSA and supporting documentation have not yet been updated but this does not change the results of the analysis supporting the LAR because the cables remaining (after the modifications are complete) that may impact coordination have been included in the risk analysis as associated cables. These remaining coordination issues do not result in any additional VFDRs; the cables that impact the coordination have been incorporated into the fire risk evaluation process. The only remaining tasks are to update the SSA and supporting documents with the information that has already been generated.

Enclosure I to L-2014-056 Page 62 of 75 PSL SSA RAI 07 LAR Attachment B, Section 3.3.3.2 interlocked circuits whose spurious operation could affect shutdown identifies the compliance strategy as "Aligns with Intent". Provide an explanation of the portion of the NEI 00-01, Rev. 1, Section 3.3.3.2 that is not met.

RESPONSE

All portions of NEI 00-01 Section 3.3.3.2 are met; the alignment statement for Section 3.3.3.2 was conservatively assigned "with intent" based on the close association with Section 3.3.3.3 which recommends a relational database. After this review, it is determined that the assignment of "Aligns" is appropriate for Section 3.3.3.2 when taken as a stand-alone statement. Attachment B will be changed appropriately. See the attached markup of LAR Attachment B, Table B-2.

Section 3.3.3.2 refers to Figure 3-3, Safe Shutdown Equipment Selection. This figure refers to Figure 3-4 where Steps 4 through 7 identify interlocked circuits, power requirements and assignment of cables to equipment. NEI 00-01 Section 3.3.3.3 refers to a relational database including interlocks that are typically tabulated with cable associations and applied to affected components as dependencies (e.g., main generator lockout circuits are listed as components with associated cables, then these components are listed as dependencies in the analyses of affected components). Although the overall method recommended in NEI 00-01 Section 3.3.3.3 had not been implemented (no relational database including interlocks), the results are still valid because the safe shutdown analyses include all cables required for equipment function as well as cables that could prevent equipment function or cause equipment mal-operation.

Enclosure I to L-2014-056 Page 63 of 75 PSL SSA RAI 07 - Mark-up Attachment B NEI 04-02 Table B-2, Nuclear Safety Capability Assessment - Methodology Review 2.422 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.

NEI 00-01 Guidance 3.3.3.2 Identify Interlocked Circuits and Cables In revieAng each control circuit investigate Interlocks that may lead to additional circuit schemes, cables and equipment. Assign to the equipment Whose Spurious Operation or Mat-operation any cables for Interlocked circuits that can affect the equipment Could Affect Shutdovn While investigating the interlocked circuits, additional equipment or power sources may be Oscovered. Include these interlocked equipment or poaer Applicability t~igvment Statemnent\\

A&IGNiS WrITH 09ENTj sources in the safe shuldv*y.n equipment list (refer to Figure 3-3) if they can impact the operation of the equipment under consideration.

Alignment Basis Reference The cable selection included all cables required for equipment function as well PSL-.ENG-SEMS-98-035 Rev. 4 [2.3.2] - St. Lucle Unit 1 Appendix as cables that could prevent equipment function or cause equipment R Validation Effort Safe Shutdem.AnaJysis malopera3lon. The analysis does consider all potential Interlocks and cables PSL-ENG-SEMS-98-067 Rev, 4 [2.3.2] - St. Lucie Unit 2 Appendix that could affect equipment operaion in the analysis of the required cables for R Vatidnon Effort Safe Shutdown Analysis affected equipment.

Fire Safety Analys*s Data Manager (4.129)

FPL - St. Lucie Run: 0311 SnOl 3 01:15 Page: 63 of 102

Enclosure I to L-2014-056 Page 64 of 75 PSL SSA RAI 08 LAR Attachment B, Section 3.3.3.4, "Routing Cables," and 3.3.3.5, "Routing Raceway," state that "[... ] the raceway details would only provide an additional layer of information however, this additional level of detail does not affect the end results since the location cable information is ultimately determined[... ]. The end result was several reports from the cable routing system which listed cables by fire area as well as listing cables by fire zone routes. As a result, the raceway information is not important to the results and is not captured in the SSA."

For individual fire scenarios with zones of influence (ZOIs) smaller than a fire zone; explain how targets are identified and included in the analysis. Identify any assumption made for cables without routing information.

RESPONSE

The intent of NEI 00-01 Section 3.3.3.4 and 3.3.3.5 is to determine the fire area location for each cable. This supports a deterministic analysis by fire area which is the subject of NEI 00-01. St.

Lucie has a cable to fire zone (area) relationship but not a raceway to fire zone relationship in its database. This is the case as stated in FAQ 07-0039 that the methodology differs but the results remain valid. The assumed method in NEI 00-01 is that the primary fire zone (area) information is raceway to fire zone. However, the intended result is a list of cables in a fire zone (area). St.

Lucie has the intended result of a list of cables in a fire zone (area) but does not use the raceway to fire zone (area) relationship to develop that. For individual fire scenarios in the Fire PRA, walkdowns were performed wherein fixed ignition sources were confirmed, as necessary, and transient ignition sources were identified. Both vertical and horizontal Zones of Influence (ZOI) were calculated based on the properties of the ignition source(s) identified above. An additional walkdown was performed to identify all the equipment, conduit and trays within the zone of influence. These were identified as targets for the scenario. Cable/raceway data is then extracted from the PSL Cable and Raceway Schedule (CARS). Database queries are used to identify all cables routed through the identified raceway, panel or component targets from the CARS database. Any cable without cable routing was walked down and the routing information entered into the CARS database. The component/cable data from the safe shutdown analysis is compared, via database queries, to the cables impacted by the scenarios to generate a list of components impacted by each scenario. It should be noted that the PSL CARS has been transferred to an Edison database. However, the methodology is the same in that raceways (cable tray or conduit) are identified via walkdown and the cables contained within those raceways is obtained from queries in the Edison database.

The scenarios, including the ignition source and the targets are summarized in Attachment A of Report 0493060006.104. The "A" scenarios with zone numbers representing actual plant zones represent the equivalent of an Appendix R exposure fire where everything in the compartment is assumed to fail at the compartment frequency. Most other scenarios are those with a ZOI smaller than a fire zone.

PRA scenario development methodology and the use of raceway information were not utilized to answer/address the concerns of the deterministic approach used to generate Attachment B of the LAR, which is the deterministic approach, or NSCA.

Enclosure I to L-2014-056 Page 65 of 75 PSL SSA RAI 09 LAR Attachment B, Section 3.5.2.1, "Circuit Failures Due to an Open Circuit," with respect to current transformers (CTs) states that "the potential for secondary fires (especially for CT circuits on ammeters not associated with safe shutdown equipment) requires additional analysis.

The licensee commits to ensuring that CT circuits with the potential for secondary fire will be protected from fire-induced open circuits as part of its implementation for NFPA 805." This appears as though future analysis or work will be required.

LAR Attachment S, Table S-2, "Implementation Items," number 10 states, "Review NUREG/CR-7150 Vol. 2 when published to determine if conclusion that an open secondary CT is not a concern [... ]." Provide an explanation of what work is still outstanding, and estimate the potential impact to the fire protection program (FPP) as presented in the LAR. Justify why this additional work should not be added as an Implementation Item.

RESPONSE

LAR Section 4.2.1.1 identifies the process used by PSL to document compliance with NFPA 805 Section 2.4.2. The initial methodology review, documented in the LAR Attachment B, evaluated the existing post-fire safe shutdown analysis (SSA) methodology against the guidance provided in NEI 00-01, Revision 1, Chapter 3, "Deterministic Methodology," as discussed in Appendix B-2 of NEI 04-02. At the time of this initial review, there was no definitive industry or regulatory guidance related to the topic of fire induced open circuits on CT secondary windings (NEI 00-01 Section 3.5.2.1). Subsequent to this initial review, but before the PSL LAR was submitted in March 2013, NEI 00-01 Revision 2 and NUREG/CR-7150 Volume I were issued. LAR Section 4.2.1.1 documents the review of NEI 00-01 Revision I to Revision 2 (i.e., "gap analysis"). This review included the regulatory guidance contained in the NUREG and concluded that open circuit on the secondary CT windings would not impact the NSCA and would not be considered.

PSL LAR Attachment S, Table S-2 Implementation Item 10 is to review NUREG/CR-71 50 Volume 2 (when issued) to determine if the information published in Volume 2 impacts the conclusion that an open circuit on the secondary circuit of a CT does not impact the NSCA and does not need to be considered in the Fire PRA. FPL considers that the information that will be contained in NUREG/CR-7150 Volume 2 will not change the conclusion in this section of the LAR and no future work beyond this review and an update to the supporting documents (PSL-ENG-SEMS-98-035, PSL-ENG-SEMS-98-067) will be required. LAR Attachment B Section 3.5.2.1 was not updated with the results of the NEI 00-01 revision I to Revision 2 gap analysis.

A markup of the LAR Attachment B, Section 3.5.2.1 is attached to the response to this RAI.

PSL SSA RAI 09 - Mark-up Attachment B NEI 04-02 Table B-2, Nuclear Safety Capability Assessment - Methodology Review Enclosure I to L-2014-056 Page 66 of 75 2.4.2.2 Nuclear Safety Capability Circuit Analysis 2.4.2.2 Nuclear Safety Capability Circuit Analysis NEI 00-01 Ref.

3.5.2.1 Circuit Failures Due to an Open Circuit NEI 00-01 Guidance This section provides guidance for addressing the effects of an open circuit for safe shutdown equipment. An open circuit is a fire-induced break in a conductor resulting in the loss of circuit continuity. An open circuit will typically prevent the ability to control or power the affected equipment. An open circuit can also result in a change of state for normally energized equipment. For example, a loss of power to the main steam isolation valve (MSIV) solenoid valves [for BWRs] due to an open circuit will result in the closure of the MSIV.

NOTE: The EPRI circuit failure testing indicated that open circuits are not likely to be the initial fire-induced circuit failure mode. Consideration of this may be helpful within the safe shutdown analysis. Consider the following consequences in the safe shutdown circuit analysis when determining the effects of open circuits:

Loss of electrical continuity may occur within a conductor resulting in de-energizing the circuit and causing a loss of power to, or control of, the required safe shutdown equipment.

In selected cases, a loss of electrical continuity may result in loss of power to an interlocked relay or other device. This loss of power may change the state of the equipment. Evaluate this to determine if equipment fails safe.

GO A_."

GaG,-61 OR" a_ high,-1tagG 4egA.

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[Refer to hard copy of NEI 00-01 for Figure 3.5.2-1]

NEI 00-01 Rev 2: Open circuit on a high voltage (e.g., 4.16 kV) ammeter Open circuit No. 1:

current transformer (CT) circuit may result in secondary damage, possibly An open circuit at location No. 1 will prevent operation of the subjereutninheocrneofaadtoalfeinheoainofheC Open circuit No. 2:

An open circuit at location No. 2 will prevent opening/starting of the subject equipment, but will not impact the ability to close/stop the equipment.

Applicability Applicable Alignment Statement ALIGNS WITH INTENT Alignment Basis PSL safe shutdown analysis generally considers open circuits. Open circuits were considered on CT circuits on ammeters associated with safe shutdown equipment thi pnfitnfi2l fnr seonnyi, /furec (hsp-c!-i--,' on GT rir:'itc MR ammoc ntO A

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[related to the loss of indication. Based on the guidance provided in NUREG/CR-7 150 Vol. I regarding the potential for a secondary fire due to fire damage to CT circuits, PSL has concluded Ithat open circuits on the secondary of CTs would not impact the NSCA and would not be considered.

Refer to LAR Section 4.2.2. 1, "Comparison to NEI 00-0 1 Revision 2".

Reference PSL-ENG-SEEJ-11-001 Rev. 0 - NFPA 805 Review of St. Lucie Units 1 & 2 4160V Switchgear Breakers (EC 273329)

PSL-ENG-SEMS-98-035 Rev. 4 [2.5.1.3.b] - St. Lucie Unit 1 Appendix R Validation Effort Safe Shutdown Analysis PSL-ENG-SEMS-98-067 Rev. 4 [2.5.1.3.b] - St. Lucie Unit 2 Appendix R Validation Effort Safe Shutdown Analysis

',NUREG/CR-7150, Volume 1, 10/2012 [6.2]-Joint

'Assessment of Cable Damage and Quantification of Effects from Fire (JAQUE-FIRE)I Fire Safety Analysis Data Manager (4.129)

FPL - St. Lucie Run: 03/1812013 01:15 Page: 74 of 102

Enclosure I to L-2014-056 Page 67 of 75 PSL SSA RAI 10 LAR Section 4.2.1.1, "Compliance with NFPA 805 Section 2.4.2" states that "the only current transformers with larger turns ratio are located in the turbine building for the power feeds from the startup and auxiliary transformers and the generator output and current transformers located in the switchyard. These do not impact the NSCA (no credit for offsite power)." Clarify what is meant by "no credit for offsite power." Provide more explanation why potential secondary fires do not affect the NSCA. Describe how secondary fires due to CTs were addressed in the Fire Probabilistic Risk Assessment (FPRA).

RESPONSE

The NSCA for both units do not credit offsite power to achieve and maintain post fire safe and stable conditions. The analysis assumes a loss of offsite power unless that assumption would prevent a mal-operation that could impact safe and stable conditions. The Offsite Power feed to the 4.16 kV Essential/Vital Switchgear is from the 4.16 kV Non-Vital Switchgear located in the Turbine Building Switchgear area. Therefore, any failures in the Turbine Building Switchgear or the feeds to the switchgear from the transformers (auxiliary or startup) do not impact the NSCA.

A secondary fire in the Turbine Building Switchgear will not impact the NSCA.

LAR Section 4.2.1.1 documents the review of the guidance provided in NUREG/CR-7150 Volume 1. The regulatory guidance contained in the NUREG concludes that the ignition of a secondary fire from an open circuited CT secondary circuit with a turns-ratio of 1200:5 or less is incredible. As documented in LAR Section 4.2.1.1, all the CTs in safety related areas are 1200:5 or less. For CT ratios larger than 1200:5, NUREG/CR-7150 Volume 1 states that the likelihood of secondary fires was judged to be very low. Evidence could not be found that the open circuit failure would cause a secondary fire, but this could not be proven without additional testing. The open circuit failure would drive the voltage in the CT to higher than the insulation rating and would cause local arcing within the windings of the CT. The arcing drops the voltage as soon as any current flows. This cycle may repeat causing local overheating of the CT but would not damage anything beyond the switchgear compartment where the CT is located. The damage is essentially self-limiting.

With respect to the Fire PRA, a secondary fire due to an open circuit CT need not be considered because NUREG/CR-6850 states that an open circuit is not the primary failure mode. In addition, even given the open circuit, the probability of causing damage beyond the affected CT is also very low. Based on this a fire induced open circuit CT causing a secondary fire that induces additional fire induced failures is considered such a low probability that it would not impact the final risk results. In addition to the judgment that the failure mode is low likelihood and low risk, there is no guidance or accepted approaches for modeling the phenomenon in the Fire PRA. Modeling this type of scenario in the Fire PRA, without an established approach, is considered to be inappropriate without additional testing or guidance. Therefore, secondary fires due to open circuit CTs are not included in the Fire PRA.

Enclosure I to L-2014-056 Page 68 of 75 PSL SSA RAI 11 LAR Attachment B, Section 3.4.1.5, "Repairs," states that "St. Lucie Unit I does not credit any repairs for cold shutdown. St. Lucie Unit 2 takes credit for limited repairs during cold shutdown (i.e., inserting fuses that are part of the normal shutdown procedures)." Provide a more detailed explanation of those repairs. For those fire areas where repairs are necessary to achieve safe and stable, describe the location of these actions. Describe how these are procedurally controlled with parts/fuses/components pre-staged, and how is training conducted to perform this work.

Describe whether the cold shutdown repairs are included in the FPRA

RESPONSE

The referenced section of the LAR pertains to the installation of fuses for the components identified in the table below. The fuse installation occurs at the identified RTGBs in the Control Room. The PIs are the Pressurizer Pressure Low Range Indicators and the valves are the Safety Injection Tank Vents.

Component to be Action*

Energized PI-1103, PI-1104 Insert RTGB -203 120V AC "SA" fuse F21 PI-1105, PI-1106 Insert RTGB -203 120V AC "SB" fuse F21 V3733 Insert RTGB -206 125V DC "SA" F53, F54 V3734 Insert RTGB -206 125V DC "SB" F117, F118 V3735 Insert RTGB -206 125V DC "SA" F51, F52 V3736 Insert RTGB -206 125V DC "SB" Fl 15, F116 V3737 Insert RTGB -206 125V DC "SA" F57, F58 V3738 Insert RTGB -206 125V DC "SB" F47, F48 V3739 Insert RTGB -206 125V DC "SA" F59, F60 V3740 Insert RTGB -206 125V DC "SB" F49, F50 During normal operation these fuses are not installed in the circuit until directed by 2-GOP-305, REACTOR PLANT COOLDOWN-HOTSTANDBY TO COLD SHUTDOWN, to enable operation of the respective component while proceeding to cold shutdown conditions. Therefore, this activity is required for all fire areas to energize the protected train equipment because the fuses are normally removed (power removed during power operation). The components identified in the table above are not subject to fire damage in any Fire Area where they are part of the safe shutdown protected train. The activity to install the fuse is part of the normal plant shutdown procedure as directed by 2-GOP-305, the capability is already maintained by the station, and the activity is not the result of fire damage, therefore, such a condition does not constitute a Cold Shutdown (CSD) repair and is not identified as a VFDR because the action is in the control room and part of the normal control room duties during plant cooldown. LAR Attachment B, Section 3.4.1.5 was overly conservative in identifying this activity as a cold shutdown repair. LAR Attachment B Section 3.4.1.5 has been revised and is attached to this RAI response. The insertion of these fuses is part of the normal plant procedures and is included in the internal events PRA as well as the Fire PRA for sequences that require cooldown to cold shutdown conditions.

PSL SSA RAI 11 - Mark-up Enclosure I to L-2014-056 Page 69 of 75 Attachment B NEI 04-02 Table B-2, Nuclear Safety Capability Assessment - Methodology Review 2.4.2.4 Fire Area Assessment.

NEI 00-01 Ref.

3.4.1.5 [Repairs]

Applicability Applicable Alignment Statement ALIGNS NEI 00-01 Guidance Where appropriate to achieve and maintain cold shutdown within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, use repairs to equipment required in support of post fire shutdown.

Alignment Basis St. Lu-cei Un_.it 1 dogs not credit any repai*rs for cold shu-tdown. St. Lucia U.nt 2 tkaks re.dit for limited repairs during cold shutdown (i.e. inserting fuses that are part..f the normal shutdown pr.ocedues). NFPA 805 only requires that the plntbepaced in a sfaf an~d stable condfitin post fire. Tho report PS;L=F=PER-Reference PSL-ENG-SEMS-98-035 Rev. 4 [2.1] - St. Lucie Unit 1 Appendix R Validation Effort Safe Shutdown Analysis PSL-ENG-SEMS-98-067 Rev. 4 [2.1 - St. Lucie Unit 2 Appendix R Validation Effort Safe Shutdown A ysis PSL-FPER-11-012 Rev. 0[1.0, 2.0] NFPA 805 Fire Protection Evaluation to Define Safe and Stabh for Use at St. Lucie Units 1 & 2 1 1-01:2

."NiFP A!

2805 Fire Prot Use at St. L.uck Units 1 & 2" d-atQ4-on~u itA~mised-2 tfion to Define Safe and St2ble for nd identified the St. Lucie specific z

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and stable condition.

,St. Lucie Units I and 2 do not credit any repairs for cold shutdown.

' NFPA 805 only requires that the plant be placed in a safe and stable

'condition post fire. The report PSL-FPER-11-012, "NFPA 805 Fire Protection Evaluation to Define Safe and Stable for Use at St. Lucie Units 1 & 2" determined and identified the St. Lucie specific plant operational mode at which condition the fuel can be maintained in a safe and stable condition.

Fire Safety Analysis Data Manager (4.129)

FPL - St. Lucie Run: 03/18/2013 01:15 Page: 92 of 102

Enclosure I to L-2014-056 Page 70 of 75 PSL SSA RAI 12 LAR Attachment B, Table B-2, Section 3.1, "C. Spurious Operations," indicates that spurious signals, impact on high-low pressure interfaces, multiple spurious operation (MSO), and common enclosure/power supplies were analyzed for their impact on NSCA. However the staff noted the referenced analysis does not address all of these elements. The referenced analysis does not appear to reflect additional components and power supplies that were added for NFPA 805 considerations of MSO, common enclosure/power supplies, or equipment added for the NSCA. Describe how the additional equipment of the FPRA were added to the nuclear safety equipment list.

RESPONSE

The subject RAI requests how additional equipment of the FPRA were added to the nuclear safety equipment list. For St. Lucie the equipment required for safe shutdown under Appendix R is the essential equipment list on each unit. This contains the equipment required to mitigate the consequences of a fire as well as the required power supplies for the equipment to function.

These essential equipment lists have not yet been updated with equipment/power supplies, etc.

required by the FPRA. During the NFPA 805 project it was recognized that the St. Lucie documentation had some weaknesses in that there was no cable-to-component relationships. The safe shutdown analysis was a cable by cable analysis where the cable description was used to describe the cable to component relationship. However, this makes the safe shutdown analysis labor intensive and does not lend itself to any automated processes. As a result, the project developed a separate cable-to-component evaluation that included additions made for FPRA, MSO considerations, common power supplies, etc., to support the identification of VFDRs and performance of fire risk evaluations. This report was updated to capture the latest available information required to the FPRA. This is currently maintained as a separate report (FPLSL 120-PR-001) and has not yet been incorporated into the essential equipment list or the safe shutdown analysis.

As part of the transition process after the Safety Evaluation is received, all of this information will be captured in a revised essential equipment list as well as the nuclear safety capability assessment (NSCA) which will be developed from the existing SSA and the above referenced report. This activity is addressed by LAR Attachment S, Table S-2 Implementation Item 16. The above referenced report captured information that currently exists in the corrective action program as well as outstanding engineering changes that have not yet been incorporated into the bases analysis. The response to RAI SSA 13 contains additional information.

Enclosure I to L-2014-056 Page 71 of 75 PSL SSA RAI 13 LAR Attachment B, Table B-2, Section 3.2.1.6, "Spurious Components," indicates that "the guidance related to multiple spurious operations from NEI 00-01 Revision 2 has been addressed and included in the SSA or entered into the plant corrective action program." A similar statement was made for Section 3.5.1.5, "Circuit Failure Risk Assessment." Provide more detail regarding the resolution of the identified corrective actions. Describe what FPRA equipment/power supplies are still awaiting disposition in the corrective action program and how this will affect the results of the NSCA currently presented in the LAR.

RESPONSE

The issues related to MSOs identified from the process outlined in Chapter 4 of NEI 00-01 Revision 2, which were not already included in the SSA, were entered into the corrective action program to evaluate appropriate compensatory measures and obtain resolution. In several cases the actual SSA of record had not been completely updated. However, if any corrective action indicated that a VFDR is required, a VFDR was generated and properly analyzed in the fire PRA and documented in Report FPLSL 120-PR-001, "Update Review for St. Lucie Units 1 and 2, Impact of ARs and ECs on FREs and Cable to Component Relationships". The appropriate cable to component and cable routing data was developed for these corrective actions and supplied to the Fire PRA. However, some of this data has not been formally incorporated into permanent plant documentation. The open corrective actions are tracking the update of the permanent plant documentation but will not impact the fire PRA or associated NSCA. There are no FPRA equipment or power supplies awaiting disposition in the corrective action program for inclusion in the LAR or the NSCA.

For example, the following Condition Reports/Action Requests had not been fully incorporated into the SSA at the time of the LAR submittal, but the identified VFDRs were generated from the cable to component and cable routing data developed for the issue (Reference 1) for evaluation in the FPRA and inclusion in the LAR.

AR 01680630- VFDR-1C-121, VFDR-IA-78 AR 00459049-VFDR-2B-592, VFDR-2F-261, VFDR-2J-10, VFDR-20-1 10

Enclosure I to L-2014-056 Page 72 of 75 PSL SSA RAI 14 The Fire Risk Evaluations (FREs) described in LAR Attachment C, NEI 04-02, Table B-3, "Fire Area Transition," all refer to evaluating VFDRs generally as "this condition was evaluated for compliance using the performance-based approach of NFPA 805, Section 4.2.4. An FRE determined that applicable risk, defense-in-depth (DID), and safety margin criteria were satisfied without further action." However, there is no specific description of the associated criteria or any other details identified in the LAR. LAR Section 4.5.2.2 generally defines DID and safety margin as listed in NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," but does not describe the methodology, controls, and systems for providing DID and safety margins. Provide additional detail, specific to St. Lucie, of the methods and criteria for evaluating DID and safety margins.

RESPONSE

The following methods and criteria for evaluating DID and safety margins were extracted from PSL-FPER-1 1-001, St. Lucie Nuclear Power Plant NFPA 805 Fire Risk Evaluations, Revision 1:

Defense-in-Depth (DID)

Guidance A review of the impact of the change on DID was performed, using the guidance below from NEI 04-02. NFPA 805 defines DID as:

" Preventing fires from starting

" Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage

" Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

In general, the DID requirement was considered satisfied if the proposed change does not result in a substantial imbalance amnong these elements (or echelons). The review of DID was qualitative and addressed each of the elements with respect to the proposed change. Fire protection features and systems relied upon to ensure DID were identified in the assessment (e.g., detection, suppression system).

Consistency with the DID philosophy is maintained if the following acceptance guidelines, or their equivalent, are met:

" A reasonable balance is preserved among 10 CFR 5 0.48(c) DID elements.

" Over-reliance and increased length of time or risk on performing programmatic activities to compensate for weaknesses in plant design is avoided.

" Pre-fire nuclear safety system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges to the system and uncertainties (e.g., no risk outliers). (This should not be construed to mean that more than one NSCA train must be maintained free of fire damage.)

" Independence of DID elements is not degraded.

" Defenses against human errors are preserved.

Enclosure I to L-2014-056 Page 73 of 75 The intent of the General Design Criteria in Appendix A to 10 CFR Part 50 is maintained.

DID Process Each Fire Area was evaluated for the adequacy of DID. In accordance with NFPA 805, Section 2.4.4, Plant Change Evaluation, "the evaluation process shall consist of an integrated assessment of the acceptability of risk, DID, and safety margins." NFPA 805, Section 4.2.4.2 refers to the acceptance criteria in this section. Therefore fire protection systems and features required to demonstrate an adequate balance of DID are required by NFPA 805 Chapter 4.

The VFDRs and the associated Fire Area risk (CDF) and scenario consequences (CCDP values) were evaluated to identify general DID echelon imbalances. Potential methods to balance the DID features were identified ensuring an adequate balance of DID features is maintained for the Fire Area. To aid in the consistency of the review of DID, the following guidance is provided in Table 5-2:

Table 5 Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 1: Prevent fires from starting Combustible Control Combustible and hot work controls are fundamental elements of DID Hot Work Control and as such are always in place. The issue to be considered during the FREs is whether this element needs to be strengthened to offset a weakness in another echelon thereby providing a reasonable balance. Considerations include:

Creating a new Transient Free Areas Modifying an existing Transient Free Area The fire scenarios involved in the FRE quantitative calculation should be reviewed to determine if additional controls should be added.

Review the remaining elements of DID to ensure an over-reliance is not placed on programmatic activities to compensate for weaknesses on plant design.

Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage Detection system Automatic fire suppression Portable fire extinguishers provided for the area Hose stations and hydrants provided for the area Fire Pre-Fire Plan Automatic suppression and detection may or may not exist in the Fire Area of concern. The issue to be considered during the FRE is whether installed suppression and or detection is required for DID or whether suppression/detection needs to be strengthened to offset a weakness in another echelon thereby providing a reasonable balance. Considerations include:

If a Fire Area contains both suppression and detection and fire fighting activities would be challenging, both detection and suppression may be required If a Fire Area contains both suppression and detection and fire fighting activities would not be challenging, require detection and manual fire fighting (consider enhancing the pre-plans)

" If a Fire Area contains detection and a recovery action is required, the detection system may be required.

Enclosure I to L-2014-056 Page 74 of 75 Table 5 Considerations for Defense-in-Depth Determination Method of Providing DID Considerations If a Fire Area contains neither suppression nor detection and a recovery action is required, consider adding detection or suppression.

The fire scenarios involved in the FRE quantitative calculation should be reviewed to determine the types of fires and reliance on suppression should be evaluated in the area to best determine options for this element of DID.

Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed Walls, floors ceilings and If fires occur and they are not rapidly detected and promptly structural elements are rated or extinguished, the third echelon of DID would be relied upon. The have been evaluated as issue to be considered during the FRE is whether existing separation adequate for the hazard.

is adequate or whether additional measures (e.g., supplemental Penetrations in the Fire Area barriers, fire rated cable, or recovery actions) are required offset a barrier are rated or have been weakness in another echelon thereby providing a reasonable evaluated as adequate for the balance. Considerations include:

hazard.

0 If the VFDR is never affected in the same fire scenario, Supplemental barriers (e.g.,

internal Fire Area separation may be adequate and no ERFBS, cable tray covers, additional reliance on recovery actions is necessary.

combustible liquid dikes/drains, 0

If the VFDR is affected in the same fire scenario, internal etc.)

Fire Area separation may not be adequate and reliance on a Fire rated cable recovery action may be necessary.

Reactor coolant pump oil E

If the consequence associated with the VFDRs is high collection system (as applicable) regardless of whether it is in the same scenario, a recovery action and / or reliance on supplemental barriers should be Guidance provided to operations considered.

personnel detailing the required success path(s) including 0

There are known modeling differences between a Fire PRA recovery actions to achieve and nuclear safety capability assessment due to different nuclear safety performance success criteria, end states, etc. Although a VFDR may be

criteria, associated with a function that is not considered a significant contribution to CDF, the VFDR may be considered important enough to the NSCA to retain as a recovery action.

The fire scenarios involved in the FRE quantitative calculation should be reviewed to determine the fires evaluated and the consequence in the area to best determine options for this element of DID.

Defense-in-Depth - Recovery Action Considerations Reliance on Recovery Actions in lieu of protection is considered part of the third echelon of DID. Per NFPA 805, recovery actions are defined as: "Activities to achieve the nuclear safety performance criteria that take place outside of the main control room or outside of the primary control(s) station for the equipment being operated, including the replacement or modification of components."

If the VFDR is characterized as a 'Separation Issue', and the change in risk (ACDF and ALERF) is acceptable, a recovery action can be considered as a means to provide an adequate level of DID. Guidance on the need to establish/rely upon a recovery action is provided in Table 5-2. The

'additional risk presented by the use of the recovery action', if relied upon for DID, would be characterized as the calculated change in risk of the 'Separation Issue'.

Enclosure I to L-2014-056 Page 75 of 75 Safety Margin Assessment A review of the impact of the change on safety margin was performed. An acceptable set of guidelines for making that assessment are summarized below. Other equivalent acceptance guidelines may also be used.

" Codes and standards or their alternatives accepted for use by the NRC are met, and

• Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.

The requirements related to safety margins for the change analysis is described for each of the specific analysis types used in support of the FRE. These analyses can be grouped into two categories. These categories are:

" Fire Modeling

" Plant System Performance The following guidance on these topics is provided. Additional information is contained in NEI 04-02 Section 5.3.5.3.

Fire Modeling For fire modeling used in support of the FRE (i.e., as part of the Fire PRA), the results were documented as part of the qualitative safety margin review.

Plant System Performance This review documented that the Safety Margin inherent in the analyses for the plant design basis events was preserved in the analysis for the fire event and satisfied the requirements of this section.