NRC 2014-0056, Response (120 Day) to Request for Additional Information and Revision to 60 Day Response License Amendment Request 271 Associated with NFPA 805

Response (120 Day) to Request for Additional Information and Revision to 60 Day Response License Amendment Request 271 Associated with NFPA 805

ML14282A447
Person / Time
Site:	Point Beach
Issue date:	09/25/2014
From:	Mccartney E Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2014-0056, TAC MF2372, TAC MF2373
Download: ML14282A447 (69)

Text

NEXTera, ENERGY~

~

September 25, 2014 NRC 2014-0056 10 CFR 50.90 10 CFR 2.390 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Response (120 Day) to Request for Additional Information and Revision to 60 Day Response License Amendment Request 271 Associated with NFPA 805

References:

(1) NextEra Energy Point Beach, LLC, letter to NRC, dated June 26, 2013, "License Amendment Request 271, Transition to 10 CFR Q0.48(c)-

NFPA 805, 'Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,' 2001 Edition" (ML131820453)

(2) NRC e-mail to NextEra Energy Point Beach, LLC, dated September 9, 2013, "Request for Supplemental Information Regarding the Acceptability of the Proposed Amendment Request" (ML13256A197)

(3) NextEra Energy Point Beach, LLC, letter to NRC, dated September 16, 2013, "License Amendment Request 271 Supplement 1 Transition to 10 CFR 50.48(c)- NFPA 805" (ML13259A273)

(4) NRC letter to Next Era Energy Point Beach, LLC, dated September 25, 2013, "Point Beach, Units 1 and 2- Acceptance Review of Licensing Action re:

License Amendment Request to Transition to NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (ML13267A037)

(5) NRC e-mail to NextEra Energy Point Beach, LLC, dated July 8, 2014, "Point Beach Nuclear Plant, Units 1 and 2 - Final (Revised) Requests for Additional Information re: License Amendment Request Associated with NFPA 805

{TAC Nos. MF2372 and MF2373)" (ML14189A365)

(6) NextEra Energy Point Beach, LLC, letter to NRC, dated July 29, 2014, "Response (60 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805" (ML14210A645)

(7) NextEra Energy Point Beach, LLC, letter to NRC, dated August 28, 2014, "Response (90 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805" (ML14241A267)

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 Security-Related Information- Withhold from Public Disclosure Under 10 CFR 2.390:

Attachments 3, 4, 5, 6, and 7 of this letter contain security-related information. Upon removal of Attachments 3, 4, 5, 6, and 7 of this letter, this letter is uncontrolled.

Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC, (NextEra) requested to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2 (Reference 1 and supplemented via Reference 3). The NRC accepted the license amendment request (LAR) for review in Reference 4.

The NRC Staff has determined that additional information (Reference 5) is required to complete its evaluation of the license amendment request. The 60 day and 90 day responses to the Request for Additional Information (RAI) were submitted in References 6 and 7, respectively. provides several revised 60 day RAI responses and Enclosure 2 provides the NextEra response to the NRC Staff's request for additional information for the required 120 day response.

This response includes revisions to the original LAR Attachment G, Recovery Actions Transition; AttachmentS Table S-2, Plant Modifications Committed; and Attachment W, Fire PRA Insights, among other changes documented in Enclosures 1 and 2. NextEra continues to meet the regulatory requirements for the transition of its fire protection licensing basis. This response to the RAI does not alter the conclusions submitted with the original request (Reference 1) that the license amendment does not present a significant hazards consideration and the criteria have been met for categorical exclusion from the need for an environmental assessment.

Note that several planned plant modifications modeled in the Fire Probabilistic Risk Assessment (FPRA) and internal events PRA for risk reduction involve electrical or mechanical cross-ties between independent safety-related systems. LAR AttachmentS, Table S-3, Implementation Items, commits to incorporating the use of these cross-ties into abnormal operating procedure guidance (e.g., for loss of feedwater, loss of service water, or loss of power events). These procedure revisions will require evaluation under 10 CFR 50.59 and may result in the requirement to submit separate license amendment requests for prior NRC approval.

This letter contains no new Regulatory Commitments and no revisions to existing Regulatory Commitments.

NextEra requests that Attachments 3, 4, 5, 6, and 7 of this letter, which contain sensitive security-related information, be withheld from public disclosure in accordance with 10 CFR 2.390.

If you have any questions regarding this letter, please contact Mike Millen at (920) 755-7845.

NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 Security-Related Information- Withhold from Public Disclosure Under 10 CFR 2.390:

Attachments 3, 4, 5, 6, and 7 of this letter contain security-related information. Upon removal of Attachments 3, 4, 5, 6, and 7 of this letter, this letter is uncontrolled.

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

Executed on September 25, 2014.

Very truly yours, NextEra Energy Point Beach, LLC

/::~~

Site Vice President Enclosures 1 and 2 Attachments 1 - 7 cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 Security-Related Information- Withhold from Public Disclosure Under 10 CFR 2.390:

Attachments 3, 4, 5, 6, and 7 of this letter contain security-related information. Upon removal of Attachments 3, 4, 5, 6, and 7 of this letter, this letter is uncontrolled.

ENCLOSURE 1 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT REVISED RESPONSE (60 DAY) TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 271 ASSOCIATED WITH NFPA 805 Pursuant to 10 CFR 50.90, NextEra Energy Point Beach, LLC, (NextEra) requested to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2 (Reference 1 and supplemented via Reference 3). The NRC accepted the license amendment request for review in Reference 4.

The NRC Staff has determined that additional information (Reference 5) is required to complete its evaluation of the license amendment request. This Enclosure 1 provides revised NextEra responses to selected NRC Staff's requests for additional information for the 60 day response, which was previously submitted in Reference 6. The NOTES in front of each response provide the reason for the revision of each RAI response.

PBNP RAI SSA 01 a Safe Shutdown Analysis The NRC staff noted that some implementation items and modifications are referenced in the LAR, but not in sufficient detail to determine what particular implementation item or modification relates to the proposed change. For the item listed below, provide the following:

a) Identify the specific modifications in LAR Attachment S that correlate with elements 3.5.2.4 and 3.5.2.5 of LAR Attachment 8, Table 8-2. The alignment basis in LAR Attachment 8, Table 8-2, for elements 3.5.2.4 and 3.5.2.5 refer toLAR AttachmentS, Table S-2 for modifications associated with circuit coordination and common enclosure criteria. There are several modifications in LAR Attachment S associated with circuits, breakers and fuses.

NextEra Response a) **NOTE: The 60-day RAI response provided by letter dated July 29, 2014 (Reference 6) is superseded to reflect LAR Attachment S updates in response to PRA RAI-25.**

LAR Attachment B, Element 3.5.2.4, "Circuit Failures Due to Inadequate Circuit Coordination", indicates four (4) references to AttachmentS.

The specific modifications in LAR Attachment S that correspond with these references are as follows:

• MOD-26-1 for 480V Motor Control Centers - MCC B-21 Coordination (Page B-85)

• MOD-26-3 for 120VAC Distribution Panel Coordination (Page B-85)

• MOD-26-3 for 120VAC Safety Related Instrument Bus Coordination (Page B-86)

• MOD-26-3 for 120VAC Branch Circuit Coordination (Page B-86)

LAR Attachment B, Pages B-85 and B-86 are revised to add the applicable MOD number to each AttachmentS statement (see Attachment 2).

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AttachmentS, Pages S-16 and S-17 are revised to clarify the specific MOD-26 applicable to these concerns. The updated LAR AttachmentS, Table S-2 is provided in the response to PRA RAI-25 (see Attachment 5).

LAR Attachment B, Element 3.5.2.5, "Circuit Failures Due to Common Enclosure Concerns", indicates two (2) references to AttachmentS.

The specific modifications in LAR Attachment S that correspond with these references are as follows:

• MOD-24 for cables not protected for overload (Page B-88)

• MOD-26-1 and MOD-26-3 for 480V MCCs and Power Panels and 208/120 Lighting Panel cable protection (Page B-88)

LAR Attachment B, Page B-87 is revised to update Alignment Basis related to 13kV and 4kV Switchgear discussion, and Page B-88 is revised to add the applicable MOD number to each AttachmentS statement (see Attachment 2).

PBNP RAJ SSA 01 b Safe Shutdown Analysis The NRC staff noted that some implementation items and modifications are referenced in the LAR, but not in sufficient detail to determine what particular implementation item or modification relates to the proposed change. For the items listed below, provide the following:

b) Identify the implementation item(s) that address the revision to the training program and drill procedures to incorporate the feasibility evaluation results. LAR Attachment G, under the heading, "Results of Step 4," describes implementation items resulting from the feasibility evaluation including revision to the training program and revision to the drill development procedure and states these items are included in LAR Attachment S.

NextEra Response b) **NOTE This RAI response is resubmitted to correct a typographical error in the 60-day response dated July 29, 2014 (Reference 6), which included the first paragraph of this response with the previous (RAI SSA-01.a) response. An update to IMP-152 (underlined in last paragraph) is the only other update. **

LAR Attachment G, in "Results of Step 4" states that "[i]implementation items resulting from the feasibility evaluation are included in the corrective action program.

These items include:

• Development/revision of procedures

• Revisions to the Training Program to reflect procedure changes

• Revision to the drill development procedure These items are included in Table S-3."

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Several Table S-3 Implementation Items are related to the Feasibility Evaluation, to include:

~ IMP-135 (pg. S-27) Fire protection program documents will be updated and training will be provided as necessary. This includes fire protection design basis document, system-level design basis documents and procedures, the Fire Protection Evaluation Report (FPER), the Fire Hazards Analysis Report (FHAR),

the Safe Shutdown Analysis Report (SSAR), post transition change process (including Fire PRA updates), and qualification training. This is being tracked by NAMS Action Request 1882226.

~ IMP-143 (pg. S-29) A confirmatory demonstration (field validation walk-through) of the 4.2.1.3 and Attachment G feasibility for the credited NFPA 805 Recovery Actions (RA) will be performed. This will include field validation of:

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

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

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

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

This is being tracked by NAMS Action Request 1882226.

Note that IMP-152 is updated to:

~ IMP-152 (pg. S-37) Procedure FOP 1.2, "Potential Fire Affected Safe Shutdown Components," will be revised from guideline format to utilize a procedure-type format; and requisite training will be performed for the revised procedure once formally issued. The feasibility for each fire-specific safe and stable action, including a formal walk-through and a timing evaluation, will be evaluated and documented. Also. update training processes to provide clarification on drills for recovery actions. This is being tracked by NAMS Action Request 1882226.

PBNP RAI SSA 01 c Safe Shutdown Analysis The NRC staff noted that some implementation items and modifications are referenced in the LAR, but not in sufficient detail to determine what particular implementation item or modification relates to the proposed change. For the item listed below, provide the following:

c) In LAR Attachment C, the Fire Risk Summary for Fire Areas A01-B/46, A23N, and A36, states, in part, those with the proposed cable protection in Attachment S, the applicable risk, defense-in-depth, and safety margin criteria were satisfied. There were no VFDR dispositions identified in these fire areas that describe modifications.

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i. Confirm the modifications referenced in the Fire Risk Summary for the individual fire areas are not associated with a VFDR disposition.

ii. Identify the specific modification(s) item in LAR Attachment S that is/are associated with the risk summaries in Attachment C for these areas.

NextEra Response c) **NOTE: The 60-day response provided by letter dated July 29, 2014 (Reference 6) is superseded to reflect LAR AttachmentS updates in response to PRA RAI 25.**

The RAI inquires about three (3) specific Fire Areas.

~ Fire Area A01-B/46- Proposed cable protection referenced in the Fire Risk Summary for Fire Area A01-B/46 (MOD-18) is not related to VFDR dispositions. Per the Fire Risk Evaluation, it is a risk reduction modification. Subsequent to the LAR submission, additional refinements to the Fire PRA model and risk insights concluded that MOD-18 is unnecessary to support transition to NFPA 805. Refer to the response to RAI PRA 25, Changes to Modifications Described in LAR Attachment S.

The Fire Risk Summary for Fire Area A01-B/46 (pg. C-53) is revised (see Attachment

3) to remove reference to "proposed cable protection in Attachment S".

~ Fire Area A23N - Proposed cable protection referenced in the Fire Risk Summary for Fire Area A23N (MOD-12) is not related to VFDR dispositions. Per the Fire Risk Evaluation, it is a risk reduction modification.

MOD-17 protects multiple cables in various Fire Areas to preserve DC control power to multiple buses to ensure the ability of 4160V breakers to trip on an overcurrent (OCT) condition (see detailed response and discussion of OCT modifications in the response to RAI SSA 05. Some of the cables to be protected are in Fire Area A23N.

Cable ID Assoc. VFDR Remarks ZFD0206A A23N-27 VFDR disposition does not mention cable protection since it does not fully resolve VFDR.

ZFD1402A1/A2 A23N-10 VFDR disposition does not mention cable protection since it does not fully resolve VFDR.

ZFD0203A A23N-31 New VFDR as a result of EC-261 022 and further evaluation under RAI SSA-05.

ZFD0208A A23N-31 New VFDR as a result of EC-261022 and further evaluation under RAI SSA-05.

Subsequent to the LAR submission, additional refinements to the Fire PRA model and risk insights concluded that MOD-12 and MOD-17 are unnecessary to support transition to NFPA 805. Refer to the response to RAI PRA 25, Changes to Modifications Described in LAR Attachment S.

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The Fire Risk Summary for Area A23N (pg. C-264) is revised to remove reference to "proposed cable protection in AttachmentS".

~ Fire Area A36 - Proposed cable protection referenced in the Fire Risk Summary for Fire Area A36 (MOD-8, MOD-9) are not related to VFDR dispositions. Per the Fire Risk Evaluation, these are risk reduction modifications. MODs 8 and 9 intend to protect RCS pressure instrument cables in Fire Area A36 to protect against spurious opening of the pressurizer PORVs. Since the inadvertent operation can be mitigated by a Control Room action, the failure is not associated with a VFDR and the modification is only for risk reduction purposes. Subsequent toLAR submission, additional refinements to the Fire PRA model and risk insights concluded that MOD-8 and MOD-09 are unnecessary to support transition to NFPA 805. Refer to the response to RAI PRA 25, Changes to Modifications Described in LAR Attachment S.

The Fire Risk Summary for Area A36 (pg. C-376) is revised to remove reference to "proposed cable protection in Attachment S".

PBNP SSA RAI 03(b)

For those fire areas that credit electrical raceway fire barrier system (ERFBS) as described in LAR Attachment C:

b) If credited for dispositioning a VFDR, provide a discussion of the analysis or basis for the acceptability of the ERFBS in resolving the VFDR.

NextEra Response b) **NOTE: The 60-day response provided by letter dated July 29, 2014 is superseded to reflect LAR AttachmentS updates in response to PRA RAI 25.**

The following VFDRs are dispositioned by proposed modifications to protect cable(s); these modifications may utilize raceway protection (ERFBS) or explore other cable protection options (e.g., fire rated cable):

• A01-B-64 in LAR Attachment C (pg. C-38), refer to MOD-20 in AttachmentS (pg. S-11)

• A15-16 in LAR Attachment C (pg. C-210), refer to MOD-11 in AttachmentS (pg. S-7)

• A30-06 in LAR Attachment C (pg. C-324), referred to proposed cable protection as described in MOD-16. Subsequent toLAR submission, additional refinements to the Fire PRA model and risk insights concluded that MOD-16 is unnecessary to support transition to NFPA 805. Refer to the response to RAI PRA 25, Changes to Modifications Described in LAR Attachment S.

• A30-22 in LAR Attachment C (pg. C-328), referred to proposed cable protection as described in MOD-16. Subsequent to LAR submission, additional refinements to the Fire PRA model and risk insights concluded that MOD-16 is unnecessary to support transition to NFPA 805. Refer to the response to RAI PRA 25, Changes to Modifications Described in LAR Attachment S.

Modification design processes (i.e., fire protection and safe shutdown design reviews) will ensure compliance with NFPA 805 Section 4.2.3 for acceptability of resolving each VFDR.

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References (1) NextEra Energy Point Beach, LLC, letter to NRC, dated June 26, 2013, "License Amendment Request 271, Transition to 10 CFR 50.48(c)- NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants,"

2001 Edition" (ML131820453)

(2) NRC e-mail to NextEra Energy Point Beach, LLC, dated September 9, 2013, "Request for Supplemental Information Regarding the Acceptability of the Proposed Amendment Request" (ML13256A197)

(3) NextEra Energy Point Beach, LLC, letter to NRC, dated September 16, 2013, "License Amendment Request 271 Supplement 1 Transition to 10 CFR 50.48(c)- NFPA 805" (ML13259A273)

(4) NRC letter to NextEra Energy Point Beach, LLC, dated September 25, 2013, "Point Beach, Units 1 and 2 - Acceptance Review of Licensing Action re: License Amendment Request to Transition to NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (ML13267A037)

(5) NRC e-mail to NextEra Energy Point Beach, LLC, dated July 8, 2014, "Point Beach Nuclear Plant, Units 1and 2 - Final (Revised) Requests for Additional Information re:

License Amendment Request Associated with NFPA 805 (TAC Nos. MF2372 and MF2373)" (ML14189A365)

(6) NextEra Energy Point Beach, LLC, letter to NRC, dated July 29, 2014, "Response (60 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805" (ML14210A645)

Page 6 of 6

ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT REVISED RESPONSE (120 DAY) TO REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST 271 ASSOCIATED WITH NFPA 805 Pursuant to 10 CFR 50.90, Next Era Energy Point Beach, LLC, (NextEra) requested to amend renewed Facility Operating Licenses DPR-24 and DPR-27 for Point Beach Nuclear Plant (PBNP), Units 1 and 2 (Reference 1 and supplemented via Reference 3). The NRC accepted the license amendment request for review in Reference 4.

The NRC Staff has determined that additional information (Reference 5) is required to complete its evaluation of the license amendment request. This Enclosure 2 provides the NextEra response to the NRC Staff's request for additional information for the 120 day response.

References 6 (as amended in Enclosure 1) and 7 provided the 60 day and 90 day responses to the NRC Staff's request.

Probabilistic Risk Assessment (PRA) RAJ 01 - Fire PRA Facts and Observations (F&Os)

In Section 2.4.3.3 of National Fire Protection Association Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition (NFPA 805), it states that the probabilistic safety assessment (PSA) (PSA and PRA are synonymous) approach, methods, and data sha/1 be acceptable to the authority having jurisdiction (AHJ), which is the U.S. Nuclear Regulatory Commission (NRC).

Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," identifies NUREG/CR-6850, "EPRIINRC-RES Fire PRA Methodology for Nuclear Power Facilities," as documenting a methodology for conducting a fire PRA (FPRA) and endorses, with exceptions and clarifications, Nuclear Energy Institute (NEI) 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Revision 2, as providing methods acceptable to the staff for adopting a fire protection program (FPP) consistent with NFPA 805.

RG 1.200, '~n Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes a peer review process utilizing an associated American Society of Mechanical Engineers/American Nuclear Society (ASMEIANS) standard (currently ASME/ANS-RA-Sa-2009, '~ddenda to ASME/ANS RA-S-2008, Standard for Level 1 I 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 for evaluations that could influence the regulatory decision. The primary result of a peer review is the F&Os recorded by the peer review and the subsequent resolution of these F&Os.

Clarify the fa/lowing dispositions to fire F&Os and Supporting Requirement (SR) assessment identified in License Amendment Request (LAR) (Agency wide Documents Access and Management System (ADAMS) Accession No. ML13182A353) Attachment V that have the potential to impact the Fire PRA (FPRA) results and do [not] appear to be fu/ly resolved:

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a) CF-81-01 (Not Met)

For components for which information regarding cable housing and insulation was not readily available, the licensee's analysis indicated that Option #2 from NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," Section 10.5.3.2 was used to quantify the likelihood of hot short-induced spurious operations. Based on its review, the NRC staff concludes that this approach does not provide an adequate method for quantifying the likelihood of hot short-induced spurious operations for components for which information regarding cable housing and insulation was not readily available.

Replace this approach and provide an explanation of the approach used and the results in sufficient detail so the staff can make a conclusion regarding the use of the approach.

b) CS-C3-01 (Not Met)

This F&O suggested that the assumed cable routing for the turbine stop and steam dump valves was either not modeled or not documented. The disposition did not directly indicate that this would be done, rather circuitry would be modified (see LAR AttachmentS, Table S-2, MOD-14).

Clarify how the modification resolves the concern identified in the F&O.

c) FQ-A4-01 (Not Met)

The disposition to F&Os FQ-A4-01, IGN-A10, and UNC-A 1 indicate that integrated uncertainties were not performed since several of the key factors such as cable damage, zone of influence, and spurious short durations cannot be carried forward and estimated in the codes (UNCERT). The licensee concluded that because of this, the risk insights from available codes that can be gained to estimate mean CDF and LERF is minimal. SR FQ-A4-01 stipulates that HLR-QU-A SRs from Part 2 of the American Society of Mechanical Engineers/American Nuclear Society (ASMEIANS) PRA standard be applied to FPRA quantification. SR QU-A3 requires that the estimation of the mean core damage frequency (CDF) should account for the state of knowledge correlation (SOKC). The NRC staff noted that because parameter uncertainty was not propagated, the FPRA quantification did not account for the impact of the SOKC on the estimate of CDF and large early release frequency (LERF).

Explain how the disposition of these F&Os which indicates that integrated uncertainties have not been performed, account for SOKC in the risk estimates and address SOKC for component failure types, fire ignition frequency, circuit failure probability, and non-suppression probability.

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d) FQ-F1-05 (Not Met)

The disposition to this F&O states that the success of instrument air was not credited in the model, except in cases where the assumption that air was failed provided a non-conservative input to the model. The staff notes that conservative system modeling can lead to calculation of non-conservative delta (b.) CDF and b. LERF if there are variances from deterministic requirements (VFDRs) associated with the system. LAR Attachment C indicates that instrument air is credited in the shutdown path and is associated with a VFDR (i.e., A32-22).

Explain whether conservative modeling of the compliant plant could underestimate b. CDF or b. LERF. If so, address this conservatism as part of the integrated analysis performed in response to PRA RAI 3.

e) FSS-A1-01 (Met)

The disposition to this F&O states that the basis for eliminating fire scenarios involving junction boxes has been documented. The licensee's analysis states that "Junction boxes are robustly secured and well-sealed, and therefore are screened as non-damaging ignition sources." Based on its review, the NRC staff concludes that unlike electrical cabinets, there is no exclusion of a junction box from the count because it is robustly secured and well-sealed; therefore, junction boxes that route FPRA target cables that can contribute to fire risk should not be excluded as ignition sources.

Replace this approach and provide an explanation of the approach used and the results in sufficient detail so the staff can make a conclusion regarding the use of the approach.

f) FSS-C5-01 (CC-I/)

The disposition to this F&O states that this F&O will be resolved when FAQ 13-0004, "Clarifications on Treatment of Sensitive Electronics," is closed by NRC.

FAQ 13-0004 was closed on December 3, 2013 (ADAMS Accession No.

ML13322A085) after the LAR was submitted.

Since the FAQ is now closed, explain how the treatment of sensitive electronics performed for the FPRA is consistent with the guidance in FAQ 13-0004.

Include in the response, how mounting sensitive electronic components on the surface of cabinets and also how the presence of louvers or other typical ventilation means were considered in the determination of damage conditions for sensitive electronic equipment enclosed in cabinets.

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g) HRA-A2-01 (Not Met)

The F&O noted that some new fire-specific human failure events (HFEs) had not been assessed consistent with the supporting requirements. The NRC Staff recognizes that meeting the supporting requirements will not be possible until the procedures are completed, but some evaluation has been performed to quantify the PRA. Summarize how new HFEs including those as yet incomplete have been evaluated by addressing:

i. How feasibility assessment of credited HFEs is determined ii. How the operator response procedure (or draft procedure) used as the basis for the credited HFEs was evaluated for consistency with plant observations.

iii. How the plant response modeling in the FPRA associated with credited HFEs was reviewed with plant staff.

h) HRA-A3-01 (CC-1)

As noted in the F&O, the licensee's analysis provides only a list of annunciator procedures associated with the control boards of interest, and does not give any information associated with the automatic actuations resulting from the annunciator or the instruments of interest. SR HRA-A3 requires identification of undesired operator actions that might result from spurious indication of a single instrument. It is not clear whether the licensee's analysis is for annunciators only, or also includes control indication (e.g., main control board (MCB) instrument indications). Explain the following:

i. How the analysis (i.e., use of the categories: proceduralized check/verify; multiple spurious indications on redundant channels/parameters; systems, structures, and components (SSCs) not credited; spare; and other) supports the conclusion that spurious indications on an instrument cannot lead to an undesired operator action.

ii. Whether all instruments used by the plant operators to inform actions were addressed in the analysis. If not, provide justification for the excluded instruments.

ij HRA-02-01 (Met)

The disposition to this F&O states that since the peer review, a human reliability analysis (HRA) dependency analysis was performed and documented using a minimum joint human error probability (HEP) "floor" of 1E-6. Based on the NRC staff's review, the licensee has not provided sufficient justification for a floor less than 1E-5.

Provide justification for the 1E-6 minimum joint HEP for each HEP below this minimum value that is used in the FPRA.

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NextEra Response a) The circuit failure mode likelihood analysis (CFMLA, Task 10 of NUREG/CR-6850) in the FPRA was reevaluated using only Option #1, which incorporates probability estimates from tables in NUREG/CR-6850 that are based on industry tests. As per RAI PRA 04, no control power transformer (CPT) credit was provided during this analysis. The results of this update are evaluated and discussed in the integrated analysis as part of the response to PRA RAI 3.

b) The F&O is based on the Multiple Spurious Operation (MSO) Review as documented in EPM Report P2902-11 OA-001, Appendix A. Appendix A, PWROG Scenario Number 23 of this review, "Decay Heat Removal- Failure to close or spurious opening of MSIVs with concurrent failure of downstream relief valve(s) to close", indicates a failure to close the MSIV's concurrent with the spurious opening of valve(s) for downstream steam load(s) (e.g., condenser steam dumps, turbine inlet valves, some atmospheric relieve/dump valves, etc.)

may lead to an excessive cool-down event. Point Beach conservatively assumed this condition to result in core damage. LAR, AttachmentS, TableS-2 (Reference 1), "Plant Modifications Committed", has been revised to add MOD 14 (see Attachment 5) to ensure isolation of the main steam flow path to resolve the concern by reducing the core damage frequency for risk reduction.

The addition of MOD 14 addresses MSO 23 such that precise cable routings for the turbine stop and steam dump valves are no longer needed.

c) As part of the "NFPA 805 Fire PRA Quantification Notebook" (P2091-2900-02) update to Revision 2, a section on uncertainty is included to address the state of knowledge correlation (SOKC) and integrated uncertainties. The SOKC is addressed using EPRI UNCERT software, as the software uses the component failure type codes found in the CAFTA database to calculate the probability distribution of the overall FPRA. In addition to the base PRA data, the database includes uncertainty parameters for fire specific events for ignition frequencies, non-suppression probabilities, and hot short induced spurious operations.

The UNCERT analysis provides a mean core damage frequency (CDF)/Iarge early release frequency (LERF) with uncertainty that accounts for SOKC and the uncertainty parameters for fire specific events for ignition frequencies, non-suppression probabilities, and hot short induced spurious operations. This satisfies Category II for SR QU-A3, FQ-A4, IGN-A10 and UNC-A1.

d) Instrument air was not credited in the Variant Model, except in cases where the assumption that air was failed provided a non-conservative input to the model.

Instrument air was credited in the Compliant Model, except when instrument air was failed as a result of the fire. This prevented underestimating delta CDF or delta LERF.

Therefore, this treatment does not need to be addressed further as part of the integrated analysis in the response to PRA RAI 3.

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e) The junction box treatment has been updated using an approach consistent with Fire PRA FAQ 13-0006, "Modeling Junction Box Scenarios in a Fire PRA."

Since junction boxes are not easily identified by identification number in a cable route and are not all assigned a fire zone in the Point Beach cable and raceway system, the approach in Section 3.2 of Fire PRA FAQ 13-0006 will be utilized.

The Junction Box Fire Ignition Frequency for each fire compartment is determined using the Bin 18 value from NUREG/CR-6850 Supplement 1, and the cable loading fractions developed in P2091-2600-01, "Fire PRA Notebook-Compartment Analysis", Revision 2, which is identical to the process used for self-ignited cable tray fires.

The Conditional Core Damage Probabilities (CCDPs) were evaluated for each of the junction boxes with a damage set based on raceway to cable data.

Because the specific locations of the Junction Boxes are not consistently identified, the CCDP of the most risk significant junction box for each unit was conservatively applied to each fire compartment as documented in the "NFPA 805 Fire Probabilistic Risk Assessment Quantification Notebook."

The junction box scenarios have been added to the base model in the "NFPA 805 Fire Probabilistic Risk Assessment Quantification Notebook" (P2091-2900-02). The total CDF contribution from the Junction Box fires to each unit is less than 2E-6/yr, with the highest individual Junction Box CDF less than 3E-7/yr.

f. Temperature sensitive electronic equipment is considered to be any equipment that is susceptible to lower thermal damage thresholds (i.e., solid-state control components). As such, all components in an analyzed compartment have been examined to determine whether they may be immersed within temperature exposures above the damage threshold recommended by NUREG/CR-6850. Plant walkdowns were performed in the fire compartments to identify those cabinets potentially containing PRA credited sensitive electronics. Safe shutdown information (i.e., component location, including cabinet information) combined with the visual inspections of the fire compartments confirmed if any credited components were located inside of cabinets. Walkdowns also confirmed if sensitive equipment was located close enough to the floor level to avoid hot gas layer immersion. This identification is consistent with the guidance in NUREG/CR-6850, "EPRIINRC-RES Fire PRA Methodology for Nuclear Power Facilities," and Fire PRA FAQ 13-0004.

Damage to temperature sensitive plant equipment caused by radiant heat from a fire is bounded by the zone of influence (ZOI) for thermoset cable, as determined by NUREG-1805 correlations. For distances outside the thermoset flame radiation ZOI, a study using Fire Dynamics Simulator (FDS) indicates that the steel housing of temperature sensitive equipment is effective in reducing damaging heat fluxes, and preventing damage to the equipment internals. This study is discussed at length in Report R2168-1 003B-0001, Verification and Validation of Fire Modeling Tools and Approaches for Use in NFPA 805 and Fire PRA Applications, Appendix C, which documents the FAQ 13-0004 response. Therefore, treatment is consistent with the guidance in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," and Fire PRA FAQ 13-0004.

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Plant walkdowns were executed in fire compartments where detailed fire modeling was performed to identify the locations of sensitive electronic components within PRA-credited cabinets. The results of the walkdowns indicate that some cabinets do not align with the results of FAQ 13-0004.

Additional analyses justify that there are no impacts to the Fire PRA and that the current fire modeling results are bounding. Instances outside the guidance of FAQ 13-0004 are justified as follows:

Fire Cabinet Component Type Screening Justification Compartment ID D-01 D-02 Exposed electronics are test devices or 305 D-07 electronic readouts of meters or other Exposed electronics monitoring devices. Although part of a PRA on cabinet face D-08 credited cabinet, these electronics are not D-09 critical to cabinet functionality.

676 C296*

1DY-01 1DY-02 Cabinet doors do not provide the protection 2DY-01 described in FAQ 13-0004, but there are no Partially grated 318 2DY-02 additional impacts to existing fire scenarios.

cabinet door DY-OA The current fire modeling bounds the failure of this component (fails at time = 0).

DY-OB Cabinet door does not provide the protection described in FAQ 13-0004, but Glass or Plexiglas there are no additional impacts to existing 333GRP 2C-290 cabinet door fire scenarios. The current fire modeling bounds the failure of this component (fails at time= 0).

• Not a PRA-credited cabinet, but has terminations for PRA-credited cables.

With respect to louvers or other typical ventilation means, the model outlined in Report R2168-1003B-0001, Appendix C, conservatively accounts for only natural ventilation. The cabinet in the model uses openings on the top and bottom of the housing to allow natural air flow through the cabinet. Provided the hot gas layer has not descended to the cabinet level, any other ventilation means would result in increased ambient air flow through the cabinet, reducing the internal cabinet temperature. Therefore, the natural ventilation in the model is conservative and bounds the use of louvers or other ventilation means.

g. New fire-specific human failure events (HFEs) have been evaluated as follows:

i. Feasibility assessment of credited HFEs was determined by review of the HFEs by two SROs licensed at Point Beach.

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ii. Existing procedures were used with steps inserted at the appropriate points as determined by engineering and operations. Timing was determined by two SROs with knowledge of the plant and existing procedures.

iii. The plant response modeling in the FPRA associated with credited HFEs was reviewed by two licensed SROs and a PRA engineer.

The items listed above will be verified upon completion of the modifications and procedural changes as per LAR, Table S-3, "Implementation Items", IMP-142 (Reference 1), as modified by the response to PRA RAI 21 by letter dated July 29, 2014 (Reference 6).

h.

i. The analysis of spurious indications was extensively revised and updated subsequent to the fire PRA peer review to provide a traceable basis for disposition of each control room annunciator or instrumentation used in EOPs applicable after a fire, rather than to simply list which procedures were reviewed. The analysis was conducted by an engineer knowledgeable in both PRA and PWR plant operations (previously held a SRO on a Westinghouse PWR). Any instrument which could result in a spurious annunciator or which is used to direct EOP actions following a fire was reviewed and individually dispositioned based on several generic considerations:

1. If the alarm response procedure included a step to verify the indication prior to taking any action, the item was screened as having negligible risk and no further analysis was required.

2. If the alarm would require multiple spurious indications on redundant channels, the item was screened as having negligible risk and no further analysis was required.

3. If the associated sse which could be adversely impacted by the

• undesired operator action was not credited in the fire PRA and would not adversely affect other SSCs, the item was screened as having negligible risk and no further analysis was required.

4. Spare instruments were screened as having negligible risk and no further analysis was required.

Any spurious indications not screened were then further reviewed to determine if another disposition was applicable, and if so this was documented. For example, such dispositions might identify that the directed action would not apply once the reactor was tripped, or the action was only to locally monitor equipment and not take any adverse action, or the alarm might be obviously spurious (e.g., manual Sl actuated and operator would know this was not done). Each potential spurious indication was dispositioned; therefore, there are no indications which could lead to an undesired operator action which would adversely affect the ability to safely shut down after a fire.

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ii) The scope of the evaluation included all control room annunciator associated instrumentation and the EOPs used for a post-fire response.

This includes each individual annunciator window and each individual EOP step. If an instrument neither causes any alarm nor is specifically used in the EOPs, then it was judged that there was no potential for the instrument to cause an undesired operator action which adversely impacts the ability to safely shut down after a fire.

i) The human reliability dependency analysis was updated in support of the PRA integrated analysis performed for the PRA RAI 3(a) response. This updated human reliability dependency analysis performed in the "NFPA 805 Fire PRA Quantification Notebook" (P2091-2900-02, Revision 2) uses a minimum joint human error probability (HEP) "floor" of 1E-5. Therefore, no justification is required for HEPs below this minimum value.

PRA RAI 03 - Integrated Analysis NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4. 1 further states that the change in public health risk arising from transition from the current FPP to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to 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. The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates.

The PRA methods listed below have not been accepted by the NRC staff. Unless a method is eventually found to be acceptable by the NRC, that method needs to be replaced by an acceptable method. Alternatively it may be demonstrated that the FPRA results used to support transition do not exceed the change in risk acceptance guidelines if the acceptable method were used. The PRA methods currently under review in the LAR include the following:

• PRA RAJ 1. a regarding removal of the Option #2 approach in assessment of circuit failure probabilities

• PRA RA/1.c regarding the inclusion of SOKC for internal and fire event related factors

• PRA RA/1.d and PRA RA/12 regarding conservative modeling of Instrument Air

• PRA RA/1.e regarding inclusion of junction boxes as damaging ignition sources

• PRA RAJ 1.f regarding treatment of sensitive electronics

• PRA RAJ 4 regarding removal of CPT credit in assessment of circuit failure probabilities

• PRA RAJ 5 regarding Heat Release Rates lower than 317 kW for transient sources

• PRA RAJ 6 regarding other disclosed deviations from acceptable PRA methods

• PRA RAJ 9 regarding MCB fire ignition frequency

• PRA RAJ 10 regarding main control room (MCR) abandonment modeling Page 9 of 48

• PRA 12 on assumptions due to cable routing and the impact on I::!.CDF, I::!.LERF

• PRA RAJ 13 regarding credit taken for the new RCP shutdown seals

• PRA RAJ 15 regarding fire damage effects from the opposite unit Please provide the following:

a) Results of an aggregate analysis that provides the integrated impact on the fire risk (i.e., the total transition CDF, LERF, !:::.CDF, !:::.LERF) of replacing specific methods identified above with alternative methods which are acceptable to the NRC. In this aggregate analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed. For those cases where no synergy exists, a one-at-a-time analysis may be done. For those cases that have a negligible impact, a qualitative evaluation may be done. It should be noted that this Jist may expand depending on NRC's review of other RA/s in this document.

b) Explain how the FPRA model will be updated to incorporate acceptable methods before using the PRA to support self-approval. While analyses may show that methods addressed one-at-a-time have negligible impact on the change on risk for the post-transition plant, these methods may have a greater impact in future plant-change evaluations (PCEs) since post transition self-approval acceptance guidelines are smaller. Therefore these methods need to be replaced with acceptable methods.

NextEra Response a) "Integrated Analysis of Point Beach Nuclear Plant Fire PRA and NFPA 805 Compliant Model" (P2428-0009-01, Rev. 0) was performed to evaluate the integrated impact of the changes indicated in Table PRA RAI 3.a-1 below. The analysis demonstrates that the fire probabilistic risk assessment model (FPRA) results meet the risk acceptance guidelines of Regulatory Guide (RG) 1.174.

The results are provided in an updated Attachment W to Reference 1. A discussion of the individual methods referenced in the RAI is provided below and how the method/topic was addressed in the integrated analysis and reflected in the updated LAR Attachment W (Attachment 7). Full dispositions can be found with the individual RAI responses.

Table PRA RAI 3.a-1 PRA RAI TOPIC Discussion ID Removal of the The FPRA has been updated using only Option #1. The Option #2 Option #2 approach is not used in assessment of circuit failure probabilities in approach in the updated FPRA model.

1.a assessment of circuit failure probabilities.

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Inclusion of state The FPRA has been updated to include the uncertainty parameters of knowledge for ignition frequencies, non-suppression probabilities, and hot short correlation induced spurious operations. SOKC is addressed using EPRI 1.c (SOKC) for UNCERT software, which provides a mean core damage frequency internal and fire (CDF)/Iarge early release frequency (LERF) with uncertainty.

event related factors.

Conservative To avoid conservative modeling of instrument air, it was not credited modeling of in the Variant Model, except in cases where the assumption that air Instrument Air. was failed provided a non-conservative input to the model.

1.d Instrument air was credited in the Compliant Model, except when instrument air was failed as a result of the fire. This prevented underestimatin_g delta CDF or delta LERF.

Inclusion of Junction boxes have been included in the updated Fire PRA as junction boxes ignition sources using an approach consistent with FPRA FAQ 13-1.e as damaging 0006, "Modeling Junction Box Scenarios in a Fire PRA".

ignition sources.

Treatment of Plant walkdowns were performed in the fire compartments to sensitive identify those cabinets potentially containing PRA credited sensitive electronics electronics. This identification is consistent with the guidance in NUREG/CR-6850 and Fire PRA FAQ 13-0004, "Clarifications on Treatment of Sensitive Electronics."

1.f Detailed fire modeling walkdowns indicate that some cabinets do not align with the results of FAQ 13-0004. Any conflicts with the guidance of FAQ 13-0004 have been justified. There were no Fire PRA model changes required.

Removal of CPT The FPRA has been updated without taking control power credit in transformer (CPT) credit in the circuit failure mode likelihood 4 assessment of analysis (CFMLA, Task 10 of NUREG/CR-6850).

circuit failure probabilities.

Heat Release Two areas use heat release rates lower than 317 kW for transient Rates lower than sources. Based on the enhanced administrative controls that will be 317 kW for put in place for the transition to NFPA 805, the limited personnel transient traffic expected in these areas, and minimal combustibles required 5

sources. for maintenance activities in these areas, the reduced heat release rates were determined to be appropriate to represent the expected combustibles in these areas. There were no Fire PRA model changes required.

Other disclosed At the time of the LAR submittal, the statement provided in LAR deviations from Attachment V,Section V.2, "PBNP fire PRA did not use unreviewed acceptable PRA analysis methods," was accurate.

6 methods.

The FPRA has been updated, incorporating several methods acceptable to the NRC Staff provided since the LAR submittal.

These are also addressed in other RAis.

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Main control The MCB fire ignition frequency for non-abandonment cases has board (MCB) fire been updated consistent with Appendix Lin NUREG/CR-6850 9

ignition guidance. No partitioning or segmentation is used for subdividing frequency. the MCB fire frequency.

Main control For the post-transition model, main control room (MCR) room (MCR) abandonment is not being credited for non-habitability cases. In the 10 abandonment compliant case, MCR abandonment is credited in all scenarios.

modeling.

Assumptions "Credit by exception" is not used in the Point Beach Fire PRA.

due to cable Component events that do not support fire mitigating strategies are 12 routing and the failed for the Fire PRA. The failures have been assessed and no impact on LlCDF, significant over-estimation of risk is expected. There were no Fire LlLERF. PRA model changes made.

Credit taken for The FPRA has been updated based on the most recent qualification the new RCP testing of the redesigned Westinghouse SHIELD Passive Thermal shutdown seals. Shutdown Seal (SDS) (Generation Ill), which has been documented in WEC report is PWROG-14001-P/NP, "PRA Model for the Generation Ill Westinghouse Shutdown Seal, PA-RMSC-0499R2."

13 The WEC report has been submitted to the NRC, but has not yet been approved.

If the report is not approved by the NRC or requires revision, the PRA used in support of the NFPA 805 PRA will be adjusted as necessary.

Fire damage Each postulated fire in the plant is evaluated to determine the scope effects from the of fire damage, either from direct damage to the equipment or opposite unit. damage to cables powering and/or controlling the equipment. With the exception of fires originating in the unit-specific containment buildings, each postulated fire is evaluated twice, once for each unit.

15 This evaluation is performed regardless of whether the ignition source is from a Unit 1 component, a Unit 2 component, or a component that is common to both units. As such, the reported CDF, .LlCDF, LERF, and .LlLERF values for one unit implicitly include any equipment lost due to a fire originating on the opposite unit.

There were no Fire PRA model changes made.

b) The Fire Probability Risk Assessment (FPRA) model was updated to incorporate changes as described in the response to Request for Additional Information (RAI) Probabilistic Risk Assessment (PRA) 3a to support the integrated analysis reflected in the updated License Amendment Request (LAR) Attachment 7. Following incorporation of these changes, the FPRA does not have any unacceptable methods, and is, therefore, considered to be acceptable for use in post-transition plant change evaluations.

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PRA RAI 04- Control Power Transformer (CPT) Credit for Circuit Failure Probabilities NFPA 805 Section 2.4.3.3 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 a fire protection program consistent with NFPA-805. In Jetter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC Staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC Staff to complete its review of the proposed method.

The analysis regarding circuit failure mode likelihood appears to credit CPTs for a reduction factor of two. The NRC staff concludes that the effect of any CPT reduction to the hot short-induced spurious operation likelihood cannot be substantiated. The staff cannot complete its review based on the current analysis. Replace this analysis and provide an explanation of the method used and the results in sufficient detail so the staff can make a conclusion regarding the use of the method.

NextEra Response The circuit failure mode likelihood analysis (CFMLA, Task 10 of NUREG/CR-6850) in the FPRA was reevaluated using only Option #1 without control power transformer (CPT) credit. The results of this update are evaluated and discussed in the integrated analysis as part of the response to PRA RAI 3.

PRA RAI 09 - Main Control Board Fire Ignition Frequency NFPA 805 Section 2.4.3.3 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 a FPP consistent with NFPA-805. 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.

The licensee's analysis indicates that the total adjusted fire ignition frequency is 1.65E-03/year for the two units. In addition, the licensee's analysis explains that the MCB fires are divided into eleven scenarios, three of which are screened from quantification. The licensee's analysis indicates that the fire ignition frequencies applied to MCB fires is 1.50E-4/year (or the total adjusted fire ignition frequency of 1.65E-03/year divided by 11). Dividing the MCB frequency in this manner is different than NRC guidance.

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When applying the NUREG/CR-6850 Appendix L method, the frequency of a scenario involving specific target damage in the MCB should be determined by multiplying the probability of target damage, such as defined by Figure L-1 of NUREG/CR-6850, by the entire MCB frequency. Partitions or segmentation cannot be used to justify subdividing the MCB fire frequency, unless accompanied by a recalculation of Appendix L, Figure L-

1. Full partitions without openings or gaps might be used to preclude scenarios involving targets sets which extend across partitions.

The NRC staff cannot complete its review based on the current analysis. Replace this analysis and provide an explanation of the approach used and the results in sufficient detail so the staff can make a conclusion regarding the use of the approach.

NextEra Response The MCB fire ignition frequency for non-abandonment cases has been updated from 1.50E-4/year to the full MCB ignition frequency of 1.65E-3/yr as per Appendix L in NUREG/CR-6850 guidance. No partitioning or segmentation is used for subdividing the MCB fire frequency. The changes are documented in "NFPA 805 Fire PRA Main Control Room Analysis," P2091-2700-01, Revision 3. The updated MCB fire frequency will be included in the integrated analysis provided in the response to PRA RAI 03.

PRA RAI 16

• Calculation of VFDR LlCDF and LlLERF NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4. 1 further 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 NRC. RG 1. 174 provides quantitative guidelines on core damage frequency, large early release frequency, 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.

The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

LAR Attachment W, Section W. 2. 1 provides a high-level description of how the b..CDF and b..LERF for the VFDRs and the additional risk of recovery actions for each of the fire areas were determined, which does not provide enough detail to make the approach completely understood. Provide further description of the methods used to determine the change in risk values reported in LAR Attachment W, Tables W-6 and W-7 that addresses the following:

a) A detailed definition of both the post-transition and compliant plant models used to calculate the reported changes in risk and additional risk of recovery actions, including any special calculations for the MCR. (It is recognized that PRA RAJ 10 already asks questions about MCR abandonment, but this question subpart is different in that it focuses on how delta CDF and LERF are calculated). This discussion should include explanation of how VFDR and non-VFDRs modifications are addressed for both the post-transition and compliant case; Page 14 of 48

b) A description of how the reported changes in risk and additional risk of recovery actions were calculated, including any special calculations performed for the MCR. Also, include a description of PRA modeling logic and mechanisms, such as adding events or logic, and using surrogate events; c) A clarification of whether FAQ 08-0054 (ADAMS Accession No.

ML110140183) guidance was applied, and identification of the use of PRA modeling, data, or methods, added after the FPRA peer review; and, d) A description of the type of VFDRs identified, and a discussion of whether the VFDRs identified but not modeled in the FPRA impact the risk estimates. (e.g.,

LAR Attachment C, Table C-1 states that VFDRs associated with overcurrent trip concerns were not modeled in the Fire PRA.)

NextEra Response a) The post-transition model (also referred to as the variant model) is developed using the current plant configuration with additional procedure changes and modifications identified in Attachment S of the Point Beach Transition to 10 CFR 50.48(c) LAR submittal (Reference 1). The compliant model is developed starting with the post-transition model and is updated with cable protections and equipment model flags to provide the ability to block the failure of cable and equipment variances from deterministic requirements (VFDRs) on a location area basis. Risk reduction modifications were built into the model with the ability to flag them out or remove cable protections. The compliant model uses a different cable protection list and modified flag file to remove risk reduction modifications.

Modifications identified in the "Risk Reduction Modifications" section of Attachment W are applied to the post-transition model and are not applied to the compliant model. The risk reduction modifications are removed from the compliant case because they do not directly relate to resolution of any VFDRs.

Modifications that address VFDRs identified in Attachment S are applied to both the post-transition and compliant model.

The instrument air system availability was addressed differently in the post-transition and compliant models. Instrument Air was not credited in the post-transition (variant) model, except in cases where the assumption that air was failed provided a non-conservative input to the model. Instrument Air was credited in the Compliant Model, except when Instrument Air was failed as a result of the fire. This prevented underestimating delta CDF or delta LERF.

The response to PRA RAI 10 has a detailed discussion on the calculation of MCR abandonment caused by loss of control (non-habitability cases). The following table describes the application of main control room abandonment in the post-transition and compliant models.

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Abandonment Post-Transition Model Compliant Model Aspect Habitability CCDP of 0.565 for Unit 1 and Assume human actions are successful and 0.566 for Unit 2. One order of use hardware random failures as a CCDP.

magnitude lower for LERF. CCDP of 0.19 (Attachment W section W .2.1 (See RAI 10 development) of the LAR) for Units 1 and 2. One order of magnitude lower for LERF.

Loss of Control No Credit Given Assume human actions are successful and use hardware random failures as a CCDP ceiling in abandonment areas. CCDP of 0.19 (Attachment W section W .2.1 of the LAR) for Units 1 and 2. One order lower for LERF. An example of application would be if a specific fire scenario in an abandonment area had a CCDP of 0. 75 this CCDP would be replaced with the 0.19 ceiling, because it is assumed the operators have perfect judgment to initiate abandonment, when it will reduce plant risk.

b) The reported changes in risk were calculated based on determining the differences between the variant (also referred to as post-transition) and compliant model quantification results. A description of the PRA modeling logic and mechanisms supporting the compliant model are provided in the response to PRA RAI 16(a).

Additional risk of recovery actions were calculated by setting recovery events failure probabilities to zero in the post-transition model to obtain the differences in risk for each fire area. Main control room abandonment is only considered for loss of habitability in the main control room for the post-transition model.

The risk of recovery actions related to abandonment were assessed as described in the following table:

Abandonment Aspect Post-Transition Model Changes to Obtain Risk of Recovel)l Actions Risk of Recovery Actions CCDP of 0.565 for Unit 1 and Assume human actions are Habitability 0.566 for Unit 2. One order successful and use hardware lower for LERF. (See PRA random failures as a CCDP.

RAI 10 development) CCDP of 0.19 (Attachment W section W.2.1 of the LAR) for Unit 1 and 2. One order lower for LERF.

Risk of Recovery Actions No Credit Given No Credit Given Loss of Control c) FAQ 08-0054, "Demonstrating Compliance with Chapter 4 of NFPA 805,"

guidance was used to develop the Fire Risk Evaluations (FREs) and LAR 271 (Reference 1). No new PRA methods have been introduced to the FPRA after the focused scope peer review.

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d) In general, variances from deterministic requirements (VFDRs) that impact plant monitoring instrumentation are not addressed in the fire probabilistic risk assessment model (PRA), unless the instrumentation affected is also a cue for a human failure event (HFE). In the case where the instrumentation has no impact on any HFE, loss of the capability to monitor the affected parameter would not affect any plant actions necessary to achieve safe shutdown conditions following a fire, based on the realistic assumptions used for the PRA. Loss of numerous instruments as a result of fire, where the instrumentation is not used as a cue for any HFE, is not judged to be risk significant. Therefore, there would be no adverse impact on either core damage frequency (CDF) or large early release frequency (LERF), and the risk impact is zero. (Note that such instruments are also evaluated for the potential to cause an undesired operator action, if they are used in post-fire EOPs or if they cause a control room alarm.)

The loss of the ability to monitor any required parameters would still be assessed with regards to defense-in-depth and safety margins to complete the risk-informed process to disposition the variant condition, consistent with the provisions of NFPA 805. As stated previously, the Fire PRA risk assessment would not model some VFDR related instrumentation as they are not contributors to core damage or large early release sequences.

The following types of VFDRs are not modeled in the PRA:

• Control Building/Primary Auxiliary Building (PAB) heating, ventilation and air conditioning (HVAC) failure on the 8' elevation, as well as Cable Spreading Room ventilation systems - Loss of ventilation after plant trip will have negligible impact on risk based on the internal events analyses.

• Control failure of pressurizer heaters - Pressurizer heaters were evaluated as having a negligible risk impact during the Multiple Spurious Operations Review and therefore negligible impact on risk.

• Steam generator sample isolation valves -These lines were evaluated as having a negligible flow diversion during the Multiple Spurious Operations Review and therefore negligible impact on risk.

• Neutron monitoring- See instrumentation discussion above.

• Temperature indication on reactor coolant system (RCS) hot/cold legs-See instrumentation discussion above.

• Condensate storage tank (CST) level indication- Based on the small volume of the CST, the operators normally would be cued in about one hour and are expected to transfer the auxiliary feedwater (AFW) pumps to another source of water with approximately two hours required to complete the action. When CST indication is unavailable, the operators would be expected to transfer suction earlier, instead of proceeding with an uncertain plant status with the limited time available. The change in cue availability on this longer term expected Human Failure Event has negligible impact on risk and the cognitive portion has already been adjusted for fire events.

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• Spurious action of the pressurizer spray- Similar to pressurizer heaters, pressurizer spray is primarily needed to maintain normal operational parameters and is not critical to preventing core damage following a plant transient. This has a negligible impact on risk.

• Steam generator blow-down isolation valve - These lines were evaluated as having a negligible flow diversion during the Multiple Spurious Operations Review and therefore negligible impact on risk.

• Reactor coolant pump (RCP) seal leak-off line isolation valves - Closing of these valves does not fail RCP Seal Injection, and failure of these valves will not prevent cooling by injection. Therefore there is negligible impact on risk.

• Boric acid blender outlet flow control valve -This system is primarily needed for normal operations or emergency boration options. As such, it has negligible contribution to fire risk.

• Secondary fire and over-current trip (OCT) -The secondary fires and OCT impacts have been added to the fire PRA. The impact on risk is included in the integrated risk analysis as documented in the response to RAI PRA 03(a).

• Containment Recirculation Emergency flow control valve (FCV) -The valve is required to remain closed for the credited service water system line up for the support goal. Conversely, the PRA service water success criteria are evaluated with the valve fully open, therefore there is negligible impact on risk.

• Steam Generator Pressure Indication- The steam generator pressure indicator has a negligible impact on plant risk. As such, the risk, safety margin and defense-in-depth acceptance criteria are satisfied without further action.

PRA RAI 18 - Large Reduction Credit for Modifications NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4. 1 further 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 NRC. RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, 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.

The NRC staffs review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

Page 18 of 48

LAR Attachment W, Tables W-6 and W-7 report a total b. CDF of -3.95£-06/year for Unit 1 and -3.33E-4/year for Unit 2, respectively. LAR Attachment W, Section W2 identifies a set of risk reduction modifications credited in the FPRA that are unrelated to VFDRs, and explains that the risk reduction from these modifications results in higher risk for the compliant plant than for the variant plant risk in some instances. The large negative total

b. CDF reported in LAR Attachment W, Table W-7 for Unit 2 implies that the CDF for the compliant plant for Unit 2 is very high (-4E-4/year). Regarding risk reduction modifications that do not address VFDRs, Section 3.2.5 of RG 1.205, provides guidance that risk decreases may be combined with risk increases for the purposes of evaluating combined changes in accordance with regulatory positions presented in Sections 1.1 and 1. 2 of RG 1. 174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Rev. 2. Given that an overly conservative calculation of the compliant plant CDF and LERF can lead to a non-conservative calculation of the b. CDF and b. LERF, the contribution of conservative modeling of the compliant plant risk to the b. CDF and b. LERF should be evaluated in detail. In light of the above, provide the following:

a) Identification of the modifications in LAR AttachmentS, Table S-2 that are being implemented to remove VFDRs, opposed to those being implemented solely to reduce risk (i.e., non-VFDR modifications);

b) The total unit increase in the b. CDF and b. LERF from accepting the unresolved VFDRs and the total unit decrease in the b. CDF and b. LERF from implementing the non-VFDR related modifications. In these calculations the only variation between the post-transition and the compliant plant PRAs should be how they model the retained VFDRs or non- VFDR modifications; c) A summary of the risk significant scenarios for fire areas in the compliant case, including risk significant scenarios for Fire Area A23N reported in LAR Attachment W, Table W-7 as having a b. CDF of -3.08£-04/year for Unit 2; and, d) A discussion of the contribution of fire-induced failures to those scenarios and the impact of any assumptions made that significantly contribute to the variant and the compliant case risk.

NextEra Response a) The following table lists each of the modifications from the updated LAR AttachmentS, Table S-2 (see Attachment 5). Associated VFDR numbers are included, if applicable. Refer to RAI PRA 25, "Changes to Modifications Described in LAR Attachment S."

Table S-2 MOD Description Associated VFDR MODID EC 271218 Unsealed penetrations in the floor of the None EC 278396 cable spreading room to the auxiliary feedwater pump rooms will be provided with penetration seals.

Page 19 of 48

Table S-2 MOD Description Associated VFDR MODID EC 272527 Replace TDAFWPs with self-cooled models None EC 272529 LTAM The fire alarm system equipment credited in None PB-12-0038 Table 4-3 will be upgraded as necessary MOD-1 Bus duct between B-03 and B-04 buses will None be modified to prevent high-energy arcing faults MOD-2 Cross-tie TDAFWP steam supplies and None pump discharge(S) to allow opposite-Unit sup~ort MOD-3 RCP seals will be upgraded to Generation-Ill None Westinghouse Shutdown Seals MOD-4 Additional power inputs to B-08 and B-09; None power to come from 2A-06.

MOD-5 Provide redundant power supply to P-38s None from B-08, B-09 MOD-6 Restore N2 supply to primary PORVs and None verify supply is adequately sized to support PRA success criteria (24 hrs)

MOD-7 PZR PORV cables in FZ-187 will be None rerouted or protected to prevent spurious PORV operation: ZK11429A, ZP21429A MOD-10 Protect PORV instrument cables in FZ-318 A30-14 (U1)*

to prevent spurious PORV actuation. This A30-27 (U2)*

MOD does not fully remove the VFDRs.

MOD-11 Protect ZFD0406A in FZ-166 to preserve G- A15-16 03 field flash MOD-14 Provide alternate means to isolate the main None steam flow path with cabling and power supplies independent of Control Room or Cable Spreading Room MOD-19 Protect G-03, G-04 interlock cables in FZ- A24-01 305 MOD-20 Protect D-108 cables in A01-B A01-B-64 MOD-21 Protect 1AF-04002 cables in FZ-318 A30-01*

MOD-22 Protect 1AF-04000 cables in FZ-318 A30-01*

MOD-23 Reduce dependence on instrument air for None P-38 AOVs by providing accumulators with 24 hr pneumatic supply MOD-24 BKR coordination MODs None MOD-25 Cable Thermal Withstand MODs None MOD-26 BKR coordination/ protection MODs None MOD-27 Unsealed penetrations in floor of non-vital None switchgear room to vital switchgear room to be provided with penetration seals.

EC 272841 Charging Pump low suction pressure trip None EC 261021

• VFDR does not appear 1n Table S-2 because MOD does not fully resolve VFDR.

Page 20 of 48

b) Attachment W of the LAR (Attachment 7) has been updated to include the total unit decrease in the b. CDF (Core damage frequency) and b. LERF (large early release frequency) from implementing the non-VFDR (variance from deterministic requirements) related modifications.

The following two delta risk metrics were used to describe the transition risk and the risk decrease for non-VFDR modifications:

Vr-Cnr = b.CDF fLERFFRE= Provides the final transition risk Cr-Cnr = b.CDF fLERFRR Mod =total unit decrease in the 1:::. CDF and 1:::. LERF from implementing the non-VFDR related modifications.

Where, Vr = CDFILERF of variant case with risk modifications incorporated Cnr = CDFILERF of compliant case without modifications incorporated Cr = CDFILERF of compliant case with risk modifications incorporated The totals as presented in Attachment W (see Attachment 7) are:

Unit 1 Fire Risk Evaluations b.CDF /LERFFRE = -6.4E-06 I 4.5E-07 Unit 2 Fire Risk Evaluations b.CDF /LERFFRE = -2.9E-04 I -1.3E-07 Unit 1 Non-VFDR Modification b.CDF /LERFRR Mod= -3.4E-05 I -1.5E-07 Unit 2 Non-VFDR Modification b.CDF /LERF RR Mod = -3.3E-04 I -9.2E-07.

c) The following tables contain risk insights on the LAR scenarios for each unit that contribute

>1% to CDF or >1% LERF to the risk of the compliant models used for the Fire Risk Evaluations in the LAR submittal (Reference 1).

Fire Area A23N risk for Unit 2 compliant case is driven mostly by the bus duct fires (FC304N.BUSDUCT-S2-FO and -F1 ). The risk is reduced for the Unit 2 variant case for fire scenario FC304N.BUSDUCT-S2-FO because of a modification that replaces a bus duct with a cable, which eliminates the potential for a high energy arcing fault (HEAF) fire for the variant case. This modification is considered a risk reduction modification (not related to VFDRs). This risk reduction modification is credited in the variant (post-transition) case, but not the compliant case. Significant risk reduction modifications can result in the compliant case risk being greater than the variant case.

Outside of the bus duct modification, other significant risk reduction modifications ensure the availability of additional AFW trains that are credited in the variant case but not in the compliant case. Therefore, some fire areas have a compliant case risk greater than the variant case. That is, the compliant case may have a single train of AFW available, while the variant case may have two or more trains, greatly reducing risk and resulting in a negative delta risk.

Risk-insights related to assumptions from this review are provided in PRA RAI 18(d) response.

Page 21 of 48

Table 1 - Unit 1 CDF for Scenarios with >1% Risk Contribution Compliant Fire CDF Scenario ID Scenario Description Risk Insights CDF Area Contribution (per rx-yr)

This fire scenario causes a failure of busses 1B-03 and 1B-04, causing a loss of motor-ELECTRICAL FIRE driven AFW pumps P-38A, P-38B, and 1P-FC318.1X14-F1 A30 CAB 1X14 10.3% 53. Random failure of the turbine-driven 6.64E-06 (TARGETS) AFW pumps results in a loss of all AFW, and Sl is unavailable in recirculation mode due to power failures.

This fire scenario causes a loss of offsite power and failure of diesel generators G-01 HEAF FIRE and G-02, along with the turbine-driven AFW FC304S.BUSDUCT-A23S BUSDUCT S1 8.7% pump. Random failures of the remaining 5.61 E-06 S1-FO (TARGET) diesel generators results in loss of all AC power, which fails AFW and bleed-and-feed capability.

This fire scenario results in a loss of AFW pumps P-38A and B, and steam supplies to the turbine-driven pump. The fire location ELECTRICAL FIRE blocks local recovery of the turbine-driven A01-FC237.2B31-F2 CAB 2B-31 (WHOLE 5.7% pump, and the motor-driven pump 1P-53 is 3.67E-06 B/46 ROOM) randomly failed. Bleed-and-feed is failed due to fire-induced failures of RWST level indication to support transition to ECCS

---

-

recirculation.

Page 22 of 48

Table 1 - Unit 1 CDF for Scenarios with >1% Risk Contribution Compliant Fire CDF Scenario ID Scenario Description Risk Insights CDF Area Contribution (per rx-yr)

This fire scenario causes a loss of offsite power and failure of diesel generators G-01, G-02, and G-04. Random failure of the G-03 diesel generator and turbine-driven AFW pump results in loss of all AC power and ELECTRICAL FIRE decay heat removal. In other cases where FC305.D-08-F1 A24 5.2% 3.35E-06 CAB D-08(ASD) 1 the G-03 diesel generator is available, battery depletion requires operator actions to align backup power supplies or human error results in loss of AFW and pressurizer PORV control for bleed-and-feed decay heat removal.

This fire scenario directly fails the P-38A and B AFW pumps, along with the steam supply to the 1P-29 turbine-driven AFW pump. The location of the fire prevents local recovery of ELECTRICAL FIRE A01- the steam supply valve. A random failure of FC237.1B31-F2 CAB 1B-31 (WHOLE 5.0% 3.18E-06 B/46 the 1P-53 motor-driven AFW pump results in I ROOM) loss of all AFW. Bleed-and-feed is failed due to fire-induced failures RWST level indication which supports transition to ECCS recirculation.

This fire scenario directly fails the P-38A and B AFW pumps and the 1P-29 turbine-driven HEAF FIRE AFW pump. A random failure of the 1P-53 FC304S.BUSDUCT-A23S BUSDUCT S2 4.2% motor-driven AFW pump results in loss of all 2.69E-06 S2-FO (TARGETS) AFW, and the fire also impacts the capability of the Sl system in recirculation mode, failing bleed-and-feed decay heat removal.

Page 23 of 48

Table 1 - Unit 1 CDF for Scenarios with >1% Risk Contribution Compliant Fire CDF Scenario ID Scenario Description Risk Insights CDF Area Contribution (per rx-yr)

This is a dual-unit scenario with control room abandonment based on habitability and Main Control Room A31-ABAND A31 3.5% assumed failures of the 1P-29 Turbine 2.26E-06 Abandonment Driven AFW pump and the Gas Turbine Generator G05.

This fire scenario causes a partial loss of offsite power and failure of diesel generator G-01, G-02, and G-04. AFW automatic start is disabled by DC bus failures and could be ELECTRICAL FIRE FC305.1 A05-62 recovered by operator manually starting A24 CAB 1A-05 (62-66) 3.5% 2.24E-06 F1 AFW pumps, but this human action fails.

(ASD)

Bleed-and-feed cooling with the Sl pumps is similarly available but also requires manual initiation and shares similar cues with the manual actions to start AFW.

This fire scenario fails AFW pumps P-38A, P-38B, and 1P-53. Random failure of the 1P-29 turbine-driven AFW pump results in loss FC318.1 B04-G20- HEAF FIRE CAB 1B- of all AFW. The fire also impacts several A30 3.1% 2.02E-06 24H-F1 04 (20-24) battery chargers, which results in inability to align Sl for recirculation to support continued decay heat removal for bleed-and-feed cooling.

This fire scenario fails AFW pumps P-38A, P-38B, and 1P-53. Random failure of the 1P-29 turbine-driven AFW pump results in loss FC318.1 B04-G17L- HEAF FIRE CAB 1B- of all AFW. The fire also impacts several A30 3.1% 2.02E-06 19H-F1 04 (17L-19) battery chargers, which results in inability to align Sl for recirculation to support continued decay heat removal for bleed-and-feed cooling.

Page 24 of 48

Table 1 - Unit 1 CDF for Scenarios with >1% Risk Contribution Fire CDF Compliant Scenario 10 Scenario Description Risk Insights CDF Area Contribution (per rx-vr)

This fire scenario causes a loss of offsite power and failure of diesel generators G-01 and G-02, along with DC battery chargers D-07 and D-108. All AFW pumps are failed ELECTRICAL FIRE FC305.1 A03-38 when DC control power is interrupted by A24 CAB 1A-03 (38-40) 1.6% 1.03E-06 F1 battery depletion and inability to align to (ASD) backup sources, either due to random equipment failures or human errors. Bleed-and-feed cooling is also impacted by loss of control power.

This fire scenario results in a loss of AFW pumps P-38A and B, and the 1P-53 motor-ELECTRICAL FIRE driven pump. Turbine-driven pump, 1P-29, is FC333GRP.2C-170-A32 CAB 2C-290 1.6% randomly failed. Bleed-and-feed is failed due 1.01 E-06 F1 (SOURCE) to fire-induced failures RWST level indication which supports transition to ECCS recirculation.

This fire scenario causes a loss of five out of six battery chargers and AFW pumps P-38B and 1P-53. The loss of DC control power after battery depletion fails AFW pump P-ELECTRICAL FIRE FC318. D13-F1 A30 1.5% 38A, and a random failure of the 1P-29 9.46E-07 CAB D13 (AD) turbine-driven AFW pump results in loss of all AFW. The pressurizer PORVs also lose control power and cannot be opened for bleed-and-feed coolina.

This fire scenario causes a loss of offsite power and failure of all diesel generators.

ELECTRICAL FIRE Random failure of the 1P-29 turbine-driven FC318.DYOA-F1 A30 1.4% 8.95E-07 CAB DYOA (AD) AFW pump results in loss of all decay heat removal, since bleed-and-feed cooling is failed by the loss of all AC power.

Page 25 of 48

Table 1 - Unit 1 CDF for Scenarios with >1% Risk Contribution Complian~

Fire GDF Scenario ID Scenario Description Risk Insights CDF Area Contribution (per rx-yr)

This fire scenario causes a loss of offsite power and failure of diesel generators G-01, G-03, and G-04, along with several battery chargers. Random failure of the G-02 diesel generator and the 1P-29 turbine-driven AFW ELECTRICAL FIRE pump results in loss of all AC power and FC318.D14-F1 A30 1.4% 8.67E-07 CAB D14 (AD) decay heat removal. In other cases where the G-02 diesel generator is available, battery depletion results in the loss of the P-38A AFW pump. The pressurizer PORVs lose control power and cannot be opened to support bleed-and-feed cooling.

The majority (80%) of the risk is due to fire-induced LOCAs due to un-isolable letdown flow, and the unavailability of ECCS recirculation due to fire-induced valve TRANSIENT FIRE failures. The remaining (20%) of the risk is TS MONITOR due to fire-induced failure of both P-38 A and FC187GRP.TS4-F1 A01-B 1.3% 8.16E-07 TANK ROOM B AFW pumps and the 1P-29 turbine-driven (TARGETS) AFW pump, and a random failure of the motor-driven 1P-53 AFW pump, causing a loss of all AFW, and the inability to align for ECCS recirculation to maintain bleed-and-feed cooling. --

Page 26 of 48

Table 1 - Unit 1 CDF for Scenarios with >1% Risk Contribution Compliant Fire CDF Scenario ID Scenario Description Risk Insights CDF Area Contribution (per rx-vr)

This fire scenario causes a loss of offsite power and failure of diesel generators G-01 and G-02, along with failure of four of six battery chargers. Failure of a backup battery charger due to random failure or human error results in a loss of DC control power which ELECTRICAL FIRE FC305.2A03-44 fails the motor-driven 1P-53 AFW pump and A24 CAB 2A-03 (44-46) 1.1% 6.99E-07 F1 both P-38A and B pumps. Random failure of (ASD) the fire water system fails the long term water supply to the 1P-29 turbine-driven AFW pump, causing a loss of all AFW. Train A Sl is directly fire-failed, and train B Sl fails on loss of DC control power, which fails bleed-and-feed coolinQ.

This fire scenario causes a loss of offsite power and failure of diesel generators G-01 and G-02, along with failure of four of six battery chargers. Failure of a backup battery charger due to random failure or human error results in a loss of DC control power which ELECTRICAL FIRE FC305.2A03-41 fails the motor-driven 1P-53 AFW pump and A24 CAB 2A-03 (41-43) 1.1% 6.85E-07 F1 both P-38A and B pumps. Random failure of (ASD) the fire water system fails the long term water supply to the 1P-29 turbine-driven AFW pump, causing a loss of all AFW. Train A Sl is directly fire-failed, and train B Sl fails on loss of DC control power, which fails bleed-and-feed cooling.

Page 27 of 48

Table 1- Unit 1 CDF for Scenarios with >1% Risk Contribution Compliant Fire CDF Scenario ID Scenario Description Risk Insights CDF Area Contribution (per rx-yr)

This fire scenario directly fails P-38A and motor-driven 1P-53 AFW pumps. Random TRANSFORMER failures result in only the P-38B pump being FC333GRP.1XY A32 FIRE 1XY-07 1.0% available which cannot support both units, so 6.65E-07 FO (WHOLEROOM) AFW is unavailable. Bleed-and-feed cooling is fire failed due to loss of RWST level instrumentation.

This fire scenario directly fails P-38A and motor-driven 1P-53 AFW pumps. Random TRANSFORMER failures result in only the P-38B pump being FC333GRP.2XY A32 FIRE 2XY-07 1.0% available which cannot support both units, so 6.65E-07 FO (WHOLEROOM) AFW is unavailable. Bleed-and-feed cooling is fire failed due to loss of RWST level instrumentation.

Note 1: (ASD) in the scenario description denotes Automatic Suppression and Detection fire modeling cases Page 28 of 48

Table 2 - Unit 1 LERF for Scenarios with >1% Risk Contribution Compliant Fire LERF Scenario ID Scenario Description Risk Insights LERF Area Contribution (per rx-yr)

This is a dual-unit scenario with control room abandonment based on habitability and Main Control Room A31-ABAND A31 44.8% assumed failures of the 1P-29 Turbine Driven 2.26E-07 Abandonment AFW pump and the Gas Turbine Generator G05.

This fire scenario results in a station blackout results in a loss of all AFW and Safety HEAF FIRE FC304S.BUSDUCT- Injection. The dominant containment failure A23S BUSDUCT S1 7.8% 3.93E-08 S1-FO modes are early failure with vessel at high (TARGET) pressure (73% ), and failure due to pre-existing leakage (12%).

This fire scenario results in a fire-induced small or very small LOCA (80%) or a loss of TRANSIENT FIRE all AFW (20% ). The dominant containment TS MONITOR FC187GRP.TS4-F1 A01-B 6.9% failure modes are associated with a fire- 3.47E-08 TANK ROOM induced containment isolation valve failure (TARGETS)

(containment penetration #9) (73%) and early failure due to vessel breach (22% ).

This fire scenario results in either station blackout or a loss of all AFW with no bleed-and-feed cooling due to pressurizer PORVs ELECTRICAL FIRE FC305. D-08-F1 A24 3.4% failing closed. The dominant containment 1.71 E-08 CAB D-08(ASD) 1 failure modes are early failure with vessel at high pressure (69%), and failure due to pre-existing leakage (16%).

Page 29 of 48

Table 2 - Unit 1 LERF for Scenarios with >1% Risk Contribution Compliant Fire LERF Scenario ID Scenario Description Risk Insights LERF Area Contribution (per rx-vr)

This fire scenario results in a loss of all AFW with no bleed-and-feed cooling due to pressurizer PORVs failing closed. The ELECTRICAL FIRE FC187GRP.C-181- dominant containment failure modes are A01-B CAB C-181 2.7% 1.38E-08 F1 associated with a fire-induced containment (TARGETS) isolation valve failure (containment penetration #9) (85%) and early failure with vessel at hiqh pressure (11 %).

This fire scenario results in a PORV LOCA or a loss of all AFW with no bleed-and-feed due to fire-induced failure of Sl valves. The ELECTRICAL FIRE FC187GRP.1 B dominant containment failure modes are A01-B CAB 1B-42 2.2% 1.11 E-08 F1 associated with a fire-induced containment (TARGET) isolation valve failure (containment penetration #9) (76%) and early failure with vessel at hiqh pressure (20%).

This fire scenario results in a loss of all AFW and failure of bleed-and-feed at recirculation.

ELECTRICAL FIRE The dominant containment failure modes are FC318.1X14-F1 A30 CAB 1X14 1.9% 9.80E-09 associated with a pre-existing leak (53%) and (TARGETS) early failure with vessel at high pressure (36%).

This fire scenario results in loss of all AFW and Sl due to failure of DC control power, ELECTRICAL FIRE caused by fire effects and random or human FC305.1 A03-38 A24 CAB 1A-03 (38-40) 1.4% failures. The dominant containment failure 6.83E-09 F1 (ASD) modes are early failure with vessel at high pressure (75%), and failure due to pre-existinq leakaqe (11 %).

Page 30 of 48

Table 2

• Unit 1 LERF for Scenarios with >1% Risk Contribution Compliant Fire LERF Scenario ID Scenario Description Risk Insights LERF Area Contribution (per rx--yrj_

This fire scenario results in a loss of all AFW and failure of bleed-and-feed cooling at I

FC318.1 B04-G20- HEAF FIRE CAB 1B- recirculation. The dominant containment A30 1.3% 6.77E-09 24H-F1 04 (20-24) failure modes are early failure with vessel at high pressure (65%), and failure due to pre-existing leakage (24%).

This fire scenario results in station blackout and failure of the 1P-29 turbine-driven AFW ELECTRICAL FIRE pump. The dominant containment failure FC318.DYOA-F1 A30 1.2% 6.07E-09 CAB DYOA (AD) modes are early failure with vessel at high pressure (77%), and failure due to pre-existing leakage (12%).

This fire scenario results in a loss of all AFW and bleed-and-feed cooling due to fire effects ELECTRICAL FIRE and human error to start and align available FC305.1A05-62 A24 CAB 1A-05 (62-66) 1.1% pumps. The dominant containment failure 5.64E-09 F1 (ASD) modes are early failure with vessel at high pressure (57%), and failure due to pre-existing leakage (32% ).

This fire scenario results in a loss of all AFW and failure of bleed-and-feed cooling at ELECTRICAL FIRE A01- recirculation. The dominant containment FC237.2B31-F2 CAB 2B-31 (WHOLE 1.1% 5.41 E-09 B/46 failure modes are failure due to pre-existing ROOM) leakage (55%) and early failure with vessel at high pressure (34%1.

Note 1: (ASD) in the scenario description denotes Automatic Suppression and Detection fire modeling cases Page 31 of 48

Table 3 - Unit 2 CDF for Scenarios with >1% Risk Contribution Compliant CDF Scenario ID Fire Area Scenario Description Risk Insights CDF Contribution (per rx-yr)

This fire scenario directly fails all AFW pumps, including the credited safe shutdown pump (2P-53) due to the dynamic impact of the HEAF which fails the fire wrap on the protected train. The fire impacts to various HEAF FIRE FC304N.BUSDUCT- electrical components also result in loss of A23N BUSDUCT S2 73.1% 2.77E-04 S2-FO recirculation capability for both Sl trains, which (TARGET) fails long term decay heat removal by bleed-and-feed cooling. This results in a CCDP of 1.0. The variant case credits a risk reduction modification replacing the bus duct with cable which eliminates the fire scenario entirely.

This fire scenario causes a loss of offsite power and failure of diesel generator G-02, and directly fails the AFW P-38 pumps flow paths as well as the turbine-driven AFW pump SELF IGNITED minimum flow valve. A random failure of FC304N.C-F1 A23N CABLE FIRE 4.5% diesel generator G-01 fails power to the 1.69E-05 (WHOLE ROOM) motor-driven AFW pump 2P-53, resulting in a loss of all AFW. Bleed-and-feed cooling is initially successful, but fails at the time of recirculation due to fire-induced failures of electrical components.

Page 32 of 48

Table 3 - Unit 2 CDF for Scenarios with >1% Risk Contribution

,, Compliant CDF Scenario ID Fire Area Scenario Description Risk Insights CDF Contribution (per rx-yr)

This fire scenario directly fails all AFW pumps, including the credited safe shutdown pump (2P-53) due to the dynamic impact of the HEAF which fails the fire wrap on the protected train. The fire impacts to various electrical components also result in loss of HEAF FIRE recirculation capability for both Sl trains, which FC304N.BUSDUCT-A23N BUSDUCT S2 2.3% fails long term decay heat removal by bleed- 8.64E-06 S2-F1 (WHOLE ROOM) and-feed cooling. This results in a CCDP of 1.0. The variant case credits a risk reduction modification that replaces the bus duct with cable eliminates the fire scenario entirely.

The fire scenario has similar to FC304N.BUSDUCT-S2-FO except for credit for suppression.

This fire scenario results in a loss of offsite power and failure of diesel generators G-01, G-02, and G-04, and battery chargers D-08, D-09, and D-107. This results in failure of ELECTRICAL FIRE FC305.1A05-62 AFW pumps P-38A and B, and motor-driven A24 CAB 1A-05 (62-66) 1.6% 5.96E-06 F1 AFW pump 2P-53. A random failure of the (ASD) 1 turbine-driven AFW pump 2P-29 results in loss of all AFW. Both Sl trains are failed due to loss of power, so bleed-and-feed cooling is

-

unavailable.

Page 33 of 48

Table 3 - Unit 2 CDF for Scenarios with >1% Risk Contribution Compliant CDF Scenario ID Fire Area Scenario Description Risk Insights CDF Contribution (per rx-yr)

This fire scenario results in a loss of offsite power and failure of diesel generators G-01 and G-02, and battery chargers D-07, D-08, D-09, and D-107. This results in failure of ELECTRICAL FIRE FC305.1A05-57 AFW pumps P-38A and B, and motor-driven A24 CAB 1A-05 (57-61) 1.4% 5.12E-06 F1 AFW pump 2P-53. A random failure of the (ASD) turbine-driven AFW pump 2P-29 results in loss of all AFW. Both PORVs are failed due to loss of control power, so bleed-and-feed cooling is unavailable.

The fire scenario results in a loss of offsite power and failure of diesel generators G-01, G-02, and G-04, and battery chargers D-08, D-09, and D-1 07. This results in failure of ELECTRICAL FIRE AFW pumps P-38A and B, and motor-driven FC305.D-08-F1 A24 1.3% 4.93E-06 CAB D-08(ASD) AFW pump 2P-53. A random failure of the turbine-driven AFW pump 2P-29 results in loss of all AFW. Both Sl trains are failed due to loss of power, so bleed-and-feed cooling is unavailable.

Page 34 of 48

Table 3 - Unit 2 CDF for Scenarios with >1% Risk Contribution Compliant CDF Scenario ID Fire Area Scenario Description Risk Insights CDF Contribution (per rx-yr)

This fire scenario causes a loss of offsite power and failure of diesel generator G-02, and directly fails the AFW P-38 pumps flow paths as well as the turbine-driven AFW pump minimum flow valve. A random failure of CABLE FIRES DUE diesel generator G-01 fails power to the FC304N.CWC-F1 A23N TO CABLE AND 1.0% motor-driven AFW pump 2P-53, resulting in a 3.82E-06 WELDING (T) loss of all AFW. Bleed-and-feed cooling is initially successful, but fails at the time of recirculation due to fire-induced failures of electrical components. The fire scenario has I similar impacts as FC304N.C-F1 except that it has a lower severity factor.

This fire scenario results in a letdown line isolation failure causing a small LOCA. The ELECTRICAL FIRE FC187GRP.C-180- LOCA can be mitigated by Sl until A01-B CAB C-180 1.0% 3.81 E-06 F3 recirculation fails due to fire-induced failures (TARGETS) of sump isolation valves and other Sl

- - - - - - L____

equipment.

Note 1: (ASD) in the scenario description denotes Automatic Suppression and Detection fire modeling cases Page 35 of 48

Table 4 - Unit 2 LERF for Scenarios with >1% Risk Contribution Compli~nt I LERF Scenario ID Fire Area Scenario Description Risk Insights LERF I' Contribution (per rx-yr)

This fire scenario results in a loss of all AFW and unavailability of bleed-and-feed cooling.

HEAF FIRE FC304N.BUSDUCT- The dominant containment failure modes are A23N BUSDUCT S2 36.0% 5.70E-07 S2-FO early failure with vessel at high pressure (TARGET)

(49%), and failure due to pre-existing leakage (40%).

This is a dual-unit scenario with control room abandonment based on habitability and Main Control Room A31-ABAND A31 14.4% assumed failures of the 1P-29 Turbine Driven 2.28E-07 Abandonment AFW pump and the Gas Turbine Generator G05.

This fire scenario results in a small LOCA with Sl failure at recirculation. The dominant ELECTRICAL FIRE FC187GRP.C-180- containment failure modes are associated A01-B CAB C-180 8.2% 1.30E-07 F3 with a fire-induced containment isolation valve (TARGETS) failure (containment penetration #9) (74%)

and early with vessel at high pressure (24%).

The fire results in a small LOCA with Sl failure at recirculation. The dominant containment ELECTRICAL FIRE FC187GRP.C- failure modes are associated with a fire-A01-B CAB C-180A 7.3% 1.16E-07 180A-F3 induced containment isolation valve failure (TARGETS)

(containment penetration #9) (74%) and early with vessel at high pressure (24%).

This fire scenario results in a loss of all AFW and unavailability of bleed-and-feed cooling.

HEAF FIRE FC304N.BUSDUCT- The dominant containment failure modes are A23N BUSDUCT S2 3.6% 5.64E-08 S2-F1 early failure with vessel at high pressure (WHOLE ROOM)

(76%) and failure due to pre-existing leakage (13%).

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Table 4 - Unit 2 LERF for Scenarios with >1% Risk Contribution Compliant LERF Scenario lD Fire Area Scenario Description Risk l nsights LERF Contribution (per rx-yr)

This fire scenario results in a fire-induced small or very small LOCA with failure of Sl.

TRANSIENT FIRE The dominant containment failure modes are TS MONITOR FC187GRP.TS3-F1 A01-B 3.5% associated with a fire-induced containment 5.51 E-08 TANK ROOM isolation valve failure (containment (TARGETS) penetration #9) (74%) and early failure with vessel at high pressure (24%).

The fire scenario results in a fire-induced small or very small LOCA with failure of Sl.

TRANSIENT FIRE The dominant containment failure modes are TS MONITOR FC187GRP.TS2-F1 A01-B 2.9% associated with a fire-induced containment 4.67E-08 TANK ROOM isolation valve failure (containment (TARGETS) penetration #9) (74%) and early failure with vessel at high pressure (24%).

This fire scenario results in a loss of all AFW ELECTRICAL FIRE with bleed-and-feed cooling unavailable. The FC305.1A05-62 A24 CAB 1A-05 (62-66) 2.5% dominant containment failure modes are early 3.99E-08 F1 (ASD) 1 failure with vessel at high pressure (77%), and failure due to pre-existing leakage (12%).

This fire scenario results in a loss of all AFW with bleed-and-feed cooling unavailable at SELF IGNITED recirculation. The dominant containment FC304N.C-F1 A23N CABLE FIRE 2.3% 3.63E-08 failure modes are early failure with vessel at (WHOLE ROOM) high pressure (49%), and failure due to pre-existing leakage (40% ).

The fire scenario results in a loss of all AFW with bleed-and-feed cooling unavailable. The ELECTRICAL FIRE FC305. D-08-F1 A24 2.2% dominant containment failure modes are early 3.42E-08 CAB D-08(ASD) failure with vessel at high pressure (77%), and failure due to pre-existing leakage (12%).

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Table 4 - Unit 2 LERF for Scenarios with >1% Risk Contribution Compliant LERF Scenario ID Fire Area Scenario Description Risk Insights LERF Contribution (per rx-yr)

This fire scenario results in a loss of all AFW with bleed-and-feed cooling unavailable due ELECTRICAL FIRE to loss of control power to the pressurizer FC305.1A05-57 A24 CAB 1A-05 (57-61) 2.1% PORVs. The dominant containment failure 3.33E-08 F1 (ASD) modes are early failure with vessel at high pressure (77% ), and failure due to pre-existing leakage (_12%}.

This fire scenario results in a loss of all AFW with bleed-and-feed cooling unavailable. The ELECTRICAL FIRE dominant containment failure modes are FC318.2B03-G33-A30 CAB 2B-03 (G33-37) 1.3% associated with a fire-induced containment 2.03E-08 37-F2 (MS) isolation valve failure (containment penetration #9) (78%) and early failure with vessel at high pressure (18%).

Note 1: (ASD) in the scenario description denotes Automatic Suppression and Detection fire modeling cases Page 38 of 48

d) The assumption that a high energy arc fault (HEAF) can occur in the 480V bus duct and cause failure of the fire wrap for nearby cables results in a significant negative delta risk between the compliant and variant cases. Modifying the plant to remove the HEAF concern for 480V bus ducts in the Auxiliary Feedwater Pump (AFW) rooms is considered a risk reduction modification, because there are no variances from deterministic requirements (VFDRs) related to the HEAF concern. This risk reduction modification is not credited in the compliant model, which results in a significantly higher compliant model risk compared to the variant model risk for Unit 2. No other assumptions significantly contribute to the large reduction credit for modifications.

PRA RAI19- Sensitivity Analysis on FAQ 08-0048 Fire Bin Frequencies NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. NFPA 805 Section 2.4.4. 1 further 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 NRC. RG 1.174 provides quantitative guidelines on core damage frequency, large early release frequency, 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. The NRC staff review of the information in the LAR has identified the following information that is required to fully characterize the risk estimates:

The licensee's analysis indicates that the updated fire bin frequencies provided in NUREG/CR-6850, Supplement 1 (i.e., FAQ-08-0048, ADAMS Accession No. ML092190457) were used in the FPRA. The guidance in FAQ-08-0048 states that a sensitivity study should be performed using the mean fire frequency for those bins in Section 6 of NUREG/CR-6850 with an alpha value less than or equal to one. Indicate if the acceptance guidelines of RG 1.174 may be exceeded when this sensitivity study is applied to the integrated study of PRA RAJ 3. If these guidelines may be exceeded, provide a description of fire protection, or related, measures that can be taken to provide additional defense in depth, as discussed in FAQ 08-0048.

NextEra Response A sensitivity analysis has been performed to address "Guidance for the Use of Revised Fire Ignition Frequencies in NUREG/CR-6850" (FAQ 08-0048) for the updated bin frequencies for those bins with an alpha value less than or equal to one and is included in Section 7 of the "Integrated Analysis of Point Beach Nuclear Plant Fire PRA and NFPA 805 Compliant Model" (P2428-0009-01 ). The analysis shows that the guidelines of RG 1.174 have been exceeded in two locations for Unit 1 and one location for Unit 2 when select historical ignition frequencies are applied to the integrated study of PRA RAI 03. The Unit 2 total risk is at the 1E-4/yr. total risk (CDF) threshold when select historical ignition frequencies are applied to the integrated study of PRA RAI 03. The Unit 1 total risk does not exceed the 1E-4/yr. total risk CDF threshold.

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Compartment 305- This location is the vital switchgear room. The risk threshold would be only slightly exceeded for Unit 1 based on the sensitivity analysis. This location is already one of the most risk significant areas due to loss of power, and the existing defense in depth factors are expected to be sufficient. This area is provided with an automatic Halon suppression system and area-wide detection. Additionally, cables in the area are protected by 1-hour fire wrap.

Qualitative margin to offset the observed risk increase is also provided by proceduralized operator actions in AOP-1 OA (Safe Shutdown- Local Control) that are not currently credited in the PRA analysis, including alternate shutdown and efforts to restore off-site power to vital systems. In addition, defense in depth is realized by credit that has not been taken in the PRA for the latest NRC guidance, which should decrease several hot short probabilities that have increased because of removal of the CPT credit.

Compartment 187GRP- This location is the Monitor Tank Room Primary Auxiliary Building or PAB 26' Elevation). The risk thresholds would be exceeded for both Units in this area, more so for unit 2, as the risk is driven by some cable pinch points that lead to letdown LOCAs caused by multiple spurious operations. This area is provided with automatic detection and 3-hour fire wrap for select cables. Qualitative margin is provided by additional operator actions in AOP-10C (Safe Shutdown Following Fire at PAB 26' Central) that are not currently credited in the Fire PRA, which are available to isolate letdown. These actions, plus the margin expected based on the conservatively applied hot short probabilities, should compensate for the increased sensitivity of this area.

PRA RAI 25- Changes to Modifications Described in LAR Attachment 5 During the NRC staff audit conducted during the week of June 9, 2014, the licensee indicated that changes will be made to the modifications described in LAR Attachment S, Table S-2, "Plant 11 Modifications Committed.

Please specify any additions to, modifications of, or deletions from, the plant modifications 11 identified in LAR AttachmentS, Table S-2, "Plant Modifications Committed, and describe adjustments made to the PRA to credit or remove credit for the new, affected, or deleted modifications.

Justify and assumptions made to support the PRA analysis. In addition, provide a revised LAR Attachment S, Table S-2, which clearly indicates the changes made to the modifications.

NextEra Response AttachmentS, Table S-2 from the Point Beach LAR (Reference 1) has been revised based on the most recent analysis and risk insights developed in response to NRC RAis and additional refinements to the Fire PRA and Safe Shutdown models (see Attachment 5). The primary changes to Attachment S are the deletion of several modifications that are unnecessary to support transition to NFPA 805. In addition, the descriptions of several previously identified modifications have been updated to better reflect the details. The changes are listed below with a basis for the change.

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Item Proposed Modification Type of Change and Justification EC Cable ZE23213CE in Fire Area A01- Scope removed.

279326 B will be protected to preserve VNBI.

Acceptable risk results are achieved with the existing plant design and operation.

Feasible and reliable recovery actions, included in IMP-143 and Attachment G, are relied upon in lieu of the modification.

MOD-1 Bus duct between the B-03 and B-04 Scope changed.

busses will be modified so that a HEAF is no longer a concern in Fire Bus bars will be replaced with cables, which Areas A23N and A23S. are not susceptible to the HEAF failure mode. This change is addressed in the Fire Model reports for the affected areas.

MOD-4 Add additional power inputs to B-08 Clarification of description.

and to B-09. Power to come from tie line independent of switchyard. The increase load capacity is no longer required and this mod will only provide an alternate _12_ower source.

MOD-7 The following cables will be re-routed Clarification of description.

and/or protected from fire damage in fire compartment FC187GRP. The Scope reduced based on circuit analysis re-routes will be reviewed to ensure from EPM Report R2337-0010-01 RO, attachment C and W results are not Evaluation of Spurious Pressurizer PORV significantly impacted. Operation due to Instrument Failures.

ZK11429A Acceptability confirmed by final ZP21429A quantification.

MOD-8 The following PORV cables will be Scope removed.

protected in FZ 511:

Additional refinements in circuit analysis ZK11429Q from EPM Report R2337-0010-01 RO ZL 114317 determined that the cable protection was not ZN11449H required.

ZL 114300 Acceptability confirmed by final quantification.

MOD-9 The following PORV cables will be Scope removed.

protected in FZ 516:

Additional refinements in circuit analysis ZK11429Q from EPM Report R2337-0010-01 RO ZL 114300 determined that the cable protection was not ZM114317 required.

ZN11449H Acceptability confirmed by final quantification.

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MOD- The following cables will be re-routed Scope reduced to a single channel of 10 and/or protected from fire damage in cabling.

fire zone FZ-318.

Additional refinements in circuit analysis 114298-C-D-H-M from EPM Report R2337-0010-01 RO determined that the cable protection was not 114498-E-F required.

ZA1J1368, ZK11429A, ZN11449A Acceptability confirmed by final 214298-C-D-H-M quantification.

214498-E-F ZC2J 1368, ZP21429A, ZS21449A.

MOD- Cables ZE2328CA and ZE2328C8 in Scope removed.

12 FZ 304N will be protected to make P38A available. Minimal risk reduction in the final quantification results.

MOD- Protect cables Z81 8178H, Scope removed.

13 Z81A84A1, and Z81A84A2 in FZ 304S to restore power to 18-04. Acceptable results are achieved with existing plant design and operation, and additional transient combustible controls.

IMP-144 has been updated to include Fire Zone 304S ventilation area.

MOD- To address the potential fail open Scope revised.

14 scenarios associated with multiple spurious operation concerns, Model and circuit analysis has been updated solenoid valves will be installed in and is reflected in final quantification.

the air lines supplying the Condenser Steam Dump Valves and steam inlet valves to the MSRs each on both units with a manually activated switch outside the Control Room.

Cable routing and power supplies will not be in the Cable Spreading Room or the Control Room or dependent on equipment in either area.

MOD- Cables ZD2426MA, ZD2426M8, and Scope removed.

15 ZD2426MC, associated with letdown valve 2RC-427 in Fire Area A01-8, Acceptable results are achieved with will be protected to prevent spurious existing plant design and operation, and LOCA. reliance on an action in the main control room to isolate air to containment, as reflected in EPM report R2168-9999-01.

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MOD- Cable ZB 1426MC for unit 1 and Scope removed.

16 ZD2426MC for unit 2, associated with letdown valve 1/2RC-427 in FZ Acceptable results are achieved with 318, will be protected. existing plant design and operation, and reliance on an action in the main control room to isolate air to containment, as reflected in EPM rer:>ort R2168-9999-01.

MOD- Protect the following cables in the Scope removed.

17 fire areas noted to preserve DC control power to the required OCT analysis confirmed this modification breakers: was not necessary. Acceptability confirmed by final quantification.

Cable ID: Fire Area D3102A A15 D3102A A01-H D4102A A01-G D4102AA02 D4102AA06 D4102AA24 ZAD1107A A30 ZAD11 07A A23S ZFD0406A A30 ZFD0406A A68 ZFD0406A A23N ZFD0206A A23N ZFD0206A A30 ZFD1402A2 A68 ZFD1402A1 A30 ZFD1402A1/A2 A23N ZFD0208A A23N ZCD3109A1 A23S ZCD31 09A 1 A24 ZED0307A A 15 ZED01 DBA A23S MOD- Either Pump P38A or P388 is Scope removed.

18 required to be restored; to restore pump P38A, cables WK114042A and Pump P388 cable not credited in A01-B/46.

ZK11460H in FZ 237, will be Acceptability confirmed by final protected. quantification.

MOD- Reduce dependence on instrument Scope clarification.

23 air for P-38 AOVs by providing 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> pneumatic supply.

MOD- Provide coordinated fuses to prevent Scope removed.

28 cables that are remote to the switchgear from preventing an over OCT analysis confirmed this modification current trip for the following breaker was not necessary. Acceptability confirmed circuits: by final quantification ..

1A52-02, 1A52-06, 1A52-12, 1A52-85 2A52-22, 2A52-31, 2A52-49 Page 43 of 48

MOD- Provide coordinated fuses and Scope removed.

29 additional relays to prevent cables that are remote to the switchgear OCT analysis confirmed this modification from preventing an over current trip was not necessary. Acceptability confirmed for the following breaker circuits: by final quantification.

H52-22, H52-32, H52-16 1A52-05, 1A52-07, 1A52-08, 1A52-09, 1A52-1 0, 1A52-11, 1A52-13, 1A52-15, 1A52-58, 1A52-59, 1A52-84 2A52-19, 2A52-20, 2A52-21, 2A52-23, 2A52-25, 2A52-28, 2A52-30, 2A52-32, 2A52-33, 2A52-67, 2A52-73, 2A52-74, 2A52-75, 2A52-88, 2A52-89 MOD- Provide automatic backup DC power Scope removed.

30 to the following buses independent of fire area A24: OCT analysis confirmed this modification 1-A01, 1-A02, 2-A01, 2-A02 was not necessary. Acceptability confirmed Note: Providing self energized over by final quantification ..

current trip devices on the individual breakers on these buses negates the re_guirement of backup DC power.

MOD- Protect cables (ZCG0201 H, Scope removed.

31 ZCG0201J, ZCB0201T, and ZEG0101T) for breaker 2A5276 Refined fire modeling and circuit analysis between 2A-03 and 2A-05 in FZ 318 confirmed this modification was not from damage due to a fire in cabinet necessary.

DYOA.

MOD- Protect cable ZF1 NB139A for Scope removed.

32 breaker 252391 between 28-39 and D-09 in FZ 318 from damage due to Refined fire modeling and circuit analysis a fire in bus 1B-04. confirmed this modification was not necessary.

MOD- Protect cable D1208A for breaker Scope removed due to refined fire model 33 2A5276 between 2A-03 and 2A-05 in and circuit analysis.

FZ 318 from damage due to a fire in cabinet D-26.

EC Install low suction pressure trip logic New modification.

272841 to at least two charging pumps per and unit. Modification added to provide additional risk 261021 reduction.

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SSARAI 04 LAR Attachment S, Table S-2, includes several modifications associated with "protecting" cables.

If these modifications involve installation of ERFBS to protect the cable, and the modification resolves a VFDR, then identify the VFDR associated with the specific modification and specify which deterministic requirement the ERFBS is meeting (1-hour or 3-hour).

NextEra Response The updated License Amendment Request (LAR) AttachmentS, Table S-2 (see Attachment 5) provided in response to PRA RAI 25 identifies the modifications to protect cables. The two modifications that will credit proposed fire wrap I Electrical Raceway Fire Barrier System (ERFBS) to resolve Variances from Deterministic Requirements (VFDRs) are MOD-11 and MOD-20. The other cable protection modifications identified in Attachment S, Table S-2 will involve a cable re-route or the use of fire rated cable. Refer to the response to RAI FPE 09c.

MOD-11 is protecting Cable ZFD0406A in Fire Area A 15, Fire Zone 166 (2832 MCC Area) with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire barrier to resolve VFDR A 15-16. The proposed protection of this cable with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated ERFBS in conjunction with area-wide detection and wet-pipe sprinkler system that is adequate for the hazards in the area satisfies the deterministic separation requirement of NFPA 805 Section 4.2.3.3(c). The fire wrap protection is considered adequate for the fire scenarios in the area as the fire duration beyond an hour is not expected. Additionally, the compartment contains a wet-pipe sprinkler suppression system, which covers the entire compartment, with the exception of only the MCC area, and protects the entire route of conduit 004-7 through this compartment. There is also area-wide detection and low combustible loading in this area.

MOD-20 is protecting Cables ZF1494A and ZF1494C in Fire Area A01-B to resolve VFDR A01-B-64. The proposed protection of these cables in Fire Area A01-B with a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated ERFBS satisfies the deterministic separation requirement of NFPA 805 Section 4.2.3.3(a).

SSARAI 05 Numerous VFDRs describe a situation where fire damage can cause overcurrent trip (OCT) concerns that could result in a secondary fire. The VFDR disposition states that the condition has not been modeled in the FPRA. The VFDRs state that a qualitative analysis, R2168-1003c-001 Aft. 7, addresses this concern. However, the staff noted that the licensee's analysis recommends numerous modifications in order to preserve overcurrent trip capability. It appears that some of those modifications have been included in LAR AttachmentS, Table S-2.

a) Describe whether all OCT concerns identified in the analysis (R2168-1003c-001 as referenced in the LAR) have been resolved by a proposed modification.

b) Although the licensee's analysis may have recommended addressing the issue with modifications, LAR Attachment C only references the calculation. For those VFDRs that will be resolved through modifications, provide the specific modification listed in LAR Attachment S to accomplish this.

c) The modifications described in LAR AttachmentS in many cases do not reference the appropriate VFDR or provide the breaker/circuit number. For each modification that resolves a VFDR, describe the specific VFDR the proposed modification refers to.

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d) For those VFDRs that will not be resolved through modifications, provide a discussion explaining how the qualitative risk analyses performed, justifies the presence of secondary fires if the condition has not been modeled in the FPRA.

NextEra Response a, b, c) Subsequent to LAR submission (Reference 1), NextEra revisited the Overcurrent Trip (OCT)/ Secondary Fires issue. The qualitative analysis of R2168-1 003c-001 Attachment G was both updated and supplemented by a risk-informed approach as discussed in the response to RAI SSA-05.d, and the OCT logics were entered into the Fire PRA model.

The final quantification demonstrated acceptable results without the MODs recommended by the qualitative analysis; see P2091-2900-02 R2, NFPA-805 Fire PRA Quantification Notebook. As such, no OCT VFDRs are resolved by MODs.

d) Subsequent to License Amendment Request (LAR) submission (Reference 1), Next Era revisited the Overcurrent Trip (OCT)/ Secondary Fires issue. A risk-informed approach was pursued and the OCT logics were entered into the Fire PRA model. The final quantification demonstrated acceptable results without any of the modifications that had previously been considered. See P2091-2900-02 R2, NFPA-805 Fire PRA Quantification Notebook. Refer to the response to PRA RAI 25 summarizing changes to Table S-2 modifications related to OCT/Secondary Fires.

FPE RAJ 09 LAR Attachment A, Section 3.11.5, describes installed ERFBS. In the "Compliance Basis" associated with "Complies with Required Action," the LAR states, in part, "The ERFBS used at PBNP is 1-hour rated with the exception of that installed in containment, which is qualified as radiant energy shielding." In the paragraphs that follow, the LAR identifies a number of locations with 3-hour wrap. Provide the following:

a) Describe whether the 1-hour and 3-hour ERFBS described in LAR Attachment A is currently installed or planned to be installed.

b) Describe whether the ERFBS described in this section is associated with the cable protection modifications described in LAR AttachmentS, Table S-1. If so, identify the modifications listed in Attachment S and associated with this compliance statement.

c) With regard to the LAR Attachment S modifications that state that a given cable will be "protected," describe what is meant by "protected" (i.e., cable will be protected by ERFBS for risk reduction, compliance with deterministic 1-hour requirement, or compliance with deterministic 3-hour requirement).

NextEra Response a) The statement in License Amendment Request (LAR) Attachment A, Section 3.11.5 (Reference 1) that "The ERFBS used at PBNP is 1-hour rated with the exception of that installed in containment, which is qualified as radiant energy shielding" is incorrect. A markup of License Amendment Request (LAR) Attachment A, Section 3.11.5 is provided in Attachment 1. Point Beach Nuclear Plant (PBNP) currently utilizes and credits both 1-hour and 3-hour rated fire wrap in areas outside of containment as identified in the compliance basis for Section 3.11.5.

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Each Electrical Raceway Fire Barrier System (ERFBS) identified in LAR Attachment A, Section 3.11.5 (Reference 1) has been verified to be currently installed in the plant.

b) There are three (3) modifications described in LAR AttachmentS, Table S-1, Plant Modifications Completed (Reference 1): EC 250831, EC 259835 and EC259934.

These design changes are currently installed in the plant. These modifications did not directly install Electrical Raceway Fire Barrier System (ERFBS) for cable protection.

c) The updated LAR AttachmentS, Table S-2 (see Attachment 5) provided in response to PRA RAI 25 identifies the following modifications to either protect or re-reroute cables:

Item Proposed Modification Proposed Protection MOD- The following cables will be re-routed Cable re-route for risk reduction 7 and/or protected from fire damage in fire compartment FC187GRP.

ZK11429A ZP21429A MOD- The following cables will be re-routed Cable re-route with 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated cable 10 and/or protected from fire damage in to partially resolve VFDRs A30-14 (U1) fire zone FZ-318. and A30-27 (U2). 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated cable in this area satisfies the deterministic 11429B, 11429C, 11429D, 11429H, separation requirement of NFPA 805 11429M, 11449B, 11449E, 11449F, Section 4.2.3.3(c).

ZA1J136B, ZK11429A, ZN11449A, 21429B, 21429C, 21429D, 21429H, 21429M, 21449B, 21449E, 21449F, ZC2J136B, ZP21429A, ZS21449A.

MOD- The following cable will be re-routed 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated ERFBS to resolve VFDR 11 and/or protected from fire damage in A15-16. The proposed protection of this fire zone FZ-166. cable with a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated ERFBS in conjunction with area-wide detection and ZFD0406A wet-pipe sprinkler system that is adequate for the hazards in the area satisfies the deterministic separation requirement of NFPA 805 Section 4.2.3.3(c). The fire wrap protection is considered adequate for the fire scenarios in the area as the fire duration beyond an hour is not expected.

Additionally, the compartment contains a wet-pipe sprinkler suppression system which covers the entire compartment, with the exception of only the MCC area, and protects the entire route of conduit D04-7 through this compartment. There is also area-wide detection and low combustible loading in this area.

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Item Proposed Modification Proposed Protection MOD- The following cables will be re-routed 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated ERFBS to resolve VFDR 20 and/or protected from fire damage in A01-B-64. 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire-rated ERFBS in fire area A01-B. This will preserve this area satisfies the deterministic control power from D-1 08. separation requirement of NFPA 805 Section 4.2.3.3(a).

ZF1494A ZF1494C MOD- The following cables will be re-routed Cable re-route and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated cable 21 and/or protected from fire damage in to partially resolve VFDR A30-01. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire zone FZ-318. fire-rated cable in this area, in conjunction with automatic fire detection and ZA1DD6306A automatic fire suppression throughout the ZL 1MOB406A area, satisfies the deterministic separation requirement of NFPA 805 Section 4.2.3.3(c).

MOD- The following cable will be re-routed Cable re-route and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire-rated cable 22 and/or protected from fire damage in to partially resolve VFDR A30-01. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> fire zone FZ-318. fire-rated cable in this area, in conjunction with automatic fire detection and ZA 1D1 NA4000C automatic fire suppression throughout the area, satisfies the deterministic separation requirement of NFPA 805 Section 4.2.3.3(c).

References:

(1) NextEra Energy Point Beach, LLC, letter to NRC, dated June 26, 2013, "License Amendment Request 271, Transition to 10 CFR 50.48(c)- NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition" (ML131820453)

(2) NRC e-mail to NextEra Energy Point Beach, LLC, dated September 9, 2013, "Request for Supplemental Information Regarding the Acceptability of the Proposed Amendment Request" (ML13256A197)

(3) NextEra Energy Point Beach, LLC, letter to NRC, dated September 16, 2013, "License Amendment Request 271 Supplement 1 Transition to 10 CFR 50.48(c)- NFPA 805" (ML13259A273)

(4) NRC letter to NextEra Energy Point Beach, LLC, dated September 25, 2013, "Point Beach, Units 1 and 2 -Acceptance Review of Licensing Action re: License Amendment Request to Transition to NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants" (ML13267A037)

(5) NRC e-mail to NextEra Energy Point Beach, LLC, dated July 8, 2014, "Point Beach Nuclear Plant, Units 1and 2 - Final (Revised) Requests for Additional Information re: License Amendment Request Associated with NFPA 805 (TAC Nos. MF2372 and MF2373)" (ML14189A365)

(6) NextEra Energy Point Beach, LLC, letter to NRC, dated July 29, 2014, "Response (60 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805" (ML14210A645)

(7) NextEra Energy Point Beach, LLC, letter to NRC, dated August 28, 2014, "Response (90 Day) to Request for Additional Information License Amendment Request Associated with NFPA 805" (ML14241A267)

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ATTACHMENT 1 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR ATTACHMENT A, NEI 04-02 TABLE B-1, TRANSITION OF FUNDAMENTAL FIRE PROTECTION PROGRAM & DESIGN ELEMENTS 1 page follows

NextEra PBNP Attachment A- NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements NFPA 805 Ch. 3 Ref. Requirements/Guidance Compliance Statement Compliance Basis Reference Document Complies. with Reauired Per Section 5.4.3 of the Fire Protection Evaluation Report Action FPER, "The use of ERFBS (FPER), Rev. 13 I Section 5.4.3 (commonly known as "cable wraps") at PBNP supports the Internal WE Correspondence NPM requirement to separate 96-0020 from Ksobiech to Maxfield redundant trains of safe dated January 19, 1996 I All shutdown equipment where installation of classical Modification MR 99-033, Fire Wrap physical structural barriers Cables in the AFW Pump Room to may not be practical. Cable Meet Appendix R Separation wraps installed to meet the Requirements I All requirements of Section III.G.2.a of 10 CFR 50 NAMS Action Request 1837986, Appendix R, shall be 3-hour "Ensure Credited Fire Wrap fire rated. One hour fire rated Included in Periodic Surveillance" barriers are acceptable in accordance with the NAMS Work Order 370104, "U2 EC requirements of 10 CFR 50 13399, Replace AF Mtr. Repower Appendix R,Section III.G.2.c from DC Bus." I Task 18, "EC provided fire detection and 13399, 3 Hr Fire Wrap of Conduits,"

suppression are also Task 63, "AFW Phase 2 Fire Wrap furnished. Inside containment EC 13399 (Cond. For 2MS-202012,"

cable wrap may also be used and Task 64, "AFW Phase 3 Fire to provide a non-combustible Wrap EC 13399 (Cond. 004-11)"

radiant energy shield in accordance with Section NAMS Work Order 370133, "Aux III.G.2.f of Appendix R. +Ae Control Board and AFW Mod" I ERFBS used at PBNP is 1 Task 14, "EC 13401, 1 Hr Fire hour rated with the Wrap U1 AFW & U2 South Remote exoeption of that installed in SO Room" oontainrnent, whish is qualified as radiant energy NAMS Work Order 37 4010, "EC shielding." 13403, Install Electrical for New AFW Motors/Instruments" I Task All ERFBS credited for NFPA 24, "EC 13403, 3 Hr Fire Wrap of 805 Chapter 4 compliance at Conduits" PBNP is 3M lnteram E-50 series fire wrap. The wrap NAMS Work Order 374013, "Shift meets the requirements of Power Supply Per EC 13398 Generic Letter 86-1 0 1P-29, And Remaining AFW Mods" Supplement 1 as discussed in I Task 26, "3 Hr Conduit Fire Wrap NPM 96-0020. EC 13398" In fire area A01-A, raceways NAMS Work Order 381971, "AF I Revision o Page A-139 Revision 0 Page C-508

ATTACHMENT 2 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR ATTACHMENT B, NEI 04-02 TABLE B-2, NUCLEAR SAFETY CAPABILITY ASSESSMENT METHODOLOGY REVIEW 4 pages follow

NextEra PBNP Attachment B- NEI 04-02 Table B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location required to meet Appendix R requirements. The issues were reviewed for compliance to NFPA 805 and no additional changes were identified. Refer to Action Request Nos. AR # 01298593, AR # 01298594 for disposition of those instances where there was a lack of coordination. See Alignment Basis for 3.5.2.5 for a discussion of loss of breaker coordination due to loss of DC control power.

480V Switchgear Calculation 2001-0049 identified nine instances where there was a lack of coordination. New breaker settings were identified by the Calculations Reconstitution Project and CAP #028771 (NAMS AR 01224052) was initiated. This CAP is shown as being complete and closed 3/11/2003. With the recommended breaker settings made, the breaker coordination for the 480 volt switchgear busses evaluated in this calculation is complete. In addition, Action Requests 01224052 and 01339558 were issued to disposition many of the issues identified in the calculation. The dispositions were reviewed and there are no additional impacts on NFPA-805.

480V Motor Control Centers Calculation 2004-0030 identified various instances where coordination could not be demonstrated. In each of these cases, the technical evaluation review was able to identify a technical justification to support that there is no impact to NFPA-805 compliance.

Full selective coordination does not exist between the supply breaker to MCC B-21 and all load breakers. Modify MCC B-21 supply breaker settings to provide full coordination. Refer to AttachmentS Table S-2 MOD-26-1.

120VAC Distribution Panels (Addressed in FPTE-2007 -001)

The technical evaluation review identified a potential coordination issue associated with 120VAC distribution panels 1Y103, 1Y104, 2Y103 and 2Y104. Each of these panels contain a 100 ampere breaker associated with Radiation Monitoring System. Since these branch breakers are large thermal magnetic breakers they require long time durations to trip open on low fault levels.

The inverter output breakers associated with powering these distribution panels are designed to trip when fault levels reach 260 amperes or more for a duration in excess of 5 seconds. Since the inverters are capable of generating fault currents in this order of magnitude, this condition could result in the tripping of the associated inverter output breaker upon a fault of the cable(s) providing power to the Radiation Monitoring System. As a result, AR # 01316511 has been initiated to identify this issue. See AttachmentS Table S-2 MOD-26-3.

PBNP does not have a single calculation which addresses both safe shutdown and non-safe shutdown branch feeder protective devices with respect to coordination. As a result, it was necessary for the technical evaluation review to review the various distribution panels required for safe shutdown. With exception of the four 120VAC buses identified above, the technical Revision 0 Page B-85

NextEra PBNP Attachment B- NEI 04-02 Table B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location evaluation was successful in demonstrating that most of the circuits from these panels are routed in alternate shutdown areas or that electrical coordination exists. Since many of the cables are in alternate shutdown fire areas, the failure of these circuits is not of concern with respect to the failure of these power supplies. PBNP has alternate shutdown capability which is independent of these fire areas.

Full selective coordination does not exist regarding the 120VAC safety related instrument busses, as documented in Sargent and Lundy Report, "Safe Shutdown 120VAC Distribution Panel Coordination Evaluation". Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-26-3.

125VDC Distribution Panels A review of calculation N-92-005 Revision 2 concluded that the only lack of coordination potentially impacting NFPA-805 compliance is the lack of coordination offuse D72-14-02 on panel D-14 with fuse D72-02-06 on panel D-02. The replacement offuse D72-02-06 is to be accomplished by AR # 01232138.

Numerous fuses associated with 13.8KV system 125VDC control power have ratings below the normal operating voltage and require replacement. Refer to AR 01877063.

120VAC and 125 VDC Distribution Panels Calculation no. V878-15-CA-02 performed a review of both the 120VAC and 125VDC distribution panels. The 125 VDC branch circuits are coordinated and, as a result, have no impact on NFPA-805 compliance. For the 120 VAC branch circuits, the inverter backup power supply has high available fault current which results in a lack of coordination. In order to achieve coordination, modifications will be performed. Refer to AttachmentS Table S-2 MOD-26-3.

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NextEra PBNP Attachment B- NEI 04-02 Table B Nuclear Safety Capability Assessment Methodology Review NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location NEI 00-01 Ref NEI 00-01 Section 3.0 Guidance 3.5.2.5 Circuit Failures Due to The common enclosure associated circuit concern deals with the possibility of causing secondary failures due to fire Common Enclosure damage to a circuit either whose isolation device fails to isolate the cable fault or protect the faulted cable from Concerns reaching its ignition temperature, or the fire somehow propagates along the cable into adjoining fire areas.

The electrical circuit design for most plants provides proper circuit protection in the form of circuit breakers, fuses and other devices that are designed to isolate cable faults before ignition temperature is reached. Adequate electrical circuit protection and cable sizing are included as part of the original plant electrical design maintained as part of the design change process. Proper protection can be verified by review of as-built drawings and change documentation. Review the fire rated barrier and penetration designs that preclude the propagation of fire from one fire area to the next to demonstrate that adequate measures are in place to alleviate fire propagation concerns.

Applicability Comments Applicable See "Alignment Basis" for exceptions to 'Align with Intent' grade.

Alignment Statement Alignment Basis Reference Documents Aligns with Intent The primary issue is the ignition of a secondary fire which then results in failure of cables local SSAR Section 3.3.2.1 to that fire. FPTE-2007-001 Calculations:

The PBNP SSAR, Section 3.3.2 identifies requirements for the Common Enclosure concern. 2004-0009 Section 6.0 of FPTE-2007-001 provides a qualitative assessment of the Common Enclosure 2001-0049 concern for Point Beach Units 1 and 2. 2004-0030 2005-0005 Overall, the existing PBNP methodology for analysis of Circuit Failures Due to Common N-92-005 Enclosure Concerns is consistent with NEI 00-01, Section 3.5, however existing cable protection V878-15-CA-02 issues for proper circuit protection could lead to loss of electrical distribution equipment affecting R2167-1019-001 Section 8.3.4.1 safe shutdown equipment operability. R2168-1003c-001 Att G P2091-2900-02 FPTE-2007-001 was performed to further resolve these common enclosure issues. Summary results are as follows:

13kV & 4kV Switchgear Calculation 2004-0009 either demonstrated adequate cable protection or a technical justification was provided in the evaluation.

Point Beach reviewed the loss of DC control and the effect of that loss on the ability of breakers to clear a fault. The concern is ignition of the faulted cable in a location other than the affected fire area. Point Beach determined that several breakers at the 4 kV and 13 kV level were susceptible to this failure mode, refer to R2168-1 003c-001 Att G. A risk-informed approach was utilized and this condition was incorporated into the Fire PRA model. The final quantification demonstrated acceptable results. See P2091-2900-02. NFPA-805 Fire PRA Quantification Notebook.

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NFPA 805 Section: 2.4.2.3 Nuclear Safety Equipment and Cable Location 480V Switchgear Calculation 2001-0049 identified several instances where cable protection was in question.

However, it was demonstrated that these cables were either in the same fire area as their associated power supplies or that the fault current only exceeds the thermal withstand for a short distance from the power supply at which point the fault current decreased to an acceptable level.

480V Motor Control Centers Nine cables were identified in Sargent and Lundy Calculation 2013-01785 that are not fully protected for overloads in the short time region. Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-24.

Numerous cables were identified in Sargent and Lundy Calculation 2013-01785 that are not fully protected for short circuits in the instantaneous region. However, the fault current only exceeds the thermal withstand for a short distance from the power supply at which point the fault current decreases to an acceptable level.

Evaluation of cable protection for cables supplied from non-safe shutdown 480V MCCs and power panels, and 208/120V lighting panels is incomplete (Ref. AR 1877063). Modifications are being implemented to eliminate this vulnerability. See AttachmentS Table S-2 MOD-26-1 and MOD-26-3.

120VAC Distribution Panels Calculation 2005-0005 addresses the thermal withstand capability of power cables directly connected to the individual panel busses. Although a large portion of the cables were shown to have sufficient thermal withstand capability, there were a number cables which were deficient. As a result, FPTE-2007-001 addressed many of these deficiencies by either identifying those circuits routed in a single fire area or in Alternate Shutdown fire areas only. Only single circuit failed meet one of these conditions. Further analysis was performed on the cable to demonstrate that its thermal withstand capability would not be exceeded for a fire outside of the fire area of the power supply.

125VDC Distribution Panels No issues have been identified regarding protection of 125VDC system cables.

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ATTACHMENT 3 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR, ATTACHMENT C NEI 04-02 TABLE B-3, FIRE AREA TRANSITION 82 pages follow

ATTACHMENT 4 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR, ATTACHMENT G RECOVERY ACTIONS TRANSITION 7 pages follow

ATTACHMENT 5 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR, ATTACHMENTS, TABLE S-2, PLANT MODIFICATIONS COMMITTED 25 pages follow

ATTACHMENT 6 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR, ATTACHMENTS TABLE S-3, IMPLEMENTATION ITEMS 1 page follows

ATTACHMENT 7 TO ENCLOSURE 2 NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT MARKED-UP LAR ATTACHMENT W FIRE PRA INSIGHTS 33 pages follow