ML14114A458

90-Day 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

ML14114A458
Person / Time
Site:	Saint Lucie
Issue date:	03/25/2014
From:	Jensen J Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML14114A457	List: ML14114A458
References
L-2014-083
Download: ML14114A458 (23)

Text

Enclosure 2 contains Security Related information - Withhold under 10 CFR 2.3 90.

Upon removal of Enclosure 2, this document is decontrolled. - .. .

March 25, 2014 FPL, L-2014-083 10 CFR 50.90 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555-0001 Re: St. Lucie Units I and 2 Docket Nos. 50-335 and 50-389 90-Day Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

References:

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

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

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

4. St. Lucie Plant Units 1 and 2 Request for Additional Information on License Amendment Request to Adopt National Fire Protection Association Standard 805 Performance-Based Standard for Fire Protection (TAC Nos. MF 1373 and MF 1374) dated December 26, 2013.

Per Reference I above, Florida Power and Light Company (FPL) requested an amendment to the Renewed Facility Operating License (RFOL) for St. Lucie Units 1 and 2. The License Amendment'Request (LAR) will enable FPL to adopt a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a) and (c) and the guidance in Revision I of Regulatory Guide (RG) 1.205.

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

By letter dated December 26, 2013 (Reference 4) NRC Staff requested additional information regarding the LAR. Based on discussions with the NRC Staff, the additional information Florida Power & Ught Company 6501 S. Ocean Drive, Jensen Beach, FL 34957

L-2014-083 10 CFR 50.90 requested was prioritized and the response to the request for additional information will be provided in three separate submittals. The attachments to this letter provide the 90-day response to the request for additional information.

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

FPL requests that Enclosure 2 to this letter, which contains sensitive security-related information, be withheld from public disclosure in accordance with i 0 CFR 2.390.

This letter makes no new comipmitments or changes to existing commitments.

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

I declare under penalty of perjuy-1 that the foregoing is true and correct.

Executed on March25l, 2014.

Respectfully St. Lucie Plant JJ/rcs

Enclosures:

1. FPL's St Lucie Units I and 2 NFPA 805 LAR 90-Day RAI

Response

2. FPL's St Lucie Units 1 and 2 NFPA 805 LAR 90-Day RAI Response - Withheld from Public Disclosure cc: Ms. Cindy Becker, Florida Department of Health

Enclosure I to L-2014-083 Page 1 of 21 Enclosure 1 FPL's St Lucie Units 1 and 2 NFPA 805 LAR 90-Day RAI Response PSL FM RAI 03a PSL PRA RAI 02a PSL RR RAI 01 PSL FM RAI 03b PSL PRA RAI 03a PSL FM RAI 03c PSL PRA RAI 03b PSL PRA RAI 03c PSL PRA RAI 05 PSL PRA RAI 06 PSL PRA RAI 11 a PSL PRA RAI lI b PSL PRA RAI II c PSL PRA RAI II d PSL PRA RAI II e PSL PRA RAII Ihf PSL PRA RAI 16 PSL PRA RAI 17a PSL PRA RAI 17b

Enclosure I to L-2014-083 Page 2 of 21 PSL FM RAI 03a NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the fire PRA development (NFPA 805 Section 4.2.4.2). Reference is made to LAR Attachment J, "Fire Modeling V&V," for a discussion of the verification and validation (V&V) of the fire models that were used.

Furthermore LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states "Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805."

Regarding the V&V of fire models:

Provide the V&V basis for the methods and correlations used to calculate flame height, HRR, flame spread, incident heat flux to targets, etc. in the fire hazard assessment of exposure to safe shutdown raceways in Unit 1 developed in support of the licensee's request for 10 CFR 50 Appendix R KI Exemption.

RESPONSE

The 10 CFR 50 Appendix R KI Exemption is based on Report 6372, "Fire Hazards Assessment of Exposure to Safe Shutdown Raceways, St. Lucie Unit 1," which uses fire modeling calculations to determine the heat flux and thermal plume exposures to raceway targets in the containment building. The fire modeling methods that are used include a flame height calculation, a heat release rate calculation, a flame spread rate calculation, and flame spread distance calculation, a fire duration calculation, and a cable tray heat flux model. The Verification and Validation (V&V) basis for the fire modeling calculations is described in Report 0027-0009-014-002, Revision 1.

LAR Attachment J will be updated to include Report 0027-0009-014-002, Revision 1 as the V&V basis for Report 6372. A markup of LAR Attachment J will be provided in conjunction with the response to the 120 day RAIs.

PSL FM RAI 03b NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the fire PRA development (NFPA 805 Section 4.2.4.2). Reference is made to LAR Attaclunent J, "Fire Modeling V&V," for a discussion of the verification and validation (V&V) of the fire models that were used.

Furthermore LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states "Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805."

Regarding the V&V of fire models:

Enclosure I to L-2014-083 Page 3 of 21

b. Provide the V&V basis for the methodology that was used to convert damage times for thermoplastic cables in Appendix H of NUREG/CR-6850 to percent damage as a function of heat flux.

RESPONSE

The methodology basis is provided in the response to FM RAI 02.e, which was submitted with the February 24, 2014 RAI responses, and is an approach used for environmental qualification (EQ) which utilizes an Arrhenius analysis. The approach taken to convert damage times to a percent damage applies the Arrhenius methodology to the test data provided in NUREG/CR-6850 Appendix H. Calculations that utilize this methodology have been verified as part of the associated document review process.

PSL FM RAI 03c NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."

LAR Section 4.5.1.2, "Fire PRA" states that fire modeling was performed as part of the fire PRA development (NFPA 805 Section 4.2.4.2). Reference is made to LAR Attaclunent J, "Fire Modeling V&V," for a discussion of the verification and validation (V&V) of the fire models that were used.

Furthermore LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," states "Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805."

Regarding the V&V of fire models:

b. Provide the V&V basis for the model that was used to calculate the HGL temperature in the NSP calculations for the containment. Provide technical details to demonstrate that the model has been applied within the validated range of input parameters, or to justify the application of the model outside the validated range reported in the V&V basis documents.

RESPONSE

A V&V evaluation has been performed for the containment hot gas layer (HGL) manual non-suppression probability (NSP) calculation, this analysis confirms that the evaluation is within the validation ranges of the fire models used. The analysis was performed using the Method of Byler (NUREG- 1824, Volume 3, Table 3-1, Predicting Hot Gas Layer Temperature in a Fire Room With Door Closed, Excel File Name 02.3 Temperature_CC.xls) to determine the heat release rate required to produce a hot gas layer temperature of 205 C (thermoplastic damage temperature) at a time of 90 minutes (corresponding to an NSP of 0.001 per NUREG/CR-6850, Supplement 1, Chapter 14 for PWR Containment Fires) in the St. Lucie (PSL) containment. The calculation for hot gas layer temperature used a 2 meter concrete boundary whereas the St. Lucie containment is a free standing steel shell. Based on this the calculated hot gas temperature is over predicted because the steel shell would absorb a significant amount of heat prior to the transfer of heat to the concrete, resulting in a lower hot gas layer temperature at the 90 minute time frame.

Enclosure 1 to L-2014-083 Page 4 of 21 The NUREG/CR-6850 formula for cable tray fires, Section R.3, p. R-4, provides the relationship between the area of a burning tray and its associated heat release rate. Using the heat release rate associated with the 0.001 manual non-suppression probability specified above, the length of tray required to cause a hot gas layer to reach the thermoplastic cable damage temperature is over 150 feet of a 2 foot wide tray. This analysis is extremely conservative in that it does not consider the full containment volume and the length of tray specified would be required to be involved from time zero till the 90 minute timeframe specified.

Of the parameters specified in NUREG-1934 the parameter of primary importance for the containment hot gas layer analysis is the compartment aspect ratio. For the containment, the aspect ratio is calculated as being beyond the validated range indicated in NUREG-1934, Table 2-

5. In order to ensure that the analysis was within the validation range, the effective containment dimensions have been truncated, such that the compartment aspect ratio is bounded by the maximum validated value. The other parameters specified in Table 2-5 are not applicable to this analysis as discussed below:

Fire Froude Number - the above analysis does not attempt to evaluate the specific characteristics of the fire plume. The analysis focuses on defining a time to hot gas layer temperature which is primarily a function of the heat release rate and the compartment height and volume.

Flame Length Ratio - The fire configuration is not impacted by ceiling effects given the height of the compartment.

Ceiling Jet Distance Ratio - This parameter is associated with evaluation of effects at a given horizontal distance from the plume. Because this evaluation, given the size of containment, is a primarily volumetric effect, this ratio does not impact the analysis.

Equivalence Ratio - The effect of ventilation of the fire is minimal given a closed but very large containment environment. The fire is expected to have sufficient ventilation in its early stages and may become under ventilated at the latter timeframes. The analysis assumes sufficient oxygen to sustain the heat release rate for the duration of the fire and therefore will provide conservative results with respect to the time to hot gas layer.

Radial Distance Ratio - This factor is important when calculating radiative heat flux to a target.

Given the volumetric nature of this analysis and the constant heat release rate input, this factor will not impact the analysis results.

This analysis confirms the bounding value for manual non-suppression probability used for the containment fire scenario. The analysis uses bounding conservative analysis parameters and provides bounding results. The use of a 90 minute containment non-suppression probability for suppression of a fire at a specific location inside containment impacting components at a remote location in containment due to a hot gas layer is considered bounding and conservative based on this analysis and associated fire size required to create the hot gas layer impact (over 150 foot tray section burning for 90 minutes).

The containment HGL NSP V&V analysis will be incorporated in the updated Hot Gas Layer and Multi Compartment analysis and a markup of LAR Attachment J will be provided in conjunction with the response to the 120 day RAIs.

Enclosure I to L-2014-083 Page 5 of 21 PSL PRA RAI 02a LAR Attachment S, Table S-1 identifies several modifications (i.e., 1B-MR-02, 2B-MR-01 and 2B-MR-02) to provide in-panel detection of fires. Address the following:

a. The risk-informed characterization column of Modification 2B-MR-01 in LAR Attachment S, Table S-I indicates that incipient detection is used to support detection of a fire well in advance of significant damage. Clarify whether this credit is consistent with FAQ 08-0046. Additionally, provide confirmation that incipient detection is not credited to limit the extent of internal fire damage within monitored panels.

RESPONSE

In Fire Zone 2-52 (Fire Area 2B) one scenario, F12-PTB for an annunciator logic cabinet, credits the preclusion of damage in both the severe and non-severe scenario. This credit will be removed from the quantification on the non-severe scenario to ensure that damage within the cabinet is not precluded. The current risk of scenario F 12-NS-PTB is 1.14E-09/year and 4.45E-1 I/year for CDF and LERF respectively. Removing the 0.02 credit for incipient detection precluding the non-severe fires will lead to an increase in plant risk of 5.7E-08/year for CDF and 2.22E-09/year for LERF. This is a small change in terms of total plant risk and will not impact the delta CDF and delta LERF as their relative contribution to MCR abandonment is exceedingly small.

The credit for incipient detection is consistent with FAQ 08-0046 and based on the above analysis will no longer be credited to limit the extent of internal fire damage within the monitored panel.

The Fire PRA will be updated in conjunction with the 120 day RAI responses to reflect this change.

PSL PRA RAI 03a Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805.

In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02.

Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

Section 7.1.1 of Fire Scenario report appears to indicate that fires within some Bin 15 cabinets above 440V, e.g., motor control centers (MCCs), are not assumed to propagate outside of the cabinet. Additionally, a review of walkdown sheets documented Attachment D of the Fire Ignition Frequency Development report does not appear to sufficiently document whether cabinets are well sealed and robustly secured, including for those credited as such in the HGL and MCA report. As a result, relative to the counting and treatment of Bin 15 electrical cabinets, address the following:

a. Per Section 6.5.6 of NUREG/CR-6850, fires originating from within "well-sealed electrical cabinets that have robustly secured doors (and/or access panels) and that house only circuits below 440V" do not meet the definition of potentially challenging fires and

Enclosure I to L-2014-083 Page 6 of 21 therefore should be excluded from the counting process for Bin 15. By counting these cabinets as ignition sources within Bin 15 the frequencies applied to other cabinets are inappropriately diluted. Clarify that this guidance is being applied. If not, provide justification for the inclusion of such cabinets in the Bin 15 count, and evaluate the impact on risk results of excluding them.

RESPONSE

Non-Severe scenarios, including scenarios which are limited to fire damage to cables within a robustly sealed panel, are currently counted in the total plant Bin 15 count, regardless of the voltage level. A review of both units identified a count of 22 Electrical Panels below 440V (or multiple panel segments, for large panels) which are associated with robustly sealed panels. The removal of these scenarios will lead to a decrease in the denominator, total bin count, for bin 15 of about 2.5% (total count of 886 is reduced to 864). The overall impact on CDF will be a fraction of the per panel bin 15 frequency increase given that the total risk is comprised of contributions by bins other than bin 15. While this will likely have a very small impact for most scenarios, the frequencies for these scenarios will be removed from the bin 15 ignition frequency count as part of the quantification revision to be performed in conjunction with the response to the 120 day RAIs.

The overall impact in terms of plant risk is expected to be small.

PSL PRA RAI 03b Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805.

In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02.

Methods that have not been detennined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

Section 7.1.1 of Fire Scenario report appears to indicate that fires within some Bin 15 cabinets above 440V, e.g., motor control centers (MCCs), are not assumed to propagate outside of the cabinet. Additionally, a review of walkdown sheets documented Attachment D of the Fire Ignition Frequency Development report does not appear to sufficiently document whether cabinets are well sealed and robustly secured, including for those credited as such in the HGL and MCA report. As a result, relative to the counting and treatment of Bin 15 electrical cabinets, address the following:

b. In addition, all cabinets having circuits of 440V or greater should be counted for purposes of Bin 15 frequency apportionment based on the guidance in Section 6.5.6 of NUREG/CR-6850. Clarify that this guidance is being applied. If not, provide justification for the exclusion of such cabinets in the Bin 15 count, and evaluate the impact on risk results of including them.

RESPONSE

Sealed cabinets were not excluded from the total bin 15 counts. As such, all electrical panels over 440V sealed, or not, have been counted towards the total bin 15 count.

Enclosure 1 to L-2014-083 Page 7 of 21 PSL PRA RAI 03c Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805.

In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02.

Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

Section 7.1.1 of Fire Scenario report appears to indicate that fires within some Bin 15 cabinets above 440V, e.g., motor control centers (MCCs), are not assumed to propagate outside of the cabinet. Additionally, a review of walkdown sheets documented Attachment D of the Fire Ignition Frequency Development report does not appear to sufficiently document whether cabinets are well sealed and robustly secured, including for those credited as such in the HGL and MCA report. As a result, relative to the counting and treatment of Bin 15 electrical cabinets, address the following:

c. Per NUREG/CR-6850, cabinets below 440V that are not well sealed and robustly secured as well all cabinets above 440V should be assumed to propagate outside the cabinet. Clarify that this guidance is being applied. If not, provide justification for not considering propagation outside of such cabinets, and evaluate the impact on risk results, considering propagation to secondary targets.

RESPONSE

All unsealed panels are assumed to propagate outside of the panel with the potential to damage all targets within the panel's zone of influence. Voltage level was not considered as a reason to exclude a bin 15 fire for panels that are not sealed. Secondly, all panels that can experience a HEAF, i.e., medium voltage switchgear, are considered to damage targets outside of the panel whether the panel is sealed or not. The HEAF zone of influence is applied in accordance with NUREG/CR-6850 Appendix M. However, 480V MCCs that are sealed, and are not subject to a HEAF because they are not considered to be switchgear per the definition of NUREG/CR-6850 Appendix M, are assumed to be impacted by a non-severe fire only and the fire is not assumed to propagate outside of the panel. This is in accordance with NUREG/CR-6850, Supplement 1, Section 8 (based on FAQ 08-0042), which notes that HEAF related fires must be considered regardless of whether the panel is sealed. For non-HEAF related fires, the only criteria for panel fires is that the panel is unvented and well-sealed, i.e., multiple points of contact to ensure warping of the panel will not allow air flow into the panel.

"In the case of electricalpanels, the panel ventilation configurationand the latching configuration of the doors are important.If the panel contains open vents, either at the top or bottom of the pane (sic), or ifpenetrations into the top or sides of the panel are notfire-sealed,fires can be assumed to be capable of spreading out of the panel to secondarycombustibles. However,for un-vented cabinets,fire spreadmay be less likely. Firespread out of the panel may still occur, unless the panel doors are attached and anchoredat multiple points.

Simple twist-handle style top-and-bottom door latches are not sufficient to contain afire within a panel. Substantialwarping of the doorface may occur due to the heat of the fire. This can allow

Enclosure 1 to L-2014-083 Page 8 of 21 gaps to open in an otherwise un-ventedpanel. In contrast,fire spread is not consideredlikely given a weather-tight or waterproofcabinet construction where multiple mechanicalfasteners secure panel access plates and where all penetrationsinto the panel are sealed."

As indicated in the FAQ, the reason for no propagation for bin 15 fires for well-sealed cabinets is that it is assumed all air will bum out of the cabinets during the fires incipient phase. For a well-sealed panel the fire will starve itself and there will be no propagation regardless of voltage level.

PSL PRA RAI 05 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing frequently asked question (FAQ) process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

Section 3.3 of the Fire Ignition Frequency Development report utilizes an approach to apportion transient and cable fire frequencies associated with hot work by using a separate hot work influence factor, which deviates from NUREG/CR-6850; however, a review of LAR Attachment H and the Fire Ignition Frequency Development report indicates that FAQ 12-0064, which defines such an influence factor, is not referenced. Additionally, a review of Table 3-4 indicates that no very high influencing factors (50) were applied and that only low (1) and medium (3) values were applied to the hot work influencing factors. Furthermore, Section 8.4 of the Fire Scenario report notes that an additional 0.1 adjustment factor has been applied to reduce transient frequencies in Unit 2 fire zone 51X. In light of these observations, clarify whether influencing factors were developed per the guidance in NUREG/CR 6850 and FAQ 12-0064, and justify why the full range of influencing factors was not implemented. If any frequency-adjustment factors deviating from the guidance presented in NUREG/CR-6850 and FAQ 12-0064 (e.g., the use of 0.1 in Unit 2 fire zone 51 X) are used, justify their use, and evaluate their impact on risk results.

RESPONSE

Transient weighting factors have been redistributed in accordance with guidance in FAQ 12-0064.

The full range of probabilities have been utilized including ratings of very high (50) and extremely low (0.1). Extremely low ratings have been used to replace the 0.1 adjustment factor previously applied to Unit 2 Fire Zone 51X. The extremely low rating is in accordance with the new zero transient zone administrative controls being applied in specified zones as needed (Refer to the responses to PRA RAI 04, included in the February 24, 2014 RAI responses, for additional detail on these controls). The overall impact of the update to the FAQ 12-0064 methodology is expected to be small with some zones seeing an increase while others see a decrease.

Enclosure 1 to L-2014-083 Page 9 of 21 PSL PRA RAI 06 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA-805. In letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02.

Methods that have not been determined to be acceptable by the NRC Staff require additional justification to allow the NRC Staff to complete its review of the proposed method.

Section 3.3 of the Fire Ignition Frequency Development report states that the potential for junction box fires is included in the fire ignition frequencies for Unit 1; however, this section appears to indicate that junction box fires were not considered for Unit 2. Describe the treatment of junction boxes in the FPRA as ignition sources, and clarify whether this treatment is consistent with guidance in NUREG/CR-6850. If not, evaluate the risk impact of considering junction box fires for Unit 2. Note that draft FAQ 13-0006 (ADAMS Accession No. ML13212A378) offers a complete technical path related to junction boxes.

RESPONSE

The Fire PRA supporting the LAR submittal included junction box ignition frequency in self-ignited cable scenarios for Unit 1 only due to the use of non-IEEE-383/thermoplastic cables. For Unit 2 where IEEE-383/thermoset cables are used no junction box related fire scenarios were applied. To address the guidance provided in FAQ 13-0006, a sensitivity analysis has been performed where scenarios have been created for both units in which the risk attributed to junction boxes is calculated. If no PRA equipment cables are routed through the junction box, then no fire was postulated. The fire risk impact of these junction box scenarios resulted in a total CDF contribution of less than 0.1% of the risk reported in the LAR for either unit. As such, no further action to incorporate the junction boxes into the analysis is planned. Quantification associated with 120 day RAIs will include a sensitivity evaluation for CDF, delta CDF, LERF and delta LERF.

PSL PRA RAI lla LAR Attachment W provides the ACDF and ALERF for the variance from deterministic requirements (VFDRs) and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe, in detail, how ACDF and ALERF or the additional risk of recovery actions were calculated. Describe the method(s) used to determine the changes in risk reported in Tables W-6 and W-7 of Attachment W. The description should include:

a. A detailed definition of both the post-transition and compliant plants used to calculate the reported changes in risk and additional risk of recovery actions.

RESPONSE

The post transition baseline or PTB plant represents the plant upon full transition to NFPA 805, after all plant modifications have been completed and the NRC has issued the SE and has granted

Enclosure I to L-2014-083 Page 10 of 21 PSL self-approval. In terms of calculating ACDF and ALERF, the PTB case credits all modifications and recovery actions required to meet risk criteria under RG 1.174.

The compliant case represents a hypothetical plant condition as it would exist if PSL had been built in complete compliance with deterministic requirements, without recovery actions. In order to quantify the risk of such a plant configuration, VFDRs (variances from deterministic requirements) were identified and the associated model basic events were set to their respective components' random failure probability (elimination of the VFDR). Any actions taken not associated with a VFDR are set equal to the value for the PTB case. As the compliant case represents the plant as it is built today, but without any variances, modifications need not be included in the compliant case because modifications that are related to a VFDR, the variances are directly eliminated by setting the associated components' failure probability to the PRA model's nominal random failure probability for the component and for a modification that is not related to a VFDR, the modification is not required to make the configuration "compliant".

The total delta risk obtained as described above is conservatively assumed to be associated with the additional risk of recovery actions. The actual additional risk of recovery actions would be a fraction of this total delta risk.

PSL PRA RAI 11 b LAR Attachment W provides the ACDF and ALERF for the variance from deterministic requirements (VFDRs) and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe, in detail, how ACDF and ALERF or the additional risk of recovery actions were calculated. Describe the method(s) used to determine the changes in risk reported in Tables W-6 and W-7 of Attachment W. The description should include:

b. A description of how the reported changes in risk and the additional risk of recovery actions were calculated. Include in this description a discussion of PRA modeling mechanisms used to determine the reported changes in risk (e.g. altering the probabilities of basic events and modeled recovery actions). Clarify whether FAQ 08-0054 guidance was used.

RESPONSE

The change in risk was determined by calculating the difference between the post-transition baseline plant (PTB) and the compliant plant. The additional risk of recovery actions was conservatively assumed to be the entirety of the delta between the PTB and compliant plant configurations (see RAI PRA 11.g response for a discussion of the methodology used for calculating the additional risk of recovery actions for Fire Areas 1A and 1C). The methodology for the FREs is based on the requirements of NFPA 805 and guidance in Section 5.3 of NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2, as modified by RG 1.205 Revision 1. The methodology also relies upon clarifications provided in FAQs 08-0054, Revision 1, and 07-0030, Revision 5.

See RAI PRA 11 a for additional discussion of the quantification of the PTB and compliant scenarios.

Enclosure 1 to L-2014-083 Page 11 of 21 PSL PRA RAI l1c LAR Attachment W provides the ACDF and ALERF for the variance from deterministic requirements (VFDRs) and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe, in detail, how ACDF and ALERF or the additional risk of recovery actions were calculated. Describe the method(s) used to determine the changes in risk reported in Tables W-6 and W-7 of Attachment W. The description should include:

c. A discussion of any exceptions to normal modeling mechanisms discussed in (b),

including those cases for which the PRA model lacks sufficient resolution to model the VFDR or those that utilize surrogate basic events or HFEs to estimate/bound the change in risk in lieu of manipulating components or actions directly associated with the VFDR.

RESPONSE

The exceptions are associated with the delta risk calculation for the Main Control Room (MCR) and the Cable Spreading Room (CSR), Fire Areas 1B, 2B, IF, and 2F. For the MCR, because the deterministic analysis only addresses MCR abandonment, the delta risk is quantified only for the MCR abandonment scenarios, the fraction of the ignition frequency for each ignition source which would result in exceeding MCR habitability criteria as specified in NUREG/CR-6850 (Section 11.5.2.11). For abandonment scenarios, the post transition baseline (PTB, or variant) case CCDPs and CLERPs have been increased from the value calculated based on specific scenario damage values to account for the potential increase in failure probability due to shutdown from outside the MCR. Calculated CCDPs less than 1E-03 are factored up to 0.1, CCDPs between 1E-03 and 1E-01 are increased to 0.2 and CCDPs greater than 1E-01 are set to 1.0. The LERF calculation is done by multiplying the factored CCDP by the ratio of the calculated CLERP to the calculated CCDP. For the MCR compliant case, a CCDP is used based on the sum of the EDG Failure to Start, EDG Failure to Run and EDG Unavailable due to Maintenance terms in the full power internal events model (the sum of these failure probabilities is 4.4E-02, the value used in the analysis was a conservatively low value of 4E-02 as a surrogate lower bound CCDP for the compliant case). Credit for the EDG is consistent with the current post fire shutdown analysis which relies on the EDGs for shutdown.

The compliant case represents the lower bound value based on the following. The compliant case represents the deterministically compliant plant where the MCR would still be abandoned due to habitability. The alternate shutdown panel would have all the controls and isolation switches to allow a successful shutdown from the alternate shutdown panel without any recovery actions. This requires only one train free from fire damage being controlled at the PCS. Using the random failures of the EDG is appropriate because that represents the least reliable component in the single train shutdown. Not using any human failure probabilities results in the compliant risk being significantly underestimated because even in the deterministic plant the control is abandoned and control is transferred to the alternate shutdown panel. In the case where the calculated CCDP is very small, indicating additional equipment is available, the use of the calculated CCDP for the compliant case and a CCDP of 0.1 for the variant case provides a conservative delta risk for a relatively low risk scenario. It would do little to improve the CCDP further for the compliant case because the additional equipment used (it would have to be the second train of emergency power) would be subject to the same limitations as in the PTB case of local actions that may or may not be proceduralized. In this case the delta CDF is likely very close to zero. Because the CCDP is set to

Enclosure 1 to L-2014-083 Page 12 of 21 0.1 for the case with a low calculated CCDP, the delta CDF for these cases is overestimated. The above methodology may significantly overestimate the delta CDF and therefore, is conservative.

For the CSR, the same compliant case specified above was used. The variant case for the CSR was associated with a scenario that would require MCR abandonment, This was the CSR hot gas layer scenario for which a conservative CCDP of 1.0 was applied for the variant case.

In some instances, VFDRs were not modeled because the specific component was not credited in the Fire PRA. An example of this is 1-HVE-1 1 (Electrical Equipment Room Exhaust Fan). While the model does not specifically credit 1-HVE- 11, it does credit redundant fans 1-HVS-5A and 5B (Electrical Equipment Room Supply Fans). The PRA success criteria is met by the availability of one of the supply fans without the need for an exhaust fan operating. The SSD analysis uses a more conservative criteria which requires operation of the exhaust fan along with one of the supply fans. A VFDR associated with 1-HVE- 11 does not result in loss of the required function based on the PRA success criteria and therefore has no impact on the PRA quantification. In other instances surrogate basic events were used in lieu of manipulating components or actions directly associated with the VFDR. This was done where a VFDR for a particular component was associated with failure of the power supply for that component. The surrogate event allowed elimination of the power supply dependency for that component rather than making the power supply and all of its supported components, which were not the subject of the VFDR, available in the compliant case.

PSL PRA RAI l1d LAR Attachment W provides the ACDF and ALERF for the variance from deterministic requirements (VFDRs) and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe, in detail, how ACDF and ALERF or the additional risk of recovery actions were calculated. Describe the method(s) used to determine the changes in risk reported in Tables W-6 and W-7 of Attachment W. The description should include:

d. A description of any modeling manipulations that use data or methods not included in the FPRA peer review.

RESPONSE

The methods used to calculate the ACDF and ALERF are described in the responses to RAIs PRA 11 a, b and c. These methods involved elimination of a VFDR by exclusion of a fire impact by reverting the basic event failure probabilities to the random failure probabilities of the associated basic events. This approach is consistent with the quantification methods and exclusions used in the base Fire PRA that was evaluated in the peer review.

Enclosure I to L-2014-083 Page 13 of 21 PSL PRA RAI lie LAR Attachment W provides the ACDF and ALERF for the variance from deterministic requirements (VFDRs) and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe, in detail, how ACDF and ALERF or the additional risk of recovery actions were calculated. Describe the method(s) used to determine the changes in risk reported in Tables W-6 and W-7 of Attachment W. The description should include:

e. Bases for not modeling VFDRs in the FPRA (e.g., as indicated in LAR Attachment C).

RESPONSE

A VFDR that is not modeled will fall into one of three categories.

The first category is for Fire Areas 1B, 1F, 2B and 2F, which are areas where MCR abandonment is modeled (Control Rooms and Cable Spreading Rooms). See the response to RAI PRA 11 .c for a detailed discussion of this treatment.

The second category that VFDRs are not considered in the Fire PRA is applicable to the Fire PRA model success criteria and is applicable for any fire area. These VFDRs deal with components or specific failure modes that are not modeled in the Fire PRA as they do not correspond to core damage scenarios. An example of this would be the failure to energize the pressurizer heaters which results in VFDRs for the deterministic analysis where the functional requirements include the need to maintain pressurizer level within the pressurizer level instrumentation indicating range.

The Equipment Selection task, as documented in the Component/Cable Selection Report, evaluates the Fire PRA model against the SSD analysis and addresses differences between these analyses.

The PRA does not consider this a core damage sequence and therefore this VFDR would not be included in the delta risk as measured by the Fire PRA model.

A third category for not modeling a VFDR is related to a conservative limitation of redundant components modeled in the PRA. In some instances VFDRs were not modeled because the specific component was not credited in the Fire PRA. An example of this is I -HVE- 11 (Electrical Equipment Room Exhaust Fan). While the model does not specifically credit 1-HVE-11, it does credit redundant fans l-HVS-5A and 5B (Electrical Equipment Room Supply Fans). The PRA success criteria is met by the availability of one of the supply fans without the need for an exhaust fan operating. The SSD analysis uses a more conservative criteria which requires operation of the exhaust fan along with one of the supply fans. A VFDR associated with 1-HVE-1 1 does not result in loss of the required function based on the PRA success criteria and therefore has no impact on the PRA quantification. Another example of a VFDR not modelled in the FPRA is related to S/G overfill scenarios. A S/G overfill is modeled as a failure mechanism for impacting availability of the steam driven AFW pump but not as an overcooling transient because the overcooling is not associated with a core damage sequence in the PRA. The basis for not modeling a particular VFDR is provided in the Fire Risk Evaluation (PSL-FPER- 11-001, Rev. 1, Section 2.2 for each fire area).

PSL PRA RAI llf LAR Attachment W provides the ACDF and ALERF for the variance from deterministic requirements (VFDRs) and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe, in detail, how ACDF and ALERF or the additional risk of recovery actions

Enclosure I to L-2014-083 Page 14 of 21 were calculated. Describe the method(s) used to determine the changes in risk reported in Tables W-6 and W-7 of Attachment W. The description should include:

f. A separate description specific to how the ACDF and ALERF and additional risk of recovery actions were calculated for the MCR and any other fire area that results in MCR abandonment due to loss of habitability or credits MCR abandonment on loss of control or function (e.g., cable spreading room).

RESPONSE

See response to RAI PRA IIc.

PSL PRA RAI 16 As documented in LAR Attachment U, the peer review of the IEPRA was performed using the criteria in NEI 00-02, and as a result, a self-assessment is required according to the guidance within RG 1.200 (Revision 2) to demonstrate that the technical adequacy of the PRA is of sufficient quality to support the application. Additional self-assessment subsequent to the July 2002 NEI 00-02 peer review have been performed; however, they do not appear to be sufficient demonstrate the technical adequacy of the PRA in accordance with the guidance in RG 1.200 (Revision. 2). In particular, Focused-scope peer reviews have only been performed on select technical elements of the ASME/ANS PRA standard (i.e., large early release (LERF) evaluation (LE), human reliability analysis (HRA), internal flooding (IF), and data analysis (DA) as well as supporting requirement (SR) related to common cause failure). Also, the July 2009 LERF focused-scope peer review was not performed against the latest version of the standard as endorsed by RG 1.200, Revision 2.

" A December 2005 independent review of the PRA was performed using a dated version of the standard (i.e., ASME RA-Sa-2003 "to guide future PRA enhancement activities"). This "basic assessment" does not appear to qualify as a self-assessment. No formal process or review criteria were cited as being used to judge the adequacy of PRA against SRs. Additionally, the exact scope of SRs and PRA documentation reviewed is unclear and is acknowledged by the review team as being limited.

" An October 2007 assessment of the St. Lucie PRA was performed to addresses deficiencies raised by the December 2005 independent review and F&Os from the July 2002 NEI 00-02 peer review; however, this assessment does not appear to follow the self-assessment process outlined in Appendix B of RG 1.200. Additionally, the results are not documented in such a manner that it is clear why each requirement is considered to have been met.

Although self-identified action items to comply with RG 1.200, Rev. 1 are briefly summarized, they are not tied to specific SRs.

Explain how the 2002 peer review and these subsequent focused-scope peer reviews and assessments are consistent with or equivalent to the peer review and self-assessment process in NEI 00-02, Revision 1, as endorsed by RG 1.200, Rev. 2, with clarifications and qualifications, for demonstrating the technical adequacy of the IEPRA for the NFPA 805 application.

Enclosure I to L-2014-083 Page 15 of 21 In addition, provide the deficiencies (or "gap") identified by the latest gap assessment in a manner analogous to the format in which F&Os are presented and dispositioned in LAR Attachment U, including an assessment of each identified gap's impact on the FPRA.

RESPONSE

As correctly identified, only common cause failure (CCF), and Supporting Requirements (SRs) of HRA, IF, and DA High Level Requirements (HLRs) were subject to focused peer reviews per ASME/ANS STD RA-Sa-2009 as endorsed by RG-1.200 Rev 2. LERF was subject to focused peer review per ASME RA-Sb-2005 as endorsed by RG-1.200 Rev 1, and the remaining HLRs were peer reviewed using the earlier NEI 00-02 process. The following table summarizes which peer review was conducted against which ASME/ANS High Level Requirement as endorsed by which industry/regulatory guide.

HLR Peer Review ASME STD Endorsing Guide IE 2002 Full Scope (CEOG) ASME RA-S-2002 NEI 00-02 Rev. 0 AS 2002 Full Scope (CEOG) ASME RA-S-2002 NEI 00-02 Rev. 0 SC 2002 Full Scope (CEOG) ASME RA-S-2002 NEI 00-02 Rev. 0 SY 2002 Full Scope (CEOG) ASME RA-S-2002 NEI 00-02 Rev. 0 HR 2011 Focused (PWROG) ASME/ANS RA-Sa-2009 RG-1.200 Rev. 2 DA 2011 Focused (PWROG) ASME/ANS RA-Sa-2009 RG-1.200 Rev. 2 QU 2002 Full Scope (CEOG) ASME-RA-S-2002 NEI 00-02 Rev. 0 LE 2009 LERF (PWROG) ASME RA-Sb-2005 RG-1.200 Rev. 1 IF 2011 Focused (PWROG) ASME/ANS RA-Sa-2009 RG-1.200 Rev. 2 One additional focused peer review was conducted in December 2013 for Interfacing System LOCA (ISLOCA) analysis against ASME/ANS RA-Sa-2009 as endorsed by RG-1.200 Rev 2. The peer review identified 4 F&Os.

Further, a new self-assessment was conducted by FPL Staff to review elements of HLRs that were peer reviewed by earlier standards to identify any gaps or impact of any differences relative to the current standard (ASME/ANS RA-Sa-2009 as endorsed by RG-1.200 Rev 2). The self-assessment was particularly focused on IE, AS, SC, SY, QU, and LE HLR elements. To ensure these HLRs are consistent with and equivalent to the self-assessment process outlined in RG 1.200 Rev. 2 and to demonstrate technical adequacy of the PRA, clarifications were added using guidance in Table B-4 of Appendix B of RG-1.200 Rev. 2.

The self-assessment has concluded that no gaps were identified in compliance with ASME/ANS RA-Sa-2009 as endorsed by RG-1.200 Revision 2 for all SRs that have been reviewed by earlier standards, except those associated with the F&Os from the most recent focused scope peer review for ISLOCA. The SRs with potential gaps are IE-C5, IE-C6, IE-C9, IE-C 10, SC-A5, SY-A2, and QU-D2. Assessment of impact of the 4 ISLOCA F&Os on both internal events models and NFPA-805 fire models will be provided as part of the 120-days responses.

Enclosure I to L-2014-083 Page 16 of 21 PSL PRA RAI 17a Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting an FPP consistent with NFPA-805. RG 1.200 describes a peer review process utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established. The primary result of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os.

Clarify the following dispositions to fire F&Os and SR assessment identified in LAR Attachment U that have the potential to impact the fire PRA results and do not appear to be fully resolved:

a. AS-6 (From CEOG 2003 against HR-Fl, HR-F2, SC-A3, SC-A6, SY-A2, SY-A22, QU-Al)

The F&O recommends including credit for Low Pressure Feed from the condensate pumps for certain accident sequences. The F&O disposition explains that in response certain initiating events were combined and recovery actions were added. It is not clear how this modification addresses the F&O. Explain how apparent current IEPRA conservatism associated with modeling Low Pressure Feed affects the fire PRA risk estimates.

RESPONSE

The response to AS-6 F&O is revised as follows:

"The current revision of plant procedure (EOP-06) is listing the referenced action in step (9.B.3.1). No credit is currently considered for low pressure feed in scenarios associated with recoverable Loss of Main Feedwater (LOMFW). Current PRA results indicate that. recoverable LOMFW scenarios represent only about 2% and 5% of internal events CDF values in Unit I and Unit 2, respectively. Improved recovery of these scenarios will neither alter nor significantly change the result insights that would warrant spending further resources on."

The text above will replace the current text in Attachment 'U' of the St. Lucie LAR. See the markup to LAR page 'U-6'.

The Fire PRA does not credit use of condensate pumps and/or main feedwater pumps for secondary heat removal given the potential for loss of these systems due to unavailability of associated non-safety related support systems. Therefore, the treatment of Low Pressure Feed does not affect the fire PRA risk estimates.

Enclosure I to L-2014-083 Page 17 of 21 Florida Power & Ught Attachment U - Internal Events PRA Quality Table U St. Lucde PRA Facts and Observations F&O Source Description Significance Recommended Plant Resolutlon SR Impact on Resolution Fire PRA level. The level does not need 1o be ly restore 10 opera1ll1w Plant A3-81, slow sOutdowrn odian. The leve need only be aNove toward Soc. use of SC-M.

ft hot leI RWT ad IIAMs Is aU-Al.

requasa WMen RCS needs F mP do to smsMed Lev Orop 2250 psla at 600 F (0.0217 It-JN3m) t0 100 pSle at 30OF (0.017M stm). Given slnkiagn f ei Rx roloi RCS NMu* vo1ume of 10.400 flA3, ti s Wn"s trip. 711e reviewer ap'laxefy 18,500 galens are reluNk t reo1010 fe assumes toa OPS Wil Replace this text with the following text: conMue to SOC, even I ease if RW nonur were (7U c.wr) n io of plant procedure (oOP06) is listing the referenced actios step to *Aw Vg mkeaq (9.3.3.1). No cae*& is cumwtly considered foc low pressure feed in scenarios associated with to tw olPWtu m RCA coo"i cMOnlue at a recoverable Loss of Main Feedwater (LOMFW). Cmat PRA resalts indicate that rate of 100FRW, and RCS recoverable LOMFW scenario rmepmst only about 2% and 5% of oternal events Ct)? evel amps d~. to values in Ui 1and Unit 2,respectvely. Ioved recovery of thes scmios will neither aout 'gone lovw.

EOP-M2 SWe 4.5 Prevet alW nor pnficantly chma the result insit that would wasrat spemdinmg btlhm resotuces 19em1 1 V behavior by requirng; I*i OP ne The Fire PRA does not credit use of condensate pumps and/or main feedwater pumps for W3S kwenlory coNIN Is secondary heat removal given the potential for loss of these systems due to unavailability of leve IsrestOMe betee associated non-safety related support systems." 30% IND35%.SDC WO not CO"th" wum PZR leve isrest00d The0 oMnend sellwlois Woc~ as not redtal.

A" ClOG Consid adh low presure feed ("",,g cor"ensate LeOWs ConSdetoxing LOw LOMPW ofevents reatod It events11ve wor AS-A3, P50 Idl 2003 pumfs) tb Vie model for acdent spquences MWoN Pre-s-*e Feed Otm I* Corle" dift ned FWnW.Z a"eUpdale ASmU. No mpa IOM of Ot MFWIAPW. condensate ixrs in ~cnern mw aed a $woeselon Ifa If AS-A5 accident sequence IMt p~areview.

recove n weeoswere AS-A7.

Usb'g contlereae ptms, to feed to WSs is InboMh ocude TLOF. (Curnt model remains 00"0Wio based an VIC AS-WOP 6 -oW Loss of Free amd IO-.iS Tundloni 8 orn-) AIO.

Recovery Puscedwa. Opmetns Is dFlcte1 t11o ue IOW AS-8i.

pressue foed in i.EOP.46 (stO &.B.3.1). Ceing low- HR-El.

presme feed will envrwi twos Core damaMe HR*.F1 sequmes where ie MFW pnMs ae M. bute10 HR41Z condmnsie pw~s we ale. ft*V to TMs ae not SC-A.

avalMg, lIen IWe hot wee mWe-w conamh syStmm (or SC-A6, Page U-6 Revision 0 RevIsion 0 Page U-

Enclosure 1 to L-2014-083 Page 18 of 21 PSL PRA RAI 17b Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a fire PRA and endorses, with exceptions and clarifications, NEI 04-02, revision 2, as providing methods acceptable to the staff for adopting an FPP consistent with NFPA-805. RG 1.200 describes a peer review process utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established. The primary result of a peer review are the F&Os recorded by the peer review and the subsequent resolution of these F&Os.

Clarify the following dispositions to fire F&Os and SR assessment identified in LAR Attachment U that have the potential to impact the fire PRA results and do not appear to be fully resolved:

b. HR-G6 (From PWROG 2011 against HR-G6)

The F&O cites a lack of reasonableness check of dependency analysis results and very low joint HEPs (e.g., between 1E-10 to 1E-16). Section 6.2 of NUREG-1921 addresses the need to consider a minimum value for the joint probability of multiple HFEs.

Considering this guidance, confirm that joint HEP values that appear in PRA cutsets are appropriate.

RESPONSE

The detailed HRA analysis, including the details of dependency analysis, and joint HEP values thereof, has been independently reviewed by one FPL PRA staff engineer who used to hold a SRO-license and worked as a Shift Manager at St. Lucie. Given the various guidance associated with use of a floor value with understanding of potential impact of its application on overall results, there were, and still are, no issues identified by the independent reviewer as "unreasonable", and thus the current use of joint HEP is appropriate.

A sensitivity analysis was performed for Units 1 and 2 Internal Events results to evaluate the impact of artificially setting a floor value of 1E-05 or 1E-06 for joint HEPs, as suggested by the NRC references. Using a floor value of IE-05, the risk increase for Unit 1 has ranged from two to three folds for CDF and LERF, respectively, while Unit 2 increase was very small (less than 1%).

Using a floor value of 1E-06, the risk increase is significantly less than that of a IE-05 floor value.

Results of Unit I are summarized below.

Enclosure I to L-2014-083 Page 19 of 21 PSL Unit 1 Sensitivity Results CDF LERF Baseline (zero floor) 5.34E-06 per year 7.79E-07 per year New value (IE-05 floor) 1.57E-05 per year 2.86E-6 per year Percent Increase 195% 267%

Baseline (zero floor) 5.34E-06 per year 7.79E-07 per year New value (1 E-06 floor) 6.37E-06 per year 9.86E-07per year Percent Increase 19% 26%

Review of sequences associated with the increased risk indicated that these sequences are dominated by joint HEPs that are associated with long term cooling for Unit 1, while Unit 2 sequences are dominated by equipment failures with less dependency on human actions and joint HEPs thereof.

Sequences associated with long term cooling in Unit 1 are dependent on cross-connection to Unit 2 condensate water storage resources and the operator actions associated with establishing such function. The Unit 1 Condensate Storage Tank (CST) is much smaller than the Unit 2 CST, and thus when Unit 1 CST is depleted and long term makeup from the Treated Water Storage Tank (TWST) fails, the operator will have to open a cross-connection between Unit 1 and Unit 2 CSTs so that the Unit 1 CST takes supply of water reserved for "Unit 1 use only" in the Unit 2 CST.

These actions have a long available timeframe for execution and thus are originally considered with very low dependency and very low joint HEP, as indicated in the RAI.

The Unit 1 CST is expected to deplete in over 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. The TWST has a large capacity and thus makeup from TWST to CST will take many hours as well. If makeup fails or is exhausted, action to cross-connect with the Unit 2 CST is considered. However, while proceeding for long term cooling, substantial time exists between execution of individual actions within the joint event.

This can be credited as "very low dependency" in accordance with EPRI 1021081 guidance. Given the timing between the actions within the joint event, and the low dependency between them, a joint HEP engulfing these actions with very low probability is reasonable and appropriate.

If such dependency is artificially increased outside the uniformed rule of the associated dependency analysis to the level that the respective HEPs are relatively increased to IE-05 or IE-06 probability, it is reasonable to observe such increase in risk, summarized above, as a direct impact of replacing the original lower probability value of the respective HEPs with an arbitrary value with several orders of magnitude based on weak or no technical basis.

FPL has refuted arguments presented by defenders of maintaining floor values in the dependency analysis who provided examples, during industry/regulatory meetings, involving HRA for conditions involving concurrent transient event and heart attack, workplace violence, and misdiagnosis by operators. EPRI HRA User's group has formed a subcommittee to research and advise the industry, including NRC, on this same issue. No conclusion has yet been reported on the subject.

Enclosure I to L-2014-083 Page 20 of 21 PSL RR RAI 01 Where containment and confinement cannot be relied upon, describe whether an analysis [has]

been performed to demonstrate that the instantaneous release limits in the Technical Specifications or other 10 CFR 20 limits will not be exceeded.

For example, in reference to "RCA-Yard" area, the LAR states, "[...] there are no engineering controls to prevent radioactive release [...]." The LAR then states, "[...] fire suppression activities

[in "RCA-Yard" areas] will not result in a radioactive release that exceeds the 10 CFR 20 limits."

The LAR states, "The evaluations performed to meet the radioactive release performance criteria are qualitative in nature." Provide additional information on either qualitative assessment or a quantitative assessment.

RESPONSE

ML102920405, NRC Close Out of FAQ 09-0056, indicates that compliance with the radioactive release goals, objectives and performance criteria can be demonstrated by providing an analysis that the instantaneous release of radioactive effluents specified in the unit's Technical Specifications are met.

The St. Lucie Technical Specifications (TS) include requirements for a Radioactive Effluent Controls Program and identify that the requirements will be contained in the St. Lucie Offsite Dose Calculation Manual (ODCM). The TS for each Unit identifies the limitation for liquid effluent release to unrestricted areas in Section 6.8.4.f(2) and identifies the limitations for gaseous effluent release to areas at or beyond the site boundary in Section 6.8.4.f(7).

PSL performed an analysis for the RCA-Outdoors/Yard compartment based on a Package Characterization Report created as a test case which presented a worse-case size Sea-Land trailer based on the dry active waste (DAW) shipments during the past 15 months. Calculations in the analysis used the methodology defined in the ODCM.

The calculation is considered bounding based on the following discussion.

10 CFR 61 addresses the Licensing Requirements for Land Disposal of Radioactive Waste.

Based on 10 CFR 61, the maximum dose rate on any shipping container transporting DAW is 200 mR/hr on contact. A bounding dose rate of 180 mR/hr for calculations prepared to support this RAI was used, which is a substantially higher dose rate than St. Lucie would store or ship containers. St. Lucie procedure RP-AA-108-1002 identifies that the administrative limits on a package or a vehicle dose rate is 80% of the DOT limit and that this limit may be exceeded only with the written approval of Radiation Protection Supervision. This would equate to a contact dose rate reading greater than 160 mR/hr. As the calculation prepared to support this response is based on 180 mR/hr, this calculation is bounding.

a Data on the radionuclides and their activity used to develop the release calculations for this RAI response are based on the 2011 DAW 10CFR61 waste stream data. 10 CFR61 is Licensing Requirements for Land Disposal of Radioactive Waste. The 2011 waste stream data is based on analysis performed a population of smears taken from various areas of the radiologically controlled area at St. Lucie in order to identify the radionuclides and their activity. The data is collected and analyzed to meet the requirements of 10CFR6 1. This is considered bounding as it is representative of the radionuclides and activity present on the St.

Lucie Site.

Enclosure Ito L-2014-083 Page 21 of 21 Florida State DOT regulations limit the gross weight for vehicles on the roadways in Florida to 80,000 lbs. The truck and trailer which typically transports waste from St. Lucie is approximately 40,000 lbs. unloaded. The truck trailer can haul a maximum of two 20' Sea-Land Containers which equates to a maximum of 20,000 lbs. each beyond which the road design basis would be exceeded.

" Review of shipping data utilized a shipping container that presented the maximum size Sea-Land Container that is stored on site for the transport of waste and was associated with a shipment from January 1, 2013. St. Lucie ships waste from the site in containers with a typical range of 8,000 lbs. - 12,000 lbs. Characterization of the shipping container for this response is based on a value of 11,000 lbs of waste. The 11,000 lb. value is greater than the average of the range and results in a total net container weight of 16,860 lbs. which is 85% of the maximum allowed weight of 20,000 lbs. for this type of container and is considered conservative.

" The 2011 nuclide/activity analysis, conservative surface and 1 meter dose rates and largest Sea-Land trailer shipment data from 2013 were then input into WMG's RadMan software.

WMG's RadMan software is industry/NRC accepted/recognized software which characterizes and classifies, manifests and documents packaged radioactive waste for shipment within the current DOT and NRC regulatory requirements. This creates a conservative nuclide inventory and activity values for a DAW shipment in a conservatively packed Sea-Land container.

The calculation was based on two different fire scenarios involving a Sea-Land trailer, containing dry active waste (DAW) stored in the Radiological Control Area (RCA) Yard Area prior to shipment offsite. Noble gasses are not present/detected in containers of DAW and therefore not included in the analysis/calculation. The two scenarios, and their results, are:

1) Ignition followed by a 15 minute pre-bum time with intervention by the Fire Brigade which would require a 20 minutes discharge of 125 GPM liquid effluent from firefighting activities released off site into the Discharge Canal prior to processing.

• The calculation determined that the Fraction of Unrestricted Area Effluent Concentration from the postulated DAW fire would be less than 0.02 and the administrative limit is 0.80.

" The calculation determined that the total dose that would be received from the release associated with the postulated DAW fire would be less than 1 mR which is less than 0.1%

of the allowed annual limits. No release rate or dose limits would be exceeded due to the postulated scenario and,

2) Ignition with no intervention by the Fire Brigade and release to atmosphere of all airborne particulate within 15 minutes.

The results of the calculation determined that if the contents of the Sea-Land container were to go airborne and drift with the wind to an area outside the site boundary that the expected dose would be less than 3 mR which based on the isotopes surveyed would result in less than 20% of the annual regulatory limit and therefore, dose limits would not be exceeded.