NL-14-0446, 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

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

ML14114A550
Person / Time
Site:	Farley
Issue date:	04/23/2014
From:	Pierce C Southern Co, Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-14-0446
Download: ML14114A550 (63)

Text

Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7872 ENCLOSURE CONTAINS INFORMATION NOT FOR PUBLIC DISCLOSURE Fax 205.992. 7601S SOUTHERN N APR 2 3 2014 COMPANY Docket Nos.: 50-348 NL-14-0446 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant 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 Ladies and Gentlemen:

By letter dated September 25, 2012, the Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) for Joseph M. Farley Units 1 and 2 (Ref. TAC NOS. ME9741 and ME9742). The proposed amendment requests the review and approval for adoption of a new fire protection licensing basis which complies with the requirements in Sections 50.48(a) and 50.48(c) to Title 10 to the Code of Federal Regulations (10 CFR), and the guidance in Regulatory Guide (RG)1.205, Revision 1, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants.By letter dated December 12, 2012, the Nuclear Regulatory Commission (NRC)Staff requested supplemental information regarding the acceptance of the license amendment (Adams Accession No. ML12345A398).

SNC provided the requested information by letter dated December 20, 2012. The NRC staff subsequently completed the acceptance review by letter dated January 24, 2013, (Adams Accession No. ML13022A158).

By letter dated July 8, 2013, the NRC Staff formally transmitted a request for additional information (RAI) related to the referenced license amendment.

SNC's responses to these RAIs are being provided by three submittals.

By letter dated September 16, 2013, SNC provided the first set of responses.

By letter dated October 30, 2013, SNC provided the second set of responses and by letter dated November 12, 2013, SNC provided the remaining set of responses.

SNC provided that supplemental responses would be provided for nine of the RAIs.

U.S. Nuclear Regulatory Commission NL-14-0446 Page 2 By letter dated March 28, 2014, the NRC Staff formally transmitted the second round of requests for additional information related to the referenced license amendment.

The enclosures to this letter provide responses to the RAIs in accordance with the agreed upon completion dates. The responses to PRA RAI 01.01, PRA RAI 06.a.01, and PRA RAI 35 will be provided by May 23, 2014. As discussed, the responses to PRA RAI 01.01, 16.a.01, 21 .a.01, 33.a.01, and 33.c.01 have been moved to a May 23, 2014 due date. Attachment G, Recovery Actions Transition, provides a replacement Table G-1. Attachment G contains sensitive information and should be withheld from public disclosure under 10 CFR 2.390. A revision to Attachment M, License Condition Changes, is also provided to docket the current proposed version of the license condition.

The No Significant Hazards Consideration determination provided in the original submittal is not altered by the RAI responses provided herein.This letter contains no new NRC commitments.

If you have any questions, please contact Ken McElroy at (205) 992-7369.Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true and correct.Respectfully submitted, C. R. Pierce Regulatory Affairs Director CRP/jkb/lac Sworn o and subscribegi before me this A-3 day of 2014.Notary Public My commission expires: _______/ __

Enclosures:

1. Supplemental Responses to Previous RIA Responses 2. Response to Safe Shutdown Analysis RAI 3. Response to Fire Modeling RIAs 4. Response to Probabilistic Risk Assessment RAIs U.S. Nuclear Regulatory Commission NL-14-0446 Page 2 Attachments:

1. Revision to Recovery Actions Transition-Attachment G 2. Revision to License Conditions

-Attachment M 3. Revision to Fire PRA Quality- Attachment V cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President

& CEO Mr. D. G. Bost, Executive Vice President

& Chief Nuclear Officer Ms. C. A. Gayheart, Vice President

-Farley Mr. B. L. Ivey, Vice President

-Regulatory Affairs Mr. D. R. Madison, Vice President

-Fleet Operations Mr. B. J. Adams, Vice President

-Engineering RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. S. A. Williams, NRR Project Manager -Farley Mr. P. K. Niebaum, Senior Resident Inspector

-Farley Mr. J. R. Sowa, Resident Inspector

-Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer Joseph M. Farley Nuclear Plant 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 Enclosure 1 Supplemental Responses to Previous RAI Responses Provided in the SNC Letter Dated November 12, 2013 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley RAI SSA 14 Item 6 of Attachment S to the submittal, states that for Fire Area 2-021 an interposing relay and fuse will be installed to protect cable 2VYDG15J from fire induced failure and to prevent the breaker from tripping.

It also states that for Fire Area 1-021 a fuse will be installed for cable 1VBJ5012F to prevent fire damage and that for Fire Area 2-041 a fuse will be installed for cable 2VAJ5007L to prevent fire damage. Provide clarification that the installation of the relays and fuses is for mitigating the secondary effects of cable damage and not to protect the cables from fire damage.Response provided by SNC letter NL-13-2269 dated November 12, 2013 The modifications identified in this RAI are intended to address common power supply associated circuits for the Fire PRA, as defined in Section 3 of NUREG/CR-6850.

The intent of the modifications is to prevent fire damage on the identified cables from causing a loss of power to circuitry associated with automatic operation of auxiliary feedwater (AFW). Each specific case is discussed below.Additional refinements to the Fire PRA model indicate that it will likely not be necessary to protect Cable 2VYDG15J with an interposing relay and fuse as originally described in Attachment S. At present, satisfactory CDF and LERF values are achieved without credit for these modifications.

Accordingly, the modifications associated with Cable 2VYDG15J will most likely be removed from Attachment S of the LAR. Final confirmation of this change will occur when the comprehensive impact of all model changes is established.

Cable 1VBJ5012F is associated with automatic operation of Unit 1 auxiliary feedwater (AFW).Depending on the specific functionality credited for affected AFW components, fire-induced failure of this cable can result in a loss of automatic AFW control. In most but not all scenarios the failure does not impact manual control from the Main Control Room. The intent of fusing this cable is to isolate the cable if it suffers a fire-induced short circuit, thereby preventing a loss of power to the automatic AFW control circuitry.

The fuse will not prevent hot-short induced spurious actuation of the automatic AFW control circuit. For these cases, the Fire PRA will reflect the failures, with no credit taken for the modification.

Cable 2VAJ5007L is associated with automatic operation of Unit 2 auxiliary feedwater (AFW). Depending on the specific functionality credited for affected AFW components, fire-induced failure of this cable can result in a loss of automatic AFW control. In most but not all scenarios the failure does not impact manual control from the Main Control Room. The intent of fusing this cable is to isolate the cable if it suffers a fire-induced short circuit, thereby preventing a loss of power to the automatic AFW control circuitry.

The fuse will not prevent hot-short induced spurious actuation of the automatic AFW control circuit. For these cases, the Fire PRA will reflect the failures, with no credit taken for the modification.

El -1 Enclosure 1 Supplemental Responses to Previous RAI Responses Supplemental Response: After further review of the proposed modification to cable 2VYDG15J and based on further model refinements completed in response to other RAIs, this plant modification is no longer required.

The modification to protect cable 2VYDG15J is no longer credited in the analysis and will be removed from Attachment S.The modifications for cables 1VBJ5012F and 2VAJ5007L are still required to address common power supply associated circuits for the Fire PRA. The plant modification for cables 1VBJ5012F and 2VAJ5007L will be updated in Attachment S based on the findings presented in the original response to RAI SSA 14. As identified in the original response to RAI SSA 14, there are only specific function states that the modification will be credited in. An updated Attachment S will be provided to reflect these changes for the three cables discussed in this RAI response with the response to RAI PRA 35.El -2 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 01(b) through 01(g)In Enclosure 6 to the supplement dated December 20, 2012 (ADAMS Accession No. ML12359A051), the results are presented for both the total and delta core damage frequency (CDF) that are actually lower than previously reported in Attachment W of the LAR. The submittal, although only the credit for the electrical cabinet factor was removed. With the additional removal of credit for the main control room (MCR) very early warning fire detection system (VEWFDS), it is expected that these CDF results would increase, consistent with the increases in the large early release frequency (LERF) values. Explain why, including any key modeling assumptions that may be relevant, the increases in total and delta-CDF are now lower than before especially in light of the higher increases in total and delta-LERF.

From the submittal, address the following:

b. In Section V.2, Sensitivity of Fire PRA Methods, specifically in Tables V.2-2 through V.2-4, discuss whether there are any delta-risk values sensitive to crediting risk reduction from installing VEWFDS in the MCR. If so, discuss how they would change if this credit were removed.c. In Table V-i, Fire PRA Peer Review -Facts and Observations, with respect to Supporting Requirement (SR) FQ-A3, discuss whether the installation of VEWFDS was credited after updating the analysis for the MCR using Appendix L of NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," including how much risk reduction is being realized.

Also, with respect to SR FSS-B2, discuss to what extent the conclusion of insensitivity is dependent upon this credit.d. In the LAR Table W-1, Summary of Total Plant Risk, and in Enclosure 3 to the letter dated September 25, 2012, Item 3 on page 2, two of the sensitivity analyses (in LAR Attachment V) increase the CDF by -2E-5/yr, bringing the total close to 1 E-4/yr. Discuss whether all values here, both these values and the sensitivities, have taken some risk reduction credit for installing VEWFDS in the MCR. If total CDF becomes >1 E-4/yr or total LERF > 1 E-5/yr, address any ramifications due to these changes relative to the guidelines on total risk in RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." e. In LAR Attachment W, Table W-2, Fire Initiating Events Individually Representing at least 1% of Calculated CDF for Unit 1, discuss whether an MCR fire scenario should appear among the dominant ones. Include consideration that LAR Table W-6 indicates an MCR abandonment CDF of -3E-7/yr, which could rise to dominate the others, if credit for MCR VEWFDS is not taken.f. For Fire Areas Ul 044 and U2 044, in LAR Table W-6, FNP Fire Area Risk Summary, discuss why CDF equals LERF and delta-CDF equals delta-LERF.

Also, discuss what these values would be without the credit being taken for installation of VEWFDS in the MCR.El -3 Enclosure 1 Supplemental Responses to Previous RAI Responses g. In LAR Table W-6, FNP Fire Area Risk Summary, discuss how the total risk and delta-risk estimates, including those for RAs, would change if VEWFDS credit in the MCR were removed.Response provided by SNC letter NL-13-2269 dated November 12, 2013 b. The increase in control room risk due to the elimination of credit for incipient detection will be addressed in conjunction with additional refinements of the NUREG/CR-6850, Appendix L credit for the main control board Fire PRA quantification.

c. Credit for VEWFDS was taken prior to and after updating the analysis in the MCR using the NUREG/CR-6850 Appendix L credit. Elimination of credit for incipient detection will be addressed in conjunction with additional refinements of the NUREG/CR-6850, Appendix L credit for the main control board Fire PRA quantification.

d. The increase in risk for the scenarios in the control room which currently credit VEWFDS will be offset by further refinement of the main control board analysis using the guidance of NUREG/CR-6850, Appendix L.e. The updated analysis eliminating the credit for VEWFDS will generate updated tables including Table W-2 which may now include control room scenarios.

f. The compliant case for these areas is conservatively set to zero which results in a conservative delta risk which is equal to the variant case risk.The analysis used in the LAR submittal was based on application of a CCDP of 0.1 for calculation of CDF. The CCDP value of 0.1 was also conservatively applied to the CLERP for calculation of the LERF value.These approaches result in the same LERF to CDF ratio and same delta-LERF to delta-CDF ratio.A revised methodology for calculating the CCDP for the abandonment scenarios is discussed in the response to RAI PRA 33c. Application of this approach will differentiate between CDF and LERF. Elimination of credit taken for VEWFDS as discussed in the response to item (b) above, will increase only those scenarios crediting VEWFDS. Further refinement of the main control board analysis using the guidance of NUREG/CR-6850, Appendix L will be applied to offset the increase in risk and delta risk resulting from the elimination of incipient detection credit.g. The change in total risk and delta risk with the elimination of VEWFDS credit will be offset by further refinement of the main control board analysis using the guidance of NUREG/CR-6850, Appendix L.El -4 Enclosure 1 Supplemental Responses to Previous RAI Responses Supplemental Response: The revised risk with VEWFDS credit removed will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -5 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 01(h)In Enclosure 6 to the supplement dated December 20, 2012 (ADAMS Accession No. ML12359A051), the results are presented for both the total and delta core damage frequency (CDF) that are actually lower than previously reported in Attachment W of the LAR. The submittal, although only the credit for the electrical cabinet factor was removed. With the additional removal of credit for the main control room (MCR) very early warning fire detection system (VEWFDS), it is expected that these CDF results would increase, consistent with the increases in the large early release frequency (LERF) values. Explain why, including any key modeling assumptions that may be relevant, the increases in total and delta-CDF are now lower than before especially in light of the higher increases in total and delta-LERF.

From the submittal, address the following:

h. Calculation SE-C051326701-008, Farley Nuclear Plant, Units 1 and 2, NFPA 805 Fire Risk Evaluations, Version Number 1, dated September 25, 2012, was discussed during the site audit. In Tables 2-1a through 2-2b of the calculation, Fire PRA Variant Case Results, (Non-) Abandonment Trains A (and B) Alignment, three non-suppression probabilities are assumed -1, 0.1 and 0.02. Discuss the bases for the latter two including their use with respect to crediting VEWFDS installed in-cabinet in the MCR panels.Response provided by SNC letter NL-13-2269 dated November 12, 2013 Both non-suppression probabilities (NSP), 0.1 and 0.02, are related to the credit for VEWFDS. For the instrument rack areas (416 and 471), the 0.02 NSP is used as presented in Section 13.2 of NUREG/CR-6850 Supplement

1. These two rooms were considered physically separated from where the operators are physically located by the MCR panels themselves.

The 0.1 NSP is the assumed value applied to the scenarios postulated in the control room (401) which is continuously occupied by the operators.

The credit for VEWFDS for the control room will be removed and the related fire scenarios will be revisited in conjunction with NUREG/CR-6850, Appendix L credit as discussed in the response to RAI PRA 01a.Supplemental Response: The revised risk with VEWFDS credit removed will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -6 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 06(a)Section 10 of NUREG/CR-6850, Supplement 1, states that a sensitivity analysis should be performed when using the fire ignition frequencies in the Supplement instead of the fire ignition frequencies provided in Table 6-1 of NUREG/CR-6850. Summarize the details and provide the results of the sensitivity analysis of the impact on using the Supplement 1 frequencies instead of the Table 6-1 frequencies on CDF, LERF, delta (A) CDF, and ALERF for all of those bins that are characterized by an alpha that is less than or equal to one.Response provided by SNC letter NL-13-2269 dated November 12, 2013 Section V.2.2 of the Farley NFPA 805 LAR provides the details of the sensitivity analysis related to the bins that have an alpha that is less than or equal to one.Table V.2-3 provides the results of this sensitivity for CDF, LERF, ACDF, and ALERF. This is provided for Unit 1 only, however based on the similarities between the two units and similar baseline results it is judged that the results of this same sensitivity on Unit 2 will be comparable to that of Unit 1.This sensitivity analysis will be updated to reflect the final baseline risk and will be submitted via a supplemental RAI response or in response to a potential 2 nd round RAI, as appropriate.

Supplemental Response: The sensitivity analysis will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -7 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 06(b)Calculation PRA-BC-F-1 1-017, Joseph M. Farley Nuclear Plant, Units 1 and 2, Farley Fire PRA Summary Report, Version Number 1, dated September 14, 2012, was discussed during the site audit. With respect to p. D-3, Table D-1, Uncertainty and Sensitivity Matrix: Regarding Task 8, discuss the uncertainty analysis performed based on the updated fire frequencies from Supplement 1 of NUREG/CR-6850.

Discuss whether this went beyond just performing the sensitivity evaluation in part (a), i.e., discuss whether a parametric uncertainty evaluation was performed using distributional parameters for each bin represented in the CDF and LERF. Summarize the details and report the results.Response provided by SNC letter NL-13-2269 dated November 12, 2013 The uncertainty analysis that was completed and presented in Appendix D of PRA-BC-F-1 1-017 includes the parametric uncertainty based on the use of the ignition frequencies from Supplement 1 of NUREG/CR-6850.

The sensitivity analysis will be updated in conjunction with the update of various Fire PRA analyses to address RAI responses.

The results will be provided via a supplement to the RAI responses or in conjunction with responses to potential 2 nd round RAIs, as appropriate.

Supplemental Response: The sensitivity analysis will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -8 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 06(c)On page D-17, Section D.3.1, Fire Ignition Bin Frequencies, of the Calculation PRA-BC-F-1 1-017 (as referenced above), although it is recognized that the sensitivity analyses performed here were done on an earlier FPRA model, one would still expect the ratio of the resultant CDFs in this table (8.43E-5/4.66E-5

=1.81) to be roughly the same or even slightly lower than the corresponding ratio reported in Table V.2-3 of the LAR, namely 7.46E-5/5.24E-5

= 1.42. Similarly the delta-CDFs should be roughly the same, with the value in Table V.2-3 (2.22E-5/yr) perhaps slightly larger than the one here (3.77E-5/yr).

Neither case is true.It is recognized that the two sensitivity evaluations may be different, namely the one here addressed ALL the bin frequencies while that in Table V.2-3 addressed only the selected bins with alpha parameters of 1.0 or less. If that is the explanation, extract the results from here, with appropriate adjustments to reflect the final FPRA model, and include in the LAR with a discussion of the difference between the two sensitivity analyses.

If there is another explanation, provide it.Response provided by SNC letter NL-13-2269 dated November 12, 2013 The ignition frequency sensitivity provided in PRA-BC-F-1 1-017 Appendix D was completed on a previous model to the LAR submitted model. The results provided in Appendix D were also based on a sensitivity for all ignition frequency bins, not just those that have an alpha value of less than or equal to 1 in Supplement 1 of NUREG/CR-6850.

These frequencies were also not Bayesian updated like those used in the base model or the sensitivity in the LAR. It is for this reason that the two sensitivities are not comparable.

It should be noted that the sensitivity provided in LAR Attachment V included the Bayesian updated ignition frequency bins that had an alpha of less than or equal to 1. An updated sensitivity analysis following the guidance of NUREG/CR-6850 Supplement 1 will be provided in conjunction with the final baseline model results incorporating the resolution of all RAIs.Supplemental Response: The sensitivity analysis will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -9 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 08(a)It was stated at the 2010 industry fire forum that the Phenomena Identification and Ranking Table (PIRT) Panel being conducted for the circuit failure tests from the DESIREE-FIRE and CAROL-FIRE tests may be eliminating the credit for Control Power Transformers (CPTs) (about a factor 2 reduction) currently allowed by Tables 10-1 and 10-3 of NUREG/CR-6850, Vol. 2. Provide the results of a sensitivity analysis that removes this CPT credit from the PRA, showing the impact of this potential change on CDF, LERF, ACDF, and ALERF.Response provided by SNC letter NL-13-2269 dated November 12, 2013 A sensitivity analysis was performed as part of the Fire PRA development that included doubling all of the hot short probabilities, not just those associated with CPTs. This sensitivity is included in Section V.2.3 of the Farley NFPA 805 LAR, Table V.2-4. The results are provided here. This sensitivity was performed on a version of the model that preceded the model presented in the LAR and only using the Unit 1 Train A model. This is representative of the other unit and trains in the Farley analysis based on the type of changes that took place between the two models.Table V.2-4 Control Power Transformer Sensitivity Case Resultant Delta CDF Change CDF CDF -Base 4.66E-05 ......CDF -Double HS 4.92E-05 2.60E-06 5.28%Prob Interim guidance has since been accepted by the NRC for the treatment of hot shorts in the Fire PRA. This interim guidance is documented in Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis, dated June 14, 2013 (ML13165A209, ML13165A214).

These new probabilities also suggest that the sensitivities presented here will be bounding with respect to the anticipated change in hot short probability guidance.

This is based on the probabilities found in the interim guidance (which do not provide distinction on the presence of a CPT) actually being smaller than those found in NUREG/CR-6850 Table 10-2 and 10-4. The sensitivity above is based on all hot short probabilities from NUREG/CR-6850 Table 10-1 through 10-4 being doubled, while the new interim guidance, even for the CPT circuits, is less than what the probabilities would be if doubled. These new probabilities also suggest that the sensitivities presented here support the conclusions that the model is not sensitive to these changes. An updated sensitivity analysis using bounding values for hot short probability credit will be included in the final baseline Fire PRA quantification.

This will beprovided via a supplemental RAI submittal or in conjunction with a potential 2° round of RAIs, as appropriate.

El -10 Enclosure 1 Supplemental Responses to Previous RAI Responses Supplemental Response: The updated sensitivity analysis will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -11 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 08(b)Calculation PRA-BC-F-1 1-017, Joseph M. Farley Nuclear Plant, Units 1 and 2, Farley Fire PRA Summary Report, Version Number 1, dated September 14, 2012, was discussed during the site audit. In light of part (a), update the estimates from the table on page D-17, section D.3.2, Spurious Operation Probabilities, using the latest FPRA model, including an update to Table V.2-4 of the LAR.Response provided by SNC letter NL-13-2269 dated November 12, 2013 Since the submittal of the Farley NFPA 805 LAR, interim guidance has been accepted by the NRC for the treatment of hot shorts in the Fire PRA. This interim guidance is documented in Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis, dated June 14, 2013 (ML13165A209, ML13165A214).

An updated sensitivity analysis using bounding values for hot short probability credit (based on latest guidance available at the time) will be included in the final baseline Fire PRA quantification.

This will be provided via a supplemental RAI submittal or in conjunction with a potential 2 nd round of RAIs, as appropriate.

Supplemental Response: The sensitivity analysis will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -12 Enclosure 1 Supplemental Responses to Previous RAI Responses Farley PRA RAI 17(b)For Calculation PRA-BC-F-1 1-014, Joseph M. Farley Nuclear Plant, Units 1 & 2, Fire Scenario Development, Version Number 2, dated September 14, 2012, address the following:

a. On pages 14-1 and 14-2, Section 14.0, "Use of Generic Fire Modeling Treatments vs. Detailed Fire Modeling," potential conservatisms present in the Hughes Approach are credited as a basis for not performing detailed fire modeling.

There appears to be reliance on information available from the EPRI Fire Events Database, complete only through 2000 and currently being updated, as justification.

Discuss the potential impact on this justification in light of the fact that such data may be incomplete and, therefore, non-conservatively adapted outside the consensus approach of NUREG/CR-6850.

Response provided by SNC letter NL-13-2269 dated November 12, 2013 The discussion cited above is associated with the use of the panel split fractions for defining the fraction of the ignition frequency impacting only the ignition source panel with no impact to external targets. Use of this method was not accepted by the NRC. A sensitivity evaluation has been performed to address the elimination of this factor. The quantification of this sensitivity evaluation will be incorporated into the baseline fire quantification.

Supplemental Response: The sensitivity analysis will include the composite effect of the quantification of other RAI responses and will be provided in the response to PRA RAI 35.El -13 Joseph M. Farley Nuclear Plant 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 Enclosure 2 Response to Safe Shutdown Analysis RAI Enclosure 2 Response to Safe Shutdown Analysis RAI Farley Safe Shutdown Analysis (SSA) Request for Additional Information (RAI) 10.01 In a letter dated October 30, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13305A105), the licensee responded to SSA RAI 10 and indicated that no recovery actions (RAs) were omitted from license amendment request (LAR) Attachment G, and that the correlation of RAs to variances from deterministic requirements (VFDRs) were provided in LAR Attachment G, Table G-1, of the LAR supplement dated December 20, 2012 (ADAMS Accession No. ML12359A050).

However, the NRC staff is providing the following examples of inconsistences between: LAR Attachment C, Table C-1 and LAR Attachment G, Table G-1; with the LAR supplement and the RAI responses:

a. The LAR Attachment G, Table G-1 submitted on December 20, 2012, is missing many fire areas compared to the original LAR Attachment G, Table G-1 provided in September, 2012 (e.g., most of U2-040, U2 2-040, U2 2-041, U2 2-075, and U2 2-076... up to U2-2-021).

b. Components identified in the new LAR Attachment G, Table G-1 don't correspond to VFDRs in LAR Attachment C, Table C-1 (e.g., Q1P16V0530 and Q1P16V0593).

c. Components corresponding to VFDRs in the LAR Attachment C, Table C-1 are not identified in the new LAR Attachment G, Table G-1 (e.g., VFDRs: U1-044-PCS-040 and U1-1-040-PCS-186), and there are potential duplicate VFDRs (e.g., U1-1-040-PCS-145

&146, U1-044-PCS-127

& 128 and U2-044-PCS-079

& 155)d. A partial LAR Attachment G, Table G-1 was submitted with the RAI responses on November 12, 2013 (ADAMS Accession No.ML13318A027), which includes LAR pages G-9 through G-26. This table has corrections and multiple entries needing to be removed as duplicates.

However, this LAR Attachment G, Table G-1 still includes several components (e.g., OP-RECOV-XXXX) that are not in LAR Attachment C, Table C-1.Provide LAR Attachment G, Table G-1 and LAR Attachment C, Table C-1 that are up-to-date and correlate accordingly.

RESPONSE: a. It was determined that in the previous transmittal portions of Attachment G were inadvertently omitted. A complete replacement to Attachment G is provided, in its entirety, as an attachment to this response.

The design input is provided in the Fire Risk Evaluation document (SE-C051326701-008) where VFDR, component and categorization of DID or risk recovery action are documented.

No recovery actions are omitted in Attachment G.No changes to LAR Attachment C are required to incorporate technical E2- 1 Enclosure 2 Response to Safe Shutdown Analysis RAI updates associated with the new Attachment G. Minor administrative updates will be incorporated as an update to the source document(s).

b. The component(s) identified in LAR Attachment G, Table G-1 are component(s) made available by application of recovery action(s).

In cases where a VFDR requires a recovery action (to satisfy the risk or defense-in-depth criteria of NFPA 805, Ch. 4.2.4) alternate or redundant components are restored to meet the performance criteria.

These recovered components are identified in Attachment G, Table G-1. The components associated with the VFDR are documented in Attachment C, Table C-1. The previous LAR Attachment G tables that were submitted did not include the recovery action type to VFDR to component correlation; however, to facilitate more complete mapping this has been revised with the replacement Attachment G included in this response.c. See part b (above) to address the scope of response related to Attachment C and Attachment G component differences.

The examples of duplication have been investigated and corrected where appropriate (U1-1-040-PCS-145 was previously removed).

In some cases there are seemingly duplicate recovery actions against the same VFDR; however, the recovery is associated with a unique basic event with a different action. There are slight differences in the function states or failure modes of the same components that can be easily missed on inspection or may require review of the source documents.

See examples below for illustration of these scenarios:

BE Differences Driving Unique Recovery Actions: SE-C051326701-008 Attachment

-FRE for Unit I Fire Area 1-009 2.2.4 Required Modifications The following modifications are required as a result of the risk evaluation for hiis fire area:.Install new trip device in panel Q1R42BOO01B, breaker LB07 2.2.5 Required Recovery Actions The following recovery actions are required as a result of the risk evaluation for this fire area: Table 2-4 Required Recovery Actions Component ID Basic Event VFDR BE Description Recovery Action CompnentDescription O1R42BO1B.

EN6R ~OPERATOR RECOVERY Operator action to GIZED:ENERGRIZEDN P-RECOVo U1-1109- OF BATTERY CHARGER provide alternate cooling 60EZ 1RCBC-B HVAC-O01 1B OR 2B ROOM for battery charger room COOLING FAILS due to loss of fan OPERATOR RECOVERY Operator action to 01R42B0001B:ENER OP-RECOV-U1-1-009-OF BATTERY CHARGER provide alternate cooling GIZED:ENERGIZED-IBCSW-B HVAC-001 B ROOM COOLING FAILS- for battery charger due BC1B SW to the loss of Train B SW E2 -2 Enclosure 2 Response to Safe Shutdown Analysis RAI Function State Differences:

VFDR ID Ul-044-PCS-127 VFDR Q1R43EO001B:AVAILABLE:AVAILABLE-SEQiSHED.

SEQUENCER BUS 1G -QIR43EOOOIB

-SEQUENCER BUS 1G. This normany avaeiable, required available sequencer.

The electrical system is required to operate for various system supports.

Fire induced damage to cables in the control room and no controls on te POS prevent ability to control electrical system components, and a chalenge to lhe Vital Asetaries Nuclear Safety Performance Criteria.Disposition TiNs condition was evalualed for compliance using the performance-based approach of NFPA 80.S Section 4.2.4. A fire risk evaluation determined that recovevy actions(s) are required to meet applicable risk and DID criteria.VFDR ID U-0-44-PCS-128 VFDR O1R43E000IBAV.AILABLERAVAILABLE-SHED SEQUENCER BUS IG -O1R43E0001B

-SEQUENCER BUS 1G. This normaty avatilable, requrred available sequencer.

The elec ncal system is r d to operate toe various system supports.

Fire rnduced damage to cables in the control room and no controls on the PCS prevent ability to control electrical system components, and a challenge to the Vital Aunliaries Nuclear Safety Performance Criteria-Disposition This condition was evaluated for compliarce using the p-rormance-based approach of NF PA 805, Section 4.2.4. A fire risk evaluation determined that recovery actrons(s) are required to meet applicable risk and DID criteria.Failure Mode Differences:

VFDR ID U2-044-PCS-079 VFDR Q2B41POOO1A:ON:OFF.

RCP 2A -Q2B41POODIA-RCP 2A. This noramally on, required off pump. The RCPs are required off to remove heat generated by running pumps added to RCS and Oroti RCS inventory losses thri seals. Fire induced damage to cables and loss o1 dc control power in the contol room and no cntirols on the PCS prevent ability to control pumps, and a chatlenge to the Inventory and Pressure Control Nuclear Safety Performance Criteria.

This condition represents a variance from the deLrministc requirements of Section 4.2.3 of NFPA 805. This is a Separation Issue. Evaluate fon compliance using the performance-based approach of NFPA 805, Section 4.2.4.Disposition This condition was evablaled for conmplance using the pertormance-based approach of NFPA 805, Section 4.2.4. A tire risk evaluation determined that recovery actions(s) are reguired to Meet applicable risk and DID cemita.AFFECTED BEIGATE: U2 B31/B41-RCPPCA-ARC LOSS REASON: Loss of control Loos of power VFDR ID U2-44-PCS-155 VFDR Q2B41POO1A:ON:OFF.

RCP 2.A- 02B41POOI1A

-RCP 2A. This normally on, required off pump. The RCPs are required of4 to remove hIeat geneated by running pumps added to RCS and bimit RCS inventory losses thin seals. Fire induced damage to cables and loss of dc control power in the control room and no controls on the PCS prevent abiity to control pumps, and a challenge to the inventory and Pressure Control Nuclear Safety Performance Criteria.

This condition represents a variance from the deterministic reqrirements of Section 4.2.3 of NFPA 805. This is a Separation Issue. Evaiuate for compliance using the performance-based approach 04 NFPA 805, Section 4.2.4.AFFECTED BEIGATE: U2 TRIP-RCP-ARC LOSS REASON: Loss of control Loss of power Disposition This condition was evaluated tor compliance using the performance-based approach o0 NFPA 805. Section 4.2.4. A fire risk evaluation determined that recovery actions(s) are required to meet applicable 0io and DID criteria.d. See parts a, b and c for response to this question.E2 -3 Joseph M. Farley Nuclear Plant 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 Enclosure 3 Response to Fire Modeling RAIs Enclosure 3 Response to Fire Modeling RAIs Farley Fire Modeling (FM) RAI 01.02 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to FM RAI 01 .m and stated: "In certain areas of the plant the physical location is such that there is insufficient space for placement of a transient fire that would be equivalent to that of the 317 kW transient fire as described in NUREG/CR-6850.

For example, a hallway that may be approximately 3 feet wide would not physically allow a 317 kW fire as defined in the Generic Fire Treatments, Supplement

3. The diameter of 317 kW fire is 1.1 meters (3.6 feet). A fire of 69 kW would be associated with a fire diameter of approximately 1.5 feet which is the largest obstruction that is expected to be located in a walkway without obstructing access. Therefore, the lower heat release rate is used as a more appropriate fire size for the reduced space configuration." Provide the following information for the: "certain areas of the plant the physical location is such that there is insufficient space for placement of a transient fire..." discussed above: a. A list of the physical locations where a 69 kW transient fire was postulated based on space limitations.

b. Technical justification for each location why a 317 kW transient fire with physical dimensions that fit within the space could not be postulated.

c. A discussion of whether the Generic Fire Modeling Treatments (GFMTs) address transient configurations that are small enough to fit within the space (e.g., a hallway) without obstructing access, and that could generate a heat release rate (HRR) higher than 69 kW.d. If as a result of the response to items 2 or 3 above, locations are identified where a transient fire HRR of higher than 69 kW should have been postulated, provide a quantitative assessment of the effect of the higher HRR on plant risk (core damage frequency (CDF), delta (A) CDF, large early release frequency (LERF), and A LERF).RESPONSE: a. The Table below provides a list of all rooms where a 69kW transient fire is postulated.

A short description is also provided for why the smaller (69kW) transient fire is postulated.

Valve Approximately 6'wide, L-shaped room with valves 0152 Compartment located along the wall.Room E3- 1 Enclosure 3 Response to Fire Modeling RAIs 0202 Communication Room small pathway in room along East and North wall, remaining portion of room is taken up by equipment mounted on the ground and wall approximately 8' wide hallway, multiple doorways into/out 0210 Corridor of the hallway with piping located along the wall at the lower elevations 0211 Corridor approximately 4' wide hallway, multiple doorways into/out of the hallway 0213 Battery Service approximately 117 sq ft room, multiple doorways into/out Room of the room, one on each wall approximately 9' wide, 963 sq ft, a significant portion of the room is taken up by trays and conduits that are 0227 Cable Chase located in the room, from floor to ceiling, there is minimal floor space available for transient combustibles to be located approximately 4' wide hallway, multiple doorways into/out 0228 Corridor of the hallway, enclosed HVAC located in the middle of the room traversing vertically from floor to ceiling approximately 9' wide, 945 sq ft, a significant portion of the room is taken up by trays and conduits that are 0300 Cable Chase located in the room, from floor to ceiling, there is minimal floor space available for transient combustibles to be located 0319 Corridor approximately 4' wide hallway, multiple doorways into/out of the hallway approximately 4' wide hallway, multiple doorways into/out 0339 Corridor of the hallway, enclosed HVAC located in the middle of the room traversing vertically from floor to ceiling approximately 256 sq ft room, two doorways into/out of 0345 Hallway the room, main portion is directly in front of the elevator and stairwell exit 0465 Cable Chase approximately 9' wide, 270 sq ft, this room is only accessible via a ladder from room 0466 0466 Cable Chase approximately 9' wide, 619 sq ft, this room is only accessible via a ladder from room 0300 approximately 9' wide, 945 sq ft, this room is only accessible via a ladder from room 0466, a significant 0500 Cable Chase portion of the room is taken up by trays and conduits that are located in the room, from floor to ceiling, there is minimal floor space available for transient combustibles to be located Hot Shutdown approximately 326 sq ft room, significant portion of the Panel Room room is taken up by equipment mounted to the floor E3 -2 Enclosure 3 Response to Fire Modeling RAIs approximately 9' wide, 570 sq ft, a significant portion of the room is taken up by trays and conduits that are.hase located in the room, from floor to ceiling, there is minimal floor space available for transient combustibles to be located 2227 Cable C 2319 Corridor approximately 4' wide hallway, multiple doorways into/out of the hallway approximately 4' wide hallway, multiple doorways into/out 2339 Corridor of the hallway, enclosed HVAC located in the middle of I_ , the room traversing vertically from floor to ceiling b. The transient fire HRR values provided in NUREG/CR-6850 are based on a series of laboratory tests. A review of the tests that are the bases of the NUREG/CR-6850 recommendations found that they all had heat intensities of between 300 and 400 kW/m 2.The corresponding

'footprint' of the postulated transient fire that would have an HRR equal to the 98th percentile value would therefore have an area of approximately 1 M 2.The locations identified above, walkways and very small rooms, preclude the storage of material with more than a 1.5 ft 2 which is consistent with a 69 kW HRR transient fire. Therefore it is not considered credible for a larger transient fire to occur in these locations.

c. Supplement 3, Section 5 to the Generic Fire Modeling Treatments addresses the reduced transient fire size of 69 kW fire. The data presented in this supplement is representative of the types of transient fires discussed in this response.

The 3' X 3' fuel package fits in all areas where a fire greater than 69 kW is postulated.

d. There were no locations where a larger fire was required as a response to items 2 or 3. Therefore, no qualitative assessment is needed.E3 -3 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 01.03 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to FM RAI 01.p and stated: "The transformers cited in the RAI are filled with Dow Corning 561 Transformer Fluid, which is a dimethyl silicone insulating material for power transformers that has a substantially reduced fire hazard potential than mineral oil insulating materials.

The fluid has a Heat Release Rate (HRR) of 140 kW/m2 per ASTME 1354-90, which is approximately 11 times less than that of mineral oil. Based on the burning and ignition characteristics of the Dow Corning 561 Transformer Fluid observed in the pool fire tests (i.e., it is difficult to ignite, it produces short flame heights, and it self-extinguishes), the transformers containing silicone are considered to be similar to a dry transformer ignition source rather than a mineral oil filled transformer." Sections of NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities" (such as Chapter 8, Table 8-1 ID 23b, Chapter 11, Table 11-1), indicate that the severity factor of a motor fire can be used to characterize a dry transformer.

However, depending

'on the area of the spill, the HRR due to a pool fire from these transformers could be larger than a standalone motor fire.a. Describe and provide technical justification for the HRR and assumed fire source area that were used to characterize transformers filled with the Dow Corning 561 Transformer Fluid.b. Explain how this may impact the plant risk, if the developed fire grows larger than what was originally assumed (i.e., area greater than 0.5 m2 for the stated HRR per unit area).RESPONSE: Part a The technical justification for the assumed heat release rate is based on the properties of the Dow Corning 561 silicone liquid as described in the response to RAI FMOD 01 (p). There are three key material properties that substantially reduce the hazard of the silicone liquid relative to conventional hydrocarbon transformer oils. The key properties that are applicable to the transformer ignition sources are as follows (see "Fact Finding Report on Flammability of Less Flammable Liquid Transformer Fluids -Project No. 87NK1 7807 and Buch, R. R., Rates of Heat Release and Related Fire Parameters for Silicones, Fire Safety Journal, Volume 17, 1991): 1. The silicone liquid has a fire point that is greater than 340 0 C (6440F). This attribute suggests that the silicone liquid is difficult to ignite and requires high flame heat fluxes to sustain combustion.

2. The silicone liquid forms a silicon dioxide crust during combustion that causes the fire to self-extinguish.

This property limits the duration of a E3 -4 Enclosure 3 Response to Fire Modeling RAIs potential fire involving liquid spill and may prevent a fire from consuming the full inventory of the liquid spill.3. The heat release rate per unit area of the silicone liquid is 140 kW/m 2 (12.3 Btu/s-ft 2), which is about ten percent the heat release rate per unit area for conventional hydrocarbon transformer oils per NUREG-1805.

This property indicates that the flames that may develop on ignited liquid will have a lower power output, which reduces the potential for the formation of a hot gas layer, reduces the flame and thermal plume exposure to targets located above the burning liquid, and reduces the energy available for pyrolizing and sustaining combustion at the liquid surface relative to a conventional hydrocarbon transformer oil.Although the available test data on the silicone liquids indicate that the material is capable of supporting a fire when a sufficiently large ignition source is applied, there is a precedent for qualitatively crediting the reduced fire hazard potential for indoor transformer installations at commercial nuclear power plants. A recent example involves an exemption request for the use of operator manual actions in four fire areas at a commercial nuclear power plant (see ML110691282 and ML110700150).

The silicone oil is the predominant combustible in one of the fire areas and a significant fuel load in two other fire areas. The basis for the exemption involves the relatively low fire hazard of the silicone fuel and in particular the three key material properties previously cited.The basis for linking the heat release rate, and thus the zone of influence (ZOI), to that of a dry transformer with a 9 8 th percentile peak heat release rate of 69 kW (65 Btu/s) is that the silicone liquid would not contribute significantly to the fire hazards based on the key material properties cited above. Essentially, it is postulated that a fire at the silicone liquid transformers would be generally confined to the transformer itself, which is a comparable event to a dry transformer fire.A review of technical specifications and fire protection guidance for silicone liquid transformers provides qualitative support for this treatment:

Factory Mutual Data Sheet (FMDS) 5-4 addresses fire protection requirements for transformers with the intent of minimizing property losses. Per Section 2.2.1.8 and Tables 4 and 5 of FMDS 5-4, indoor transformers with factory mutual approved transformer liquids in factory mutual approved transformers may be located within 0.9 m (3 ft) of combustible walls. Given the range of potential combustible wall materials, this suggests that the horizontal ZOI for the transformers is expected to be on the order of 0.9 m (3 ft), which is comparable to the ZOI dimensions for the 69 kW (65 Btu/s) dry transformer.

A factory mutual approved transformer liquid is one in which the fire point is greater than 300'C (572°F), which the Dow Corning 561 silicone liquid meets.* International Standard IEC 60695-1-40 "Fire Hazard Testing: Guidance for Assessing the Fire Hazard of Electro Technical Products -Insulating Liquids," Section 7.1 states that there are no reported pool fire incidents involving transformers containing Class K liquids based on 150,000 (European) transformers with Class K liquids that are in service. Per Section 4, a Class K liquid is one in which the fire point is greater than 300'C (572'F), which the Dow Corning 561 silicone liquid meets.E3 -5 Enclosure 3 Response to Fire Modeling RAIs Based on the above data, a fire scenario at any of the transformers at the Farley Nuclear Plant containing the Dow Corning 561 silicone liquid is expected to be confined to the transformer itself rather than result in a spreading pool fire. The most appropriate ignition source bin for this type of fire scenario is the dry-type indoor transformer, which per NUREG/CR-6850 has a 98th percentile peak heat release rate of 69 kW (65 Btu/s).Part b If the fire grows larger than originally postulated, the ZOI would increase and the hot gas layer temperature would reach threshold values at shorter time intervals.

This is not unique to the transformer ignition source, but applies to any fire scenario.

In the case of the silicone liquid transformers, it is postulated that the fire remains confined to the transformer itself as discussed in Part a of this RAI response.

This is based on the properties of the silicone material and on the fire hazard guidance provided in IEC 60695-1-40 for Class K insulating liquids. As such, a pool fire having a peak heat release rate per unit area of 140 kW/m 2 (12.3 Btu/s-ft 2) is not associated with this ignition source, regardless of the spread area.E3 -6 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 01.06 In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019), the licensee responded to FM RAI 01 .e and stated: "The sensitivity analysis.. .is not directly used in the FPRA. The sensitivity analysis.. .provides an indication of the parameters selections that could lead to significant variations in the results with the intent that adjustments to the baseline scenarios be made on a case-by-case basis. The current FPRA uses only the baseline scenarios and therefore does not directly incorporate the insights provided in the sensitivity analysis of ..." In addition, the licensee stated that, "A baseline fire scenario is considered to be conservatively biased if the total probability of control room abandonment is maximized for the baseline fire scenario.

A baseline fire scenario is considered insensitive if the change in the total probability of control room abandonment remains less than fifteen percent. A baseline fire scenario is considered to be non-conservatively biased if the change in the total probability of control room abandonment exceeds fifteen percent." Provide technical justification for this 15% limit criterion and explain how it was determined (as opposed to a lower value, such as 5% or 10%).RESPONSE: The selection of a fifteen percent criterion for determining whether or not uncertainty in a parameter value has a significant effect is based on the theoretical and observed uncertainty in calorimeter heat release rate measurements as described in the SFPE Handbook of Fire Protection Engineering, Section 3-2. The theoretical uncertainty in the heat release rate measurements is reported as +/-7 -12 percent, depending on the type of carbon monoxide correction employed.

The observed uncertainty among test facilities is between 17 -23 percent. Because the heat release rate is a primary input parameter provided by NUREG/CR-6850, the effect of the uncertainty in the output parameters is resolved to a level comparable to that of the heat release rate input parameter.

When fifteen percent limit criterion is implemented in the control room abandonment parameter sensitivity analysis, it is applied only to the conservative (upper) limit. A parameter variation is thus allowed to produce a non-conservative result by a factor that exceeds fifteen percent, which implies the baseline assumption is conservative by more than fifteen percent. Consequently, the cumulative effect of all parameter variations produces a baseline configuration that is conservatively skewed so that the parameter uncertainty is bound.When viewed in context of the overall analysis uncertainty, the parameter uncertainty range of fifteen percent applicable to any given input parameter considered in Attachment 2 of Report 0005-0003-003-001

("Evaluation of Control Room Abandonment Times at the Farley Nuclear Power Plant") is significantly narrower than the uncertainty in the probability of control room abandonment attributed to uncertainty in the suppression rate parameter (i.e., A), which is the means by which the fire PRA assesses the effect of control room abandonment E3 -7 Enclosure 3 Response to Fire Modeling RAIs time uncertainty.

This may be shown by comparing the range of control room abandonment times that could result from parameter uncertainty to the range of control room times that could result from uncertainty in the suppression rate parameter.

The total probability of control room abandonment for a particular fire scenario is determined using the following equation per NUREG/CR-6850:

Pab = Z SFi "Pns, (FM 01.06-1)where Pab is the probability of control room abandonment, SF is the severity factor for the ignition source heat release rate bin, i is the ignition source heat release rate bin number, and Ps,i is the probability that the fire associated with the ith ignition source heat release rate bin will not be suppressed before abandonment occurs (i.e., probability of non-suppression).

For a specific ignition source, the severity factor array is constant; thus, the variation in the probability of control room abandonment is entirely reflected in the variation of the probability of non-suppression.

This means that the probability of control room abandonment for each ignition source as a function of time is proportional to probability of non-suppression as a function of time. The probability of non-suppression is computed using the following equation per NUREG/CR-6850:

Pns,i = MAX(0.001,exp(-Att)) (FM 01.06-2)where A is a suppression rate parameter and tL is the abandonment time for the fh ignition source heat release rate bin (min). Per NUREG/CR-6850 and NUREG/CR-6850, Supplement 1, the mean value for this parameter is 0.33 min 1 for fires in the control room however, the fifth percentile value is 0.15 min' and the ninety-fifth percentile value is 0.58 min-1.Similarly, the suppression rate parameter is equal to 0.126 min-' for transient fire scenarios located outside the control room per NUREG/CR-6850, Supplement 1, with a fifth percentile value estimated to be 0.086 min" and a ninety-fifth percentile value estimated to be 0.166 min 1 based on the range provided in Appendix P, Table P-2 of NUREG/CR-6850.

Finally, the suppression rate parameter is equal to 0.102 min 1 for electrical panel fire scenarios located outside the control room per NUREG/CR-6850, Supplement 1, with a fifth percentile value estimated to be 0.082 min" and a ninety-fifth percentile value estimated to be 0.122 min-' based on the range provided in Appendix P, Table P-2 of NUREG/CR-6850.

Figure FM 01.06-1 graphically depicts the probability of non-suppression for fires in the control room over the 5 th -9 5 th percentile uncertainty range for abandonment times between one and twenty-five minutes. Figure FM 01.06-1 also depicts the fifteen percent variation in the mean value over the same interval.

Figures FM 01.06-2 and FM 01.06-3 graphically depict the probability of non-suppression for transient and electrical panel fires in the electrical equipment area of the control room envelope.

In all cases, the fifteen percent variation and the uncertainty in the suppression rate parameter are constants, therefore the figures of the non-suppression probability are indicative and proportional to of the total probability of control room abandonment for a given ignition source. Figure FM 01.06-1 shows that the uncertainty window the MCR parameter sensitivity analysis is resolved to is much narrower than the uncertainty in the probability of control room abandonment arising from uncertainty in the suppression rate parameter for fires in the control room, which is the means by which the fire PRA assesses the effect of uncertainty in the control room abandonment times. Figures FM 01.06-2 and E3 -8 Enclosure 3 Response to Fire Modeling RAIs FM 01.06-3 indicate that the parameter uncertainty is largely bound by the uncertainty in the suppression rate parameter, except for very short times which do not arise in the analysis for fires outside the control room. The uncertainty range is less than that for fires in the control room; however, the maximum non-conservative bias for fires outside the control room any parameter considered in Attachment 2 of Report 0005-0030-003-001 is 4.55 percent, which is much smaller than the fifteen percent window.Therefore, the basis for the fifteen percent threshold for defining a significant parameter uncertainty effect is the uncertainty in the measured heat release rate input parameters.

Because the input heat release rates may have a variation of 7-23 percent, a value of fifteen percent is selected as intermediate and typical for large scale heat release rate measurements.

The baseline cases are skewed with a conservative factor within this uncertainty range. When viewed in context of the overall uncertainty in the control room abandonment times, the parameter uncertainty range of fifteen percent applicable to any given input parameter is significantly narrower than the uncertainty in the probability of control room abandonment attributed to uncertainty in the suppression rate parameter (i.e., At), which is the means by which the fire PRA assesses the effect of control room abandonment time uncertainty.

Mean suppression rate parameter (0.33 min-1)5 e percentile suppression rate parameter (0.15 min")95th percentile suppression rate parameter (0.58 min')-- -Fifteen percent uncertainty range for parameter sensitivity C 0)0.0. 0.1 F 0 z 0.01 2 0.001 0 5 10 15 20 25 Control Room Abandonment Time (min)Figure FM 01.06-1 -Probability of Non-Suppression for Control Room Fires (Proportional to the Probability of Control Room Abandonment for a Given Ignition Source).E3 -9 Enclosure 3 Response to Fire Modeling RAIs Mean suppression rate parameter (0.126 min" 1)5!h percentile suppression rate parameter (estimated, 0.086 min")9 5 th percentile suppression rate parameter (estimated, 0.166 min 1)---Fifteen percent uncertainty range for parameter sensitivity 0.a.CL CO 0 z 0 45 IL 1 0.1 0.01 0.001 0 5 10 15 20 25 Control Room Abandonment Time (min)Figure FM 01.06-2 -Probability of Non-Suppression for Transient Ignition Source Fire Scenarios Located in the Equipment Room (Proportional to the Probability of Control Room Abandonment for a Given Ignition Source).E3 -10 Enclosure 3 Response to Fire Modeling RAIs Mean suppression rate parameter (0.102 min")S5h percentile suppression rate parameter (estimated, 0.082 min-)95h percentile suppression rate parameter (estimated, 0.122 min-)--- Fifteen percent uncertainty range for parameter sensitivity 1 0 U)I-C. 0.1 0 z o6 0.01 0..0 0.00 0 5 10 15 20 25 Control Room Abandonment Time (min)Figure FM 01.06-3 -Probability of Non-Suppression for Electrical Panel Fire Scenarios Located in the Equipment Room (Proportional) to the Probability of Control Room Abandonment for a Given Ignition Source).E3 -11 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 01.07 FM RAI 01.1 requested the licensee to assure that non-cable intervening combustibles were not missed and to provide information on how intervening combustibles were identified and accounted for in the fire modeling analyses and the FREs. In a letter dated November 12, 2013 (ADAMS Accession No.ML13318A027), the licensee responded to FM RAI 01.1 and explained that during additional walkdowns, non-cable intervening combustibles have been identified, and that the quantities of the combustibles (primarily insulation materials) is limited and does not impact the current scenario quantification.

The licensee further stated that it is anticipated that the balance of walkdowns will yield similar results and that the results from the balance of the intervening combustible walkdowns will be assessed upon completion and the impact will be incorporated into the analysis in conjunction with the impact of the secondary cable combustibles addressed under probabilistic risk assessment (PRA) RAI 17.b.The staff has reviewed the licensees responses to PRA RAI 17b and also FM RAI 1 .h (which was referenced in the licensees response to PRA RAI 17b) and concluded that additional information is required to complete the review. Provide a list of the fire scenarios with non-cable intervening combustibles and explain how the contribution of non-cable intervening combustibles was accounted for in the zone of influence (ZOI) and hot gas layer (HGL) calculations.

Also discuss the impact on plant risk (CFD, A CDF, LERF and A LERF) of the fire scenarios that involve the non-cable intervening combustibles that were identified in those walkdowns.

RESPONSE: The walkdowns performed to identify non-cable intervening combustibles, including insulation materials, were done at the same time as the walkdowns to identify intervening cable combustibles.

These walkdowns did not identify any fixed non-cable intervening combustibles of a significant quantity/size that were within the zone of influence of an ignition source (the zone of influence for cable damage was used as a conservative zone of influence for ignition of non-cable intervening combustibles).

These walkdowns are documented in report 0005-0030-003-006, Walkdowns:

Expanded Zone of Influence Target Walkdowns, Rev 0. Non-cable intervening combustibles that are enclosed within a pump (pump lube oil) or are contained within a metal enclosure (paper inside a filing cabinet)were not included since they would be exposed to limited radiant energy and are not likely to ignite and therefore would not contribute to an increase in the heat release rate for the initiating fire scenario.

Should these materials ignite, they are expected to have little impact on components/cables outside of their enclosure given the limited supply of air available for the fuel. This configuration would limit the energy released inside the metal enclosures.

In addition, there is no direct means of energy transport from the enclosure to the surroundings; it requires a conduction process through the enclosure boundary and convection and thermal radiation process at the enclosure surface. This restricts further the rate of energy transfer from the ignited contents to the surroundings.

The non-cable intervening combustibles did not impact any existing scenarios or warrant creation of new scenarios, therefore, the non-cable intervening combustibles did not impact the Fire PRA risk.E3 -12 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 02.01 In letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to FM RAI 02.f and stated: "This approach is supported by the in-process Fire PRA Frequently Asked Questions (FAQ 13-0004, "Clarifications Regarding Treatment of Sensitive Electronics.")

Walkdowns to identify sensitive electronics components which are located outside of a panel enclosure are in progress.

Sensitive electronics credited for post fire shutdown which are located outside of a panel enclosure will be evaluated with respect to potential damage by ignition sources in its vicinity using the NUREG/CR-6850 Appendix H criteria for solid-state control components." Fire PRA Frequently Asked Question (FAQ) 13-0004, which has now been issued, provides a few limitations to applying this methodology:

a. FAQ-13-0004 states cable damage thresholds can be used for temperature sensitive equipment inside cabinets provided that (i) the sensitive electronic component is not mounted on the surface of the cabinet (front or back wall/door) where it would be directly exposed to the convective and/or radiant energy of an exposure fire, and (ii) the presence of louvers or other typical ventilation means do not invalidate the guidance provided.

Describe the limitations that were considered in the determination of damage condition for sensitive electronic equipment enclosed in cabinets and explain whether they are in accordance with the FAQ or some other method.b. The conclusions of the Fire Dynamic Simulator (FDS) analysis in FAQ-1 3-0004 are based on radiant heat flux exposure to the cabinet. Therefore, the 650C temperature damage criterion must still be assessed for other types of fire exposures to the enclosed sensitive electronics.

Describe what temperature damage criterion was assessed and whether it is in accordance with the FAQ or some other method.RESPONSE: a. Walkdowns were performed for identification of sensitive electronics mounted outside of electrical cabinets.

Sensitive electronic components inside a cabinet mounted on the front or back wall/door of the cabinet are not considered to be directly exposed to the convective and or radiant energy of an exposure fire as long as they are provided with a cover or face plate at or near the panel surface which protects the sensitive electronic components within the panel from exposure to convective and or radiant energy. For panels that have louvers that may contain sensitive electronics, the fire induced damage to these would be included in the Hot Gas Layer scenario.

If the panel were to be located within the Zone of Influence of the exposing ignition source, the cabinet would be failed regardless if louvers were present or not. The guidelines of Fire PRA E3 -13 Enclosure 3 Response to Fire Modeling RAIs FAQ 13-0004 were used to define the scope and criteria applied during the walkdowns.

b. The potential impact of a hot gas layer on remotely located panels (panels outside the zone of influence of the ignition source) is addressed in the Farley Fire PRA. The hot gas layer analysis is based on an 80'C hot gas layer temperature which has been defined as the bounding temperature for validity of the Generic Fire Treatment zone of influence.

A non-suppression probability is applied to the ignition frequency for the hot gas layer scenario (scenario impacting all targets within an enclosed volume, single fire zone or multiple fire zones with openings between the fire zones) based on the time required to reach a hot gas layer temperature of 80°C and the non-suppression probability associated with that timeframe (NSP for 80'C hot gas layer). When an 80'C hot gas layer temperature has been exceeded in the compartment all sensitive electronics including all cables and components in the enclosed volume are assumed to be damaged. At and below the 80'C hot gas layer temperature, sensitive electronics located inside of a electrical cabinets are considered to be undamaged, given that the difference between the 80'C hot gas layer temperature and the 65°C sensitive electronics damage temperature is accounted for by the temperature difference between the hot gas layer (which is typically at the upper elevations of the compartment) and the temperature within the enclosed panel. This temperature difference is expected to be significantly higher than the 15'C temperature difference between the 80'C criteria and the 65°C sensitive electronics damage criteria.

This is due to the difference in elevation of the hot gas layer versus the panel internals, the thermal gradient associated with hot gases in the compartment, convection boundary condition at the surface of the panel, convection boundary condition inside the panel. All of these factors result in a temperature reduction between the cabinet internals and the hot gas layer. This temperature difference is expected to be significantly larger than 15'C which ensures that sensitive electronics located within the electrical cabinet are not damaged by the hot gas layer.E3 -14 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 02.02 In letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to FM RAI 02.g and stated: "Sensitive electronics are typically not located outside of panel enclosures as well as enclosures which protect the electronic modules from the impact of external environments such as dust. As noted in the response to RAI FM 02(f), walkdowns to identify sensitive electronics are being performed to identify electronic components mounted outside of panel enclosures.

Such components relied upon for post fire shutdown will be further evaluated, as discussed in RAI FM 02.f, to ensure their availability post fire." Provide the results of the additional walkdowns and confirm that the findings, if any, have been incorporated into the FM and PRA analysis.RESPONSE: The walkdowns performed to identify exposed sensitive electronics were done at the same time as the walkdowns to identify intervening cable combustibles.

These walkdowns are documented in report 0006-0030-003-006, Walkdowns:

Expanded Zone of Influence Target Walkdowns, Rev. 0. No sensitive electronics associated with equipment credited in the Fire PRA analysis were located outside of panel enclosures.

Therefore, the results of these walkdowns had no impact on the Fire PRA analysis.E3 -15 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 07 National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, Section 2.4.3.3, on acceptability states: 'The PSA approach, methods, and data shall be acceptable to the AHJ." The staff has noted the utilization of a number of accepted tools and methods in the analyses for transition such as the Consolidated Model of Fire Growth and Smoke Transport (CFAST) and GFMTs approach.a. Identify any fire modeling tools and methods that have been used in the development of the NFPA 805 LAR that are not already documented in the LAR and where their use or application is documented.

Examples might include a methodology (empirical correlations and algebraic models) used to convert damage times for targets in Appendix H of NUREG/CR-6850 to percent damage as a function of heat flux and time or supplements to the GFMTs -Empirical Correlations and Algebraic Models.b. For any tool or method identified in "a." above, provide the Verification and Validation (V&V) basis if not already explicitly provided in the LAR (for example in LAR Attachment J).RESPONSE: a. The elimination of the "panel factors" method for which a sensitivity evaluation was documented in LAR Attachment V,Section V.2.1 (Electrical Cabinet Fire Severity Methodology) resulted in additional refinement of the Fire PRA model. The refinements applied included a more detailed evaluation of severity factors with respect to fire impact to the first external target located outside of an electrical panel. This approach includes the time to damage of the target, which utilizes the damage delay criteria provided in NUREG/CR-6850 Section H.1.5.2. This time to damage data is incorporated in the evaluation of damage time due to the impact of a hot gas layer. This approach, which is addressed in detail in the RAI FM 10 response, is an application of the data provided in NUREG/CR-6850 and the Generic Fire Modeling Treatments.

No other refinements of significance were applied beyond those already documented in the LAR.b. No methods requiring Verification and Validation beyond those addressed in the Farley NFPA 805 LAR Attachment J have been used. The use of the NUREG/CR-6850, Appendix H data described in the response to item"a" above was verified in the same manner as other Fire PRA supporting calculations.

No validation of this method is required since it is based on data specified in NUREG/CR-6850.

Further discussions of the basis for this method are provided in the response to RAI FM 10.E3 -16 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 08 Explain how high energy arcing fault (HEAF) initiated fires were addressed in the HGL and Multi Compartment Analysis (MCA) and provide technical justification for the approach that was used to calculate HGL development timing. More specifically, confirm if the guidance provided in NUREG/CR-6850, pages 11-19, fourth bullet regarding the fire growth, and the guidance provided on page M-13, sixth bullet regarding delay to cable tray ignition was followed.

Also, considering the energetic nature of the HEAF event, provide justification for the HRR used in the HGL calculations for electrical cabinet fires following a HEAF event.RESPONSE: The HEAF initiated fires are currently addressed in the Farley Fire PRA, including scenarios for HGL and MCA. The current approach uses the Zone of Influence (ZOI) based on NUREG/CR-6850 Appendix M. The heat release rate (HRR) that is currently used is based on a medium voltage switchgear (MVSG) or Load Center (LC) as applicable in conjunction with the HRR associated with any secondary combustibles that may be part of the applicable ZOI. These HRRs are defined in Report 0005-0030-003-002, Combined Ignition Source -Cable Tray Fire Scenario ZOls for Farley Nuclear Power Plant Applications Rev 1.The current HGL and MCA uses the fire growth of 12 minutes, which is associated with Electrical Cabinet fires as defined in NUREG/CR-6850 Appendix G.To account for the guidance provided in NUREG/CR-6850 Appendix M, fourth and sixth bullet, the current analysis will be updated to apply the peak HRR to the HEAF scenarios at t=0. The HRR in these cases will include that of the ignition source (MVSG or LC) and any applicable secondary combustibles.

This update to the methodology will be applied to both the HGL and MCA, and the results of this incorporation will be provided with the response to RAI PRA 35.E3 -17 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 09 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to FM RAI 01 .h and stated: "In lieu of demonstrating which scenarios are conservative and which require further analysis, new ZOI tables have been developed that are applicable to ignition source-cable tray configurations at Farley." In addition, the licensee stated that 'The method used to develop the ZOI dimensions includes the vertical cable tray stack propagation model described in NUREG/CR-6850, Appendix R, the FLASH-CAT calculation method described in NUREG/CR-7010, Volume 1, and the radiant heat flux calculation methodology described in the GFMT document." Describe the methodology used in the revised analysis to determine the ignition time of the first cable tray above an ignition source. If it was assumed that the lowest cable tray in a stack located above an ignition source will not ignite unless the tray is located below the flame tip of the ignition source fire, provide technical justification for this assumption and provide the results of a sensitivity analysis to demonstrate that the conservatism of the ZOI and HGL calculations for fires that involve cable trays as secondary combustibles is not adversely affected by the ignition criterion that was used (compared to the ignition criteria in NUREG/CR-6850 and NUREG/CR-7010, "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire").RESPONSE: The bottom cable tray in a cable tray stack is ignited one minute after the ignition source ignites in the FLASH-CAT calculations used to support the Zone of Influence (ZOI) and Hot Gas Layer (HGL) tabulations provided in Report 0005-0030-003-002, Rev. 1 and Report 0005-0030-003-003, Rev. 0 (see Assumption 18 in Report 0005-0030-003-002, Rev. 1, for example).

This corresponds to the minimum damage time for thermoset cable targets listed in Tables H-6 and H-8 of NUREG/CR-6850 and is more conservative than the generic value of five minutes that is assumed in NUREG/CR-7010, Volume 1.The physical basis for the assumption that the lowest cable tray in a stack located above an ignition source ignites if it is at or below the flame tip is the full scale test data available for cable tray ignition and propagation data. There are three basic test series upon which the empirical flame spread model for cable tray fires as provided in NUREG/CR-6850 and validated in NUREG/CR-7010, Volume 1 is based. These are as follows: " EPRI-NP-1881 (Sumitra tests);* NUREG/CR-0381 (Klamerus tests); and* NUREG/CR-7010, Volume 1 (NIST tests).The test reports for these test series document the results of about thirty-five to forty open configuration, unprotected cable tray fire tests. In all cases, the initial E3 -18 Enclosure 3 Response to Fire Modeling RAIs ignition source for the lowest cable tray within a stack is a gas burner or liquid fuel pan fire that causes flame impingement on the lowest cable tray in the stack.Further, there are no cases presented in which the thermal plume above the flame tip alone was sufficient for igniting a cable tray. An indication of this effect may be observed in the test series presented in NUREG/CR-0381, Test 28, which was a two tray stack with ceramic blanket on the top of both trays. The propane burner was sufficiently large to ignite the lower tray and to expose the upper tray to the thermal plume during the exposure fire cycle, but the upper tray did not ignite. A quantitative indication of the conditions necessary for fire ignition and surface spread is provided in NUREG/CR-5384.

Burn mode evaluations for both non-rated (thermoplastic) and low flame spread (thermoset) cables are presented and indicate that for thermoplastic cables, which bound the results for thermoset cables, a surface temperature of 5380C (1,000°F) and an internal fuel temperature of 577 0 C (1,070 0 F) are necessary for surface flames to develop (see Figure FM 09-1). Smoldering and pyrolysis occur at lower temperatures, and a deep seated fire may result if the internal fuel temperature is approximately 5380C (1,000°F) regardless of the surface temperature.

1* 4000SL 1200 OJ 10 200 0 0 0101010101

-6000 400 200 *wT 00 200 400 600 800100012001400160018 00 FUEL NTEF" TEW ('F)Figure 3.5: Burn Mode Analysis of Non-Rated Cable Fire Figure FM 09-1 -Burn Mode Analysis of Thermoplastic Cables per NUREG/CR-5384.

E3 -19 Enclosure 3 Response to Fire Modeling RAIs It can be shown using the Heskestad flame height correlation, the Heskestad virtual origin correction, and the Heskestad plume centerline temperature correlation that the temperature at the flame tip is approximately equal to 5280C (983 0 F), which is lower than the minimum temperature of 5770C (1,070 0 F)needed for surface flame spread per Figure FM 09-1. The Heskestad flame height correlation is given as follows per Section 2-1 of the SFPE Handbook of Fire Protection Engineering.

Lf = -1.02D + 0.235Q 0°4 (FM 09-1)where Lf is the flame height (m), D is the effective fire diameter (m), and 0 is the total heat release rate of the ignition source (kW). The Heskestad plume centerline temperature correlation is given as follows per Section 2-1 of the SFPE Handbook of Fire Protection Engineering and "Fire Plumes and Ceiling Jets" in the Fire Safety Journal, Vol. 11, Nos. 1 & 2: T, = T.,o + 22Q0 2/3 (Z -Zo)-5 1 3 (FM 09-2)where T, is the plume centerline temperature (0C) at an elevation Z (m) above the fire base, Too is the initial temperature (0C), and zo is the height of the virtual origin below the fire base (m). The virtual origin height is given by the following equation per Section 2-1 of the SFPE Handbook of Fire Protection Engineering zo = -1.02D + 0.083 0 0.4 (FM 09-3)where all terms have been defined. At the flame tip, the height above the fire base, Z, is equal to the flame height, Lf. Combining Equations FM 09-1, FM 09-2, and FM 09-3 results in the following:

Tc = T,, +

= Tcx + 508 (FM 09-4)where all terms have been defined. The plume centerline temperature is thus independent of both the fire diameter and the heat release rate at the flame tip and is equal to 528°C (983'F) for an ambient temperature of 200C (68 0 F).Ambient air temperatures greater than 690C (156 0 F) would be necessary to cause the peak plume temperature to exceed the minimum value of 5770C (1,070°F) necessary for surface flames to develop. There are no plant areas in which the ambient air temperature is greater than 690C (156°F); thus, the assessment is generally applicable.

These calculations are consistent with the observation provided in Section 7.2 of NUREG/CR-7010, Volume 1 that the damage threshold for cables as characterized by a heat flux is not a good indicator of ignition.

Based on the cone calorimeter tests summarized in NUREG/CR-701 0, Volume 1, a heat flux of exposure of 25 kW/m 2 (2.2 Btu/s-ft 2) is minimally sufficient to cause ignition and sustained burning for all classes of cables considered, including the thermoplastic cables which bound the results of thermoset cables. Data provided in Section 2-14 of the SFPE Handbook of Fire Protection Engineering indicates that the net heat flux to an object immersed in the fire plume at the flame height as E3 -20 Enclosure 3 Response to Fire Modeling RAIs determined from the stagnation point is between 5 -15 kW/m 2 (0.44 -1.32 Btu/s-ft 2), which is significantly less than the minimum heat flux necessary to cause sustained ignition per NUREG/CR-7010, Volume 1. This is further supported by the test data for thermoplastic cables provided in Figures 10 -12 of NUREG/CR-6931, Volume 3. A shroud temperature (exposure temperature) of about 300 -3300C (572 -626-F) is used to heat various types of thermoplastic cables in order to determine the damage times. Although the focus of the tests was not on the ignitability of the cables, the temperature profiles provide an indication of the cable response to the temperature exposure.

The figures indicate that the cables do not ignite over the ten to twenty minute exposure interval and typically show the cables reach a steady state temperature close to 3000C (572 0 F) even though damage via electrical short occurs around 200 0 C (392 0 F).The requirement for flames to impinge on the lowest cable tray before ignition is assumed is considered reasonable and supported by the available documents.

A summary of the basis is as follows:* All cable tray fire test data involves an ignition source that results in flame impingement on the lower cable tray in a cable tray stack;* A burn mode analysis of thermoplastic cables suggests that the minimum temperature required for surface flames or deep seated burning to develop is approximately 5380C (1,000°F) for thermoplastic cables, which bound the results for thermoset cables; and* Ambient air temperatures greater than 690C (1 56 0 F) would be necessary to cause the peak plume temperature to exceed the minimum value of 577°C (1,070 0 F) necessary for surface flames to develop. There are no plant areas in which the ambient air temperature is greater than 690C (156 0 F); thus, the assessment is generally applicable.

Note that the flame height is used as an indicator of ignition and the ZOI dimensions are used as an indicator of damage. Consequently, cable trays located above an ignition source may be above the flame tip but still be within the ZOI for cable damage.E3 -21 Enclosure 3 Response to Fire Modeling RAIs Farley FM RAI 10 In letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to PRA RAI 20.c and stated: "The approach discussed in Assumption 7 of the Sensitivity Analysis uses the data provided in NUREG/CR-6850 Appendix H for time to damage as a function of heat flux to define an accrual of damage based on the time at each heat flux. The value is taken from the fire model at a given distance and correlates that to a fraction of the accrued damage by dividing the time at the heat flux by the time at that heat flux required to cause damage to the cable. Cable damage occurs when the accrued damage equals 1.0. This approach uses the same principles that are applied to equipment qualification of safety related equipment including cables for post-accident environments, such as inside containment LOCA conditions.

The potential for some non-conservatism arises from the data specified in Appendix H where no damage is accrued regardless of the time exposure when the heat flux is just below the damage heat flux. To eliminate this potential non-conservatism, the analysis is being updated to assume a bounding damage accrual during the time period prior to the cable reaching the critical heat flux." The methodology assumes that the "damage rate" at a specified heat flux as the reciprocal of the failure time in Tables H-7 (for thermoset cable targets) and H-8 (for thermoplastic cable targets) of NUREG/CR-6850.

There does not appear to be a physical basis for this assumption.

Provide evidence of the validation of the methodology as a whole, and the damage rate assumption in particular.

RESPONSE: The methodology used to evaluate cable percent damage is based on the use of the time to damage data provided in Appendix H of NUREG/CR-6850 and applying an Arrhenius methodology, which is used extensively for environmental qualification (EQ) of components such as cables in a containment accident environment, to determine the time to damage of the cables. An NRC internal evaluation of the Arrhenius methodology for equipment qualification is provided in a February 24, 2000 NRR Memo from Samuel J. Collins to Ashok Thadani (ML003701987).

The NUREG/CR-6850 Appendix H, Table H-7 data provides times to target damage for thermoset cables for a set of steady state incident heat flux values.This provides a time delay for target damage beyond the damage heat flux of 11 kW/m 2.For instance, Table H-7 provides a 19-minute time-to-damage delay for a thermoset cable with a steady state incident heat flux of 11 kW/m 2.In order to apply the NUREG/CR-6850 Appendix H data to a fire with a t 2 growth rate, the EQ methodology of damage accrual is applied. The times to damage provided in NUREG/CR-6850 Appendix H were converted to damage rates by taking the reciprocal of the time to damage. For instance, the 19 minute time to damage for an 11 kW/m 2 incident heat flux in Table H-7 is converted to a min-'19 damage rate. This provides a discrete set of damage rates for the heat flux E3 -22 Enclosure 3 Response to Fire Modeling RAIs 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 Farley Fire PRA model used a damage rate profile that assumed no damage before a critical incident heat flux was reached, directly applying the Appendix H data which states that no damage occurs prior to critical heat flux. The updated methodology, referred to in the response to RAI PRA 20, referenced in this RAI, updates the model to assume a damage rate equal to the critical heat flux damage rate for incident heat flux values up to and including the critical heat flux.This approach bounds any degradation of the cable target for heat flux values below the critical heat flux. Beyond the critical heat flux, the Arrhenius curve damage rates are applied with no maximum damage rate applied, making this approach more conservative than that defined by the NUREG/CR-6850, Appendix H data. This ensures the use of a bounding damage rate curve without extrapolating data to lower heat flux values, using the critical heat flux damage rate as a minimum damage rate, and not imposing damage rate limits beyond a maximum heat flux, thereby providing a conservative, bounding analysis.Figure 1 below shows a plot of the damage rate -heat flux profile that models this approach.

The figure shows the plot of the damage rate as specified in NUREG/CR-6850, Table H-7 (square and triangular data points) versus the bounding curve as it is being use in the Farley Fire PRA (square and diamond data points). The Farley Fire PRA approach ensures conservatism in the range of heat flux values where Table H-7 does not provide data points and it conservatively, with respect to NUREG/CR-6850, extrapolates the data for higher heat flux values using the Arrhenius methodology.

The Farley approach uses bounding data points with respect to the data specified in NUREG/CR-6850, Appendix H (specifically for heat flux values below the critical heat flux and for data points where the damage rate exceeds 1.0/minute).

See Farley Hot Gas Layer and Multi-Compartment Analysis and Scenario Report for details of the application of this evaluation to the Fire PRA model.E3 -23 Enclosure 3 Response to Fire Modeling RAIs C 4-m E 0 1.5 1* Damage Accrual Approach* NUREG/CR-6850

• both methods 5 Expon. (both methods)o0 , -- T , 0 5 10 15 20 25 Heat Flux (kW)Figure 1. Damage rate -heat flux profile E3 -24 Joseph M. Farley Nuclear Plant 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 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 01.01 LAR Attachment V, Table V.2-2, provides the results of the electrical cabinet fire severity sensitivity analysis for Unit 1, also indicating similar results for Unit 2.There, the base CDF rose from 5.24E-5/y to 7.05E-5/y, an increase of 1.81 E-5/y.For A CDF, the base value rose from 8.80E-6/y to 1.03E-5/y, an increase of 1.50E-6/y.

The analogous results for LERF and A LERF were as follows: (1) a LERF increase of 2.59E-6/y from 1.26E-6/y to 3.85E-6/y; (2) a A LERF increase of 9.90E-8/y from 4.14E-7/y to 5.13E-7/y.

Subsequently, the LAR was supplemented by a sensitivity analysis which included the effect of removing credit for very early warning fire detection system (VEWFDS) in the main control room (MCR) in addition to the electrical cabinet fire severity adjustment.

The results were as follows: (1) CDF now rose only 1.41 E-5/y (vs. the previous 1.81 E-5/y); (2) A CDF now rose only 1.18E-6/y (vs. the previous 1.50E-6/y);

(3) LERF now rose more by 6.28E-6/y (vs. the previous 2.59E-6/y);

(4) A LERF now rose more by 2.88E-7/y (vs. the previous 9.90E-8/y).

In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019), as justification for the smaller increase for CDF and A CDF with credit for both VEWFDS and electrical cabinet severity adjustment removed, the licensee indicated via Table 1 that, in addition to removing credit for VEWFDS in the MCR, the following additional refinements were now included:

(1) refined main control board (MCB) fire scenarios (via App. L of NUREG/CR-6850);

(2) more realistic probabilities for HGLs; (3) refined circuit analysis for selected fire scenarios; (4)correction to anomalies in fire ignition frequencies for selected fire scenarios.

As a result, the CDF and A CDF increase for removing both VEWFDS and electrical cabinet factor credit were actually less than prior to removal of the VEWFDS credit alone. While the licensee's explanation is sound for these metrics, it remains unclear as to why the LERF and A LERF increases do not display the same trend as CDF and ACDF. If the CDF and A CDF showed a smaller increase with the additional refinements, why did not the LERF and A LERF as well? Explain why the increases in LERF and A LERF after removal of the VEWFDS credit and addition of the four refinements trended upward vs. the downward trend for the CDF and A CDF increases.

RESPONSE: The removal of credit for VEWFDS and electrical cabinet factor credit, along with the four refinements described above is included in the composite effect of the quantification of other RAI responses to be provided in the response to PRA RAI 35.E4 -1 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 06.a.01 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027) the licensee responded to PRA RAI 06(a) and stated that section V.2.2 of the LAR provides the details of the sensitivity analysis related to the bins that have an alpha that is less than or equal to one. Indicate if the acceptance guidelines of Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk- Informed Decisions on Plant-Specific Changes to the Licensing Basis," may be exceeded when this sensitivity study for those bins with an alpha less than or equal to 1 is applied to the integrated study of PRA RAI 35 (see below). If these guidelines may be exceeded, provide a description of fire protection or other measures that can be taken to provide additional defense in depth (DID) (see FAQ 08-0048).RESPONSE: The response to this RAI will be provided by May 23, 2014.E4 -2 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 16.a.01 In a letter dated October 30, 2013 (ADAMS Accession No. ML13305A105), the licensee responded to PRA RAI 16.a and partially addressed some of the criteria for assuming damage within MCR panels to be limited to the initiating panel, namely the presence of no openings and a double wall with an air gap. However, Appendix S of NUREG/CR-6850 also states that there be no sensitive electrical equipment in the adjacent cabinet (or else such equipment to have already been"qualified" above 82C), even with the double wall with air gap. Otherwise damage to such equipment should be postulated.

Explain whether these additional criteria are met or not. If the latter, explain how damage is modeled or, if not, the basis for assuming no damage. (Also see PRA RAI 33.a.01.)RESPONSE: The response to this RAI will be provided by May 23, 2014.E4 -3 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 21.a.01 In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019), the licensee responded to PRA RAI 21 .a and confirmed that the three severity factors, 5.02E-4, 4.84E-4 and 0.00158, do not derive from Figure L-1 in NUREG/CR-6850 but are specifically calculated based on the type of ignition source, scenario location and abandonment time for the MCR abandonment analysis.

The three severity factors correspond to the abandonment probabilities for transient ignition sources, equipment room fixed ignition sources and MCR fixed ignition sources, respectively.

Provide a discussion of the derivation of these factors, including their bases, e.g., as given in Section 13.2.1 of the Farley Scenario Development Report, PRA-BC-1 1-014, and Section 6 of Units 1 and 2 Control Room Abandonment Times at the Joseph M. Farley Nuclear Plant, Rev.0.RESPONSE: The response to this RAI will be provided by May 23, 2014.E4 -4 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 29.01 In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019), the licensee responded to PRA RAI 29 and indicated that 22 supporting requirements (SRs) fail to meet Capability Category (CC) II, 17 more than the staff was able to determine by review of LAR Attachment V, Table V-1. The licensee response refers to dispositions in LAR Attachment V, Table V-i, which, while indicating how the licensee addressed the related findings and observations (F&Os), do not specifically explain why failing to meet CC-Il is acceptable for transition under NFPA 805. Provide Table V-2 which explains the rationale for acceptability of less than CC-Il satisfaction for all 22 SRs.RESPONSE: During the peer review for the Farley fire PRA, peer reviewers identified 22 SRs meeting a Capability Category less than CC-Il. Among them, only the following 3 SRs had not been closed before the submittal of the LAR.* FSS-Ci* FSS-C2* FSS-E3 These 3 SRs are met at CC-I level. As explained in Table V-2 attached to this response, meeting these 3 SRs at CC-I level was determined to be acceptable for this application.

For the remaining 19 out of 22 SRs, the technical issues in the Facts and Observations (F&Os) had been resolved after the peer review and before submittal of the LAR in order to make these SRs met at the CC-Il level.The details of the closure of these 19 SRs are presented in Attachment V, Table V-1 revised in Enclosure 5 of the submittal of supplemental information (NL-12-2566 dated December 20, 2012) which was requested by the NRC for the acceptance review.Note that PRA RAI 29 RAI dated July 8, 2013 (ADAMS Accession No.ML13176A093), included SRs FSS-D7 and FSS-H5 as being categorized less than CC-Il. However, these two F&Os have been closed and are considered to be met at CC-Il as discussed below.There are 3 topics on SR FSS-D7; 1) the credited suppression system to be installed and maintained in accordance with applicable codes and standards, 2)floor non-suppression probability of 0, and 3) applicability of control room lambda to the MCR equipment rooms. The first two topics are discussed in the revised Table V-I. Regarding the applicability of control room lambda to the MCR equipment rooms, Control Room lambda (0.33) was not used for MCR Equipment Rooms. Lambdas of electrical cabinet fire (0.102) and transient fires (0.126) are used for MCR equipment rooms.The F&O topic of SR FSS-H5 is the documentation of the uncertainty related to the fire modeling for the fire scenarios.

Note that there is an F&O issued for SR FSS-E3 which is directly related to FSS-H5, but inconsistent with FSS-H5. The indicated resolution for FSS-E3 states in part that the analysis documentation E4 -5 Enclosure 4 Response to Probabilistic Risk Assessment RAIs should be enhanced to note that methods for developing the statistical representation of the uncertainty intervals and mean values currently do not exist.However, F&O FSS-H5 then asks to undertake evaluations to address uncertainty.

Since the documentation (Table D-1 of the Farley Fire PRA Summary report, PRA-BC-F-1 1-017) has been updated to include discussions related to the uncertainty for fire modeling in response to SR FSS-E3, it is believed that SR FSS-H5 is closed.E4 -6 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 33.a.01 In a letter dated October 30, 2013 (ADAMS Accession No. ML13305A105), the licensee responded to PRA RAI 33.a and referenced PRA RAI 16.a. However, neither of the responses to PRA RAI 16.a or PRA RAI 33 discussed the timing for detection and manual suppression prior to fire spread to adjacent cabinets.Furthermore, the response to PRA RAI 33.a indicates that all MCB panels are physically open to one another. Discuss the basis for assuming rapid enough detection and manual suppression prior to fire spread into the adjacent cabinet.(See also PRA RAI 16.a.01.)RESPONSE: The response to this RAI will be provided by May 23, 2014.E4 -7 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 33.c.01 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to PRA RAI 33.c and indicated an intent to revise its MCR abandonment calculation as follows: "The CCDP for the abandonment scenario is based on failure of all actions in the control room. [A] conservative basis was used for determining the abandonment CCDP based on the calculated CCDP associated with panel damage and failure of the MCR actions. The intent of this criteria is to ensure that the abandonment CCDP is an appropriate bounding value given that, shutting down the plant from outside the control room has an inherently higher risk associated with it." These criteria are presented as (1) using conditional core damage probability (CCDP) = 0.1 if FRANC calculates a CCDP < 0.001, (2) using CCDP = 0.2 if FRANC calculates a CCDP between 0.001 and 0.1, and (3) using 1.0 if FRANC calculates a CCDP > 0.1. These FRANC-calculated CCDPs are based on both MCB panel damage and failure of human actions in the MCR. Clarify how these human actions were quantified, including any detrimental effects (increased failure probabilities) due to fire effects in the MCR. If screening or other bounding values were used, specify their bases, e.g., screening/scoping approach from NUREG-1 921, "Fire Human Reliability Analysis Guidelines" (or equivalent).

RESPONSE: The response to this RAI will be provided by May 23, 2014.E4 -8 Enclosure 4 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 35 Section 2.4.3.3 of the NFPA 805 standard incorporated by reference into 10 CFR 50.48(c) states that the PSA approach, methods, and data shall be acceptable to the AHJ, which is the NRC. Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," identifies NUREG/CR-6850 as documenting a methodology for conducting a Fire PRA (FPRA) and endorses, with exceptions and clarifications, NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Rev. 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA 805. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes a peer review process utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Leveil/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications")

as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established.

In a 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.Section 2.4.4.1 of NFPA 805 states that the change in public health risk arising from transition from the current fire protection program to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the AHJ, which is the NRC. RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides quantitative guidelines on CDF and LERF and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes.As stated on page B-1 of Appendix B of PRA-BC-F-1 1-004, "Fire PRA Logic Model," the new Westinghouse Shutdown Shield (SDS) was installed in fall 2010.The internal events PRA (IEPRA), upon which the FPRA is based, takes credit for the SDS (failure rate of 0.0271/demand), limiting the leakage rate to 2 gpm where the faces of the SDS seal components remain in contact. The assumed leakage rate is increased to 19 gpm if the SDS actuates but the pump shaft continues to rotate if not tripped in a timely manner. Finally, if the SDS does not actuate at all,"existing" (Westinghouse Owners Group (WOG) 2000 or Rhodes Model) seal model leakage rates are applied. Given the July 26, 2013, 10 CFR Part 21 notification by Westinghouse concerning defects with the SDS performance, provide a sensitivity evaluation that removes all credit for the SDS package, including both probability and consequences as appropriate.

Provide revised estimates of CDF, LERF, A CDF and A LERF, including non-fire hazards for CDF and LERF, as a result of removal of this credit. Should this result in any changes to conclusions regarding the transition satisfying RG 1.174 risk/A risk guidelines, address any changes that will be made to accommodate this.E4 -9 Enclosure 4 Response to Probabilistic Risk Assessment RAIs When performing this analysis, include the composite effect from all previous re-evaluations, including any synergistic effects, specifically including the following:

a. From the LAR and the December 20, 2012 LAR Supplement, sensitivities related to the electrical cabinet fire severity method (Section V.2.1) and use of control power transformer (CPT)(Section V.2.3; also response to PRA RAI 08.a).b. From the RAI Responses dated September 16, 2013 (ADAMS Accession No. ML14038A019):

i. PRA RAI 01 .a -Removal of credit for VEWFDS in the MCR (also PRA RAI 01.01)ii. PRA RAI 15.a -Revised seismic CDF based on 2008 USGS data iii. PRA RAI 28.k -Validity of current Ignition Bin 15 fire frequencies

c. From the RAI Responses dated November 12, 2013 (ADAMS Accession No. ML13318A027):

i. PRA RAI 07.e -Use of 0.1 CCDP for MCR Abandonment ii. PRA RAI 17.d-Turbine Building Collapse iii. PRA RAI 33.c -Revised MCR Abandonment analysis (also RAI PRA 33.c.01)RESPONSE: The response to this RAI will be provided by May 23, 2014.E4 -10 Joseph M. Farley Nuclear Plant 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 Attachment M Revision to License Conditions Farley -NFPA 805 LAR Markup -Attachment M -p. 4 of Attachment to Facility Operating License alternative to the Chapter 3 element is functionally equivalent or adequate for the hazard. The licensee may use an engineering evaluation to demonstrate that a change to an NFPA 805, Chapter 3, element is functionally equivalent to the corresponding technical requirement.

A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard.The licensee may use an engineering evaluation to demonstrate that changes to certain NFPA 805, Chapter 3, elements are acceptable because the alternative is "adequate for the hazard." Prior NRC review and approval would not be required for alternatives to four specific sections of NFPA 805, Chapter 3, for which an engineering evaluation demonstrates that the alternative to the Chapter 3 element is adequate for the hazard. A qualified fire protection engineer shall perform the engineering evaluation and conclude that the change has not affected the component, system, procedure, or physical arrangement functionality using a relevant technical requirement or standard.

The four specific sections of NFPA 805, Chapter 3, are: I 'nsert Text: This License Condition does not apply to any demonstration of equivalency under Section 1.7 of* Fire Alarm and Detection Systems (Section 3.8); NFPA 805.* Automatic and Manual Water-Based Fire Suppression Systems (Section 3.9);* Gaseous Fire Suppression Systems (Section 3.10); and, P Passive Fire Protection Features (Section 3.11).(2) Fire Protection Program Changes that Have No More than Minimal Risk Impact Prior NRC review and approval are not required for changes to the licensee's fire protection program that have been demonstrated to have no more than a minimal risk impact. The licensee may use its screening process as approved in NRC safety evaluation report dated to determine that certain fire protection program changes meet the minimal criterion.

The licensee shall ensure that fire protection defense-in-depth and safety margins are maintained when changes are made to the fire protection program.Transition License Conditions (1) Before achieving full compliance with 10 CFR 50.48(c), as specified by (2)below, risk-informed changes to the licensee's fire protection program may not be made without prior NRC review and approval unless the change has been demonstrated to have no more than a minimal risk impact, as described in (2) above.(2) The licensee shall implement the following modifications to its facility to complete transition to full compliance with 10 CFR 50.48(c) by See plant specific list of modifications identified in Attachment S.(3) The licensee shall maintain appropriate compensatory measures in place until completion of the modifications delineated above.

Joseph M. Farley Nuclear Plant 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 Attachment V Revision to Fire PRA Quality Southern Nuclear Operating Company Attachment V -Fire PRA Quality Table V-2 Fire PRA- Category I Summary 1 SR Capability F&O # and Finding/Observation Status Category FSS-C1 CC-I [FSS-C1-011 Two-point fire intensity model that encompass low likelihood, but potentially risk contributing, fire events were not used in all cases. Fire scenarios were done with ignition sources characterized with one fire intensity.

To reach Capability Category II, use a two-point intensity model for all ignition sources.Utility Comment: The development of fire scenarios for the Farley Fire PRA did not identify any instances where further analysis resolution would be gained by the treatment as inferred by the requirements for CC II and CC Ill. The implications of retaining the CC I treatment in lieu of refining as described for CC II or CC III is potentially a higher calculated CDF contribution.

The CC I treatment inherently will not result in under-estimation of fire risk. As such, the current treatment is conservative.

Provided this treatment does not result in masking of risk increases in future applications, further refinements are not considered necessary.

The development of fire scenarios for the Farley Fire PRA did not identify any instances where further analysis resolution would be gained by the treatment as inferred by the requirements for CC II and CC I1l. The implications of retaining the CC I treatment in lieu of refining as described for CC II or CC Ill is potentially a higher calculated CDF contribution.

The CC I treatment inherently will not result in under-estimation of fire risk. As such, the current treatment is conservative.

Provided this treatment does not result in masking of risk increases in future applications, further refinements are not considered necessary.

Response:

The SR stipulates that a two-point model is required for CC-Il. As you stated in your comment, Farley feels that the one-point model is conservative and justified.

This would be viewed as the proposed resolution, but the F&O stands.FSS-C2 C C-1[FSS-C2-01]

Ignition source intensity were characterized such that fire is initiated at full peak intensity and ignition sources that are significant contributors to fire risk were not characterized using a realistic time-dependent fire growth profile. Generic methods from the Hughes Associates Generic Fire Modeling Treatments were used to characterize ignition source intensity.

These generic methods did not incorporate fire growth curves.The only readily available reference for a time dependent growth rate that could be considered in the analysis is 12 minutes as recommended in NUREG/CR-6850.

The treatment would involve a t 2 growth rate. If a particular source/target interaction has a spacing where the target is at the critical damage spacing threshold, such a treatment may provide some benefit as successful suppression with that time period would prevent target damage. However, if the target is located well within the Rev 01 Page V-31 Rev 01 Page V-31 Southern Nuclear Operating Company Attachment V -Fire PRA Quality Table V-2 Fire PRA- Category I Summary 1 SR Capability F&O # and Finding/Observation Status Category Characterize ignition sources that are significant contributors to fire risk using a realistic time-dependent fire growth profile.Utility Comment: The only readily available reference for a time dependent growth rate that could be considered in the analysis is 12 minutes as recommended in NUREG/CR-6850.

The treatment would involve a t 2 growth rate. If a particular source/target interaction has a spacing where the target is at the critical damage spacing threshold, such a treatment may provide some benefit as successful suppression with that time period would prevent target damage. However, if the target is located well within the calculated damage distance, the corresponding time to reaching the damage threshold is very short and effectively precludes any meaningful credit for suppression.

In the case of the Farley Fire PRA, the majority of the target spacing for the dominant risk contributors is such that no meaningful credit for suppression is available.

In other dominant risk contributors, the scenario involves high energy arcing fault (HEAF) events were no growth time is applicable.

The implications of retaining the CC I treatment in lieu of refining as described for CC I1/111 is potentially a slightly higher calculated CDF contribution.

The CC I treatment inherently will not result in under-estimation of fire risk. As such, the current treatment is conservative.

Provided this treatment does not result in masking of risk increases in future applications, further refinements are not considered necessary.

calculated damage distance, the corresponding time to reaching the damage threshold is very short and effectively precludes any meaningful credit for suppression.

In the case of the Farley Fire PRA, the majority of the target spacing for the dominant risk contributors is such that no meaningful credit for suppression is available.

In other dominant risk contributors, the scenario involves high energy arcing fault (HEAF) events were no growth time is applicable.

The implications of retaining the CC I treatment in lieu of refining as described for CC Il/111 is potentially a slightly higher calculated CDF contribution.

The CC I treatment inherently will not result in under-estimation of fire risk. As such, the current treatment is conservative.

Provided this treatment does not result in masking of risk increases in future applications, further refinements are not considered necessary.

Response:

The Farley modeling was found to be consistent with CC-I but did not meet the requirements of CC-Il. The comment provides the basis for stating that the existing treatment is adequate.

It does not provide evidence that a time-dependent heat release rate model was used.FSS-E3 CC-I [FSS-E3-01]

Supporting requirement E3 asks to provide a mean value of, and The documentation has been updated to include discussions Rev 01 Page V-32 Southern Nuclear Operating Company Attachment V -Fire PRA Quality Table V-2 Fire PRA- Category I Summary 1 SR Capability F&O # and Finding/Observation Status Category statistical representation of, the uncertainty intervals for the related to the uncertainty for fire modeling.

See Table D-1 of the parameters used for fire modeling the fire scenarios.

Farley Farley Fire PRA Summary report, PRA-BC-F-1 1-017.performed fire size and heat release rate selection in accordance with NUREG/CR-6850 and/or applicable FAQs. The associated SR was dispositioned as CC I which is judged to However, the methods for developing the statistical be sufficient given the two concerns noted.representation of the uncertainty intervals and mean values currently do not exist. However, this is not reported in the documentation.

In the documentation, explain that it is understood that methods for developing the statistical representation of the uncertainty intervals and mean values currently do not exist.Utility Comment: This specific F&O was issued against a technical element and the indicated resolution involves a documentation clarification.

This documentation clarification will be implemented.

1 _ All Fire PRAs SRs characterized as Capability Category I were identified as Findings in the Peer Review.Rev 01 Page V-33 Rev 01 Page V-33