ML13123A231

Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805

ML13123A231
Person / Time
Site:	Brunswick
Issue date:	05/15/2013
From:	Gratton C Plant Licensing Branch II
To:	Annacone M Carolina Power & Light Co
Gratton Chris NRR/DORL/LPL2-2
References
TAC ME9623, TAC ME9624
Download: ML13123A231 (32)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 May 15, 2013 Mr. Michael J. Annacone, Vice President Brunswick Steam Electric Plant Carolina Power & Light Company Post Office Box 10429 Southport, North Carolina 28461

SUBJECT:

BRUNSWICK STEAM ELECTRIC PLANT, UNITS 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING VOLUNTARY RISK INITIATIVE NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 (TAC NOS.

ME9623 AND ME9624)

Dear Mr. Annacone:

By letter dated September 25, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12285A428), Carolina Power & Light Company proposed to amend the operating license for the Brunswick Steam Electric Plant, Unit Nos. 1 and 2, by adopting a new risk-informed performance-based fire protection licensing basis in accordance with National Fire Protection Association Standard 805.

The Nuclear Regulatory Commission (NRC) staff has reviewed the licensee's application and has identified the need for additional information to complete its evaluation of the proposed change.

On April 16, 2013, the NRC staff forwarded via email a draft of the request for additional information (RAI) to the licensee. On May 2, 2013, the NRC staff and representatives of the licensee held a conference call to provide the licensee with an opportunity to clarify any portion of the information request and to discuss the response schedule. As a result of that conference call, the licensee agreed to the following schedule for responding to the staffs information request:

60-day Responses Section title Question Number(s)

Programmatic 1,2,3,4,5,6,7 Fire Protection Engineering 1 Safe Shutdown Analysis 3,4,6,7,8,10,12 1A, 1B, 1C, 1D, 1F, 1G, 1H, 11, 1K, 1N, 10, 1P, 1Q, Probabilistic Risk Assessment 1R,4,5,9, 10, 17, 18 Fire Modeling 1A, 'I E, 1F, 1G, 1H, 2A, 2B, 5A, 5B

M. Annacone -2 gO-day Responses Section title Question Number(s)

Radiation Release 1,2,3 Fire Protection Engineering 3,4,5,6,7,8,9,10,11,12,13,14,15,16,17,18,19, 20, 21 Safe Shutdown Analysis 1,2,5,9,11,13,14 Probabilistic Risk Assessment 1J, 1M, 2, 3, 6, 7,11,12,13,14,15,16 Fire Modeling 1B, 2C, 5, 5C 120-day Responses Section title Question Nurnber(s)

Fire Protection Engineering 2 Safe Shutdown Analysis 15 Probabilistic Risk Assessment 1, 1E, 1L, 8 Fire Modeling 1C, 10, 11,2,20,3,4,6 The NRC staff's information request is enclosed. Please note that review efforts on this task are continuing and additional RAls may be forthcoming.

If you have any questions regarding this letter, please feel free to contact me at (301) 415-1055.

Sincerely, Christopher Gratton, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

Request for Additional Information cc w/encl: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 REQUEST FOR ADDITIONAL INFORMATON VOLUNTARY FIRE PROTECTION RISK INITIATIVE CAROLINA POWER & LIGHT BRUNSWICK STEAM ELECTRIC PLANT. UNITS 1 AND 2 DOCKET NOS. 50-325 AND 50-324 Safe Shutdown Analysis (SSA) Request for Additional Information (RAI) 01 In Attachment S, Table S-1, Item #1, of the license amendment request (LAR) dated September 25, 2013 (Agencywide Documents Access and Management System (ADAMS)

Accession Nos. ML12285A428 and ML12285A430) an incipient detection system is identified to be installed in Main Control Room (MCR) cabinets. Based on the operator recognizing the impacted cabinet(s) fire location sufficiently early, describe what operator actions are necessary to limit fire impact and allow safe shutdown of the plant from the MCR or delay to alternate shutdown. Describe how the operator will be made aware of what must be done to remain in the MCR for plant shutdown. Include discussion of alarms, procedures, and training.

SSA RAI 02 Attachment S, Table S-1, Items #5 and #7 of the LAR provide an electrical raceway fire barrier system (ERFBS) wrap in the MCR. Provide more detail regarding the separation scheme being provided in the MCR by this modification. Include in the description the protection scheme provided for large early release frequency (LERF) risk reduction (Item #7). Describe the intent of the modification in the MCR. Include the hourly rating that is being provided for these configurations and describe the separation criterion that is being met.

SSARAI03 Attachment S, Table S-1, Item #10 of the LAR currently lists a modification to "address valve pressure boundary issues due to fire induced spurious actuations." The Table S-1 states "evaluate and modify valves, as necessary, to address pressure boundary concerns due to fire induced spurious actuations. Perform a study for the extent of condition for valves of concern."

Attachment S, Table S-2, Implementation Item #8, of the LAR addresses a study to evaluate the extent of condition related to spurious operation of pressure boundary valves. Describe how these components are included in the nuclear safety capability assessment (NSCA) and how they are subsequently treated in the fire probabilistic risk assessment (FPRA). Describe the scope, methods, and implications for impact to the NSCA and FPRA of this study.

Enclosure

-2 SSA RAI 04 Attachment B, Table B-2, Section 3.1.1.9, of the LAR, (72-hour coping), indicates that the alternate shutdown methodology ensures cold shutdown can be achieved in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, including repairs. However, the cold shutdown actions including repairs are not identified as variances from deterministic requirements (VFDRs). It also states that the analysis may be modified in the future because National Fire Protection Association Standard 805 "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, (NFPA 805) does not have a cold shutdown requirement. Section 4.2.1.2 of the LAR indicates that based on the criteria discussed in the NCSA calculation for safe shutdown, the NFPA 805 licensing basis is to achieve and maintain hot shutdown conditions following any fire occurring prior to establishing cold shutdown. This appears to include cold shutdown as part of the "safe and stable" plant condition being achieved, which would require actions and repairs necessary to be addressed as VFDRs. Describe the plant mode that the operator is attempting to achieve and maintain for safe and stable. NFPA 805 requires the plant to achieve and maintain safe and stable conditions. Provide additional information that would justify not identifying VFDRs for an analysis that "ensures cold shutdown can be achieved in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />."

SSA RAI 05 Section 4.2.1.2 of the LAR for safe and stable condition(s) achieved, provides a qualitative evaluation of the risk for achieving and maintaining safe and stable conditions, including the aspects of having to perform repairs in order to achieve cold shutdown in the event that it is necessary during the post-fire "long-term strategy" described in LAR Section 4.2.1.2. Provide justification for any low-risk conclusions.

Provide a more detailed description of the systems, evolutions, and resources required to maintain this condition between hot standby and cold shutdown. Include the following items:

a. Specific capabilities and required actions to maintain safe and stable for an extended duration (beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) including a qualitative description of the risk.

b. Capacity limitations for each applicable performance goal. Provide a description of capacity limitations and time-critical actions for other systems needed to maintain safe and stable conditions (e.g., gas/air supply for control valves, boron supply, direct current (DC) battery power, diesel fuel, water resources).

c. Describe in more detail the resource (staffing) requirements and timing of operator actions to recover NSCA equipment to sustain safe and stable conditions. Describe how soon "off-shift" personnel will be required to perform functions necessary to maintain safe and stable.

d. Provide a more detailed description of the risk of failure of operator actions and equipment necessary to sustain safe and stable conditions.

-3 SSA RAI 06 Attachment B, Table B-2, Section 3.S.2.1, of the LAR for current transformer open circuit potential of secondary fires, indicates that analysis of open circuits on high voltage (e.g., 4.16 kilo-volt (kV>> ammeter current transformers was completed, and the final disposition of this potential fire scenario is assessed as part of the analysis. Section 4.2.1.1 of the LAR states that the evaluation concludes that this failure mode is unlikely for control transformers (CTs) that could pose a threat to safe shutdown equipment. Provide a more specific description and justification of this conclusion, and include the aspects of secondary fires that may be created and subsequently impact the NSCA. Describe the analysis method and provide the outcome for damage to the safe shutdown (SSD) equipment where the CT is mounted. If fire models were performed to satisfy resolution of fire area failures, then provide verification and validation (V&V) information in Attachment J of the LAR.

SSA RAI 07 For breaker fuse coordination, describe whether cable length was considered as additional impedance in the stucfy necessary to meet maximum available short circuit current. Alternating current (AC) and DC coordination procedure (EGR-NGGC-0106) indicates that the impedance length of the cable can be 10 feet or 10 percent (%) of the cable length (whichever is less), or longer where justified. If this qualification was used, describe how this length was factored into the potential impact to the FPRA. For establishing targets in the zone of influence (ZOI) describe how cable lengths were considered and provide any justifications required for the FPRA.

SSA RAI 08 The LAR did not appear to include table entries for ERFBS by fire area. Provide a list of fire areas that rely on ERFBS for compliance with NFPA 80S. Additionally provide the reason(s) for relying on the ERFBS.

SSA RAI 09 Section 4.S.2.2, Step 3, of the LAR defines the defense-in-depth (DID) and safety margin criteria consistent with the Nuclear Energy Institute (NEI) LAR template and other submittals. However, these criteria were not discussed in Attachment C of the LAR on an area-by-area basis or in the resolution of VFDRs. Evaluations of DID and safety margin are stated to be performed as part of the area-by-area Fire Risk Evaluations. The DID echelons, as defined in NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR [Title 10 of the Code of Federal Regulations] S0.48(c)," Revision 2, and the general strategy of looking for sUbstantial imbalance in the echelons is described at a high level manner in Section 4.S.2.2 of the LAR. However, the specific criteria used to perform DID and safety margin evaluation is not provided in the LAR. Provide a more detailed description and summaries regarding the DID and safety margin established for fire areas that used the NFPA 80S performance-based Section 4.2.4 compliance strategy.

-4 SSA RAI10 Attachment C, Fire Areas RB1-1 and RB2-1 of the LAR are evaluated using both deterministic (4.2.3) and performance-based (4.2.4) methods in the same fire areas. Provide additional explanation to provide a better understanding of the approach in these areas. These areas are also identified as having recovery actions (RAs). NFPA 805 excludes the ability to classify an area as deterministically compliant with RAs. Justify the use of RAs in what appears to be deterministically compliant areas. Provide only one strategy for each fire area. Include any other fire areas that are currently represented as compliant with both deterministic and performance-based strategies.

SSA RAI11 Attachment D of the LAR describes the methods and results for non-power operations (NPO) transition. Provide the following additional information:

a. Provide a list of the components (including power supplies) added, that were not included in the at-power analysis and a list of those at-power components that have a different functional requirement for NPO.

b. Provide a list of key safety features (KSF) pinch points by fire area that were identified in the NPO fire area reviews including a summary level identification of unavailable paths in each fire area.

c. Provide a description of any actions that are credited to minimize the impact of fire induced spurious actuations on power operated valves (e.g., air-operated valves and motor-operated valves) during NPO either as pre-fire plant configuring or as required during the fire response recovery.

d. Identify locations where KSFs are achieved via RAs or for which instrumentation not already included in the at-power analysis is needed to support RAs required to maintain safe and stable conditions. Identify those RAs and instrumentation relied upon in NPO and describe how RA feasibility is evaluated. Include in the description whether these variables have been or will be factored into operator procedures supporting these actions.

e. Describe any new, changed, or deleted manual operator actions resulting from Attachment S, Item 1 of the LAR, "Implement the results of the Non-Power Operational Modes Analysis. Technical and administrative procedures and documents that relate to non-power modes of plant operating states will be revised as needed for implementation."

-5 SSA RAI12 Attachment G of the LAR under the heading, "Results of Step 4," contains an incomplete reference to the feasibility assessment as follows, "contained in Change Package BNP-."

Provide the complete reference.

SSA RAI13 Table G-1, Unit 1 Recovery Actions for CB-23E of the LAR identifies some Unit 2 components, for example:

• 2-DG4-GEN DIESEL GENERATOR NO 4 Take local control of 2-DG4-GEN at EDG #4 Control Panel, located in fire zone DG-02 and operate as required.

• 2-E4-AJ9-FTO COMPT FOR INCOMING LINE FROM SWGR 2C De-energize DC Control Power to 2-E4-AJ9 at Bus 2-E4, Compt AJ9, located in fire zone DG-14. Then verify tripped/manually trip 2-E4-AJ9, in fire zone DG-14.

The same entries are found for the U2 Recovery Actions (Table G-2).

Table G-2 Unit 2 Recovery Actions for CB-23E of the LAR identifies some Unit 1 components for example:

• 1-E6-AV4 - UNIT SUBSTATION E6 MAIN FEED BKR COMPT - Take local control of 1-E6-AV4 at Bus 1-E6 located in fire zone DG-07 and operate as required.

Provide additional information for the following:

a. Describe whether this means that some components support shutdown for both units simultaneously.

b. Describe whether these cross-connecting actions require staff from both units. If so, describe how the feasibility analysis reflects Unit 1 and Unit 2 staffing, communication, and operational interface.

c. Describe the operational impacts on the unaffected (by fire) unit created by cross-connecting these systems.

d. Describe whether the FPRA considers by analysis, only one unit shut down for a fire in the MCR. If so, provide the contribution to Unit 1 risk (core damage frequency (CDF) and LERF) due to a fire requiring shutdown in Unit 2 and vice-versa.

e. Describe whether the Technical Specifications accommodate such cross connections.

-6 SSA RAI14 Attachment G of the LAR states that "In accordance with the guidance provided in NEI 04-02, FAQ 07-0030, Revision 5, and Regulatory Guide (RG) 1.205, the following methodology was used to determine recovery actions required for compliance (i.e., determining the population of post-transition recovery actions)," and that "these actions were described in Section 6.2 of the 1984 ASCA report under "Alternative Shutdown Control Stations." The applicable safety evaluation (SE) was issued on December 30, 1986 (Serial: BSEP-86-805).

a. Describe whether all of the actions (primary control station (PCS) and RA) have been individually reviewed and approved in the 1984 SE identified in Attachment G.

b. Describe whether the location or locations of all of the actions become primary when command and control is shifted from the MCR to these other locations.

c. Describe whether the actions in both cases meet the criteria in RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, Sections 2.4 a. and b.

SSA RAI15 During the audit, it was discussed that Table B-3 of the LAR would be updated to identify dispositions for each VFDR. Provide the updated Attachment C, Table B-3 of the LAR.

Fire Protection Engineering (FPE) RAI 01 Attachment S, Table S-1, Item #1 of the LAR identifies the proposed installation of incipient detection system(s) for cabinets in the MCR. Provide more details regarding NFPA code(s) of record, proposed installation configuration (common piping or individual cabinet), acceptance testing, sensitivity and setpoint control(s), alarm response procedures and training, and routine inspection, testing, and maintenance that will be implemented to credit the new incipient detection system. If the system has not yet been designed or installed, provide the specified design features for the proposed system along with a comparison of these specified design features to their role in satisfying or supporting the risk reduction features being credited in frequently asked question (FAQ) 08-0046 (ADAMS Accession No. ML093220426). Include in this description the installation testing criteria to be met prior to operation. Describe whether this installation and the credit that will be taken will be in compliance with each of the method elements, limitations and criteria of NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," Chapter 13, and FAQ 08-0046 including the closeout memo. Provide justification for any deviations.

FPE RAI 02 Attachment S, Table S-1, Item # 11 of the LAR identifies a modification of suspended ceiling configuration to allow for an effective increase in ceiling height and associated volume of the MCR. Provide a more detailed explanation of what this modification entails. Describe whether the suspended ceiling is to be removed. Describe how this modification will affect the fire

-7 detection systems (including the potential for stratification), both current and planned detection.

Describe how this modification will be incorporated into the fire protection program.

FPE RAI 03 Section 5.5 of the LAR indicates modifications will be completed by the startup of the second refueling outage (RFO) for each unit after issuance of the SE. Describe the basis for extending completion until the end of the second RFO after approval.

FPE RAI 04 Attachment A, Table B-1, Section 3.3(2) of the LAR for design controls that are used to restrict combustibles, indicates two compliance strategies "complies" and "complies via EEEE" (existing engineering equivalency evaluation). Provide a description of what portion of this requirement "complies via EEEE."

FPE RAI 05 Attachment A, Table B-1, Section 3.3.2 of the LAR indicates two compliance strategies; "complies" and "complies via EEEE." Provide a description of what portion of this requirement "complies via EEEE." Because the references identify a structural steel fireproofing calculation for only one specific modification package dealing only with the west walls of the control building elevator shaft, describe whether it can be assumed that the "complies via EEEE" is only this specific scope and that all other aspects of the plant complies.

FPE RAI 06 Attachment A, Table B-1, Section 3.3.5.2 of the LAR identifies the requirement that only metal tray and metal conduits shall be used for electrical raceways. The compliance strategy indicates "complies via previous NRC approvaL" However, the section of the 1977 Safety Evaluation Report (SER) (5.1) cited in the LAR addresses only cable access ways in the control building for safety related equipment. Describe whether there are any non-metal tray or conduit raceways outside the control building. This "previous approval" does not encompass the extent of the NFPA 805 requirement for all tray and conduit electrical raceway. Provide additional detail sufficient to allow "previous NRC approval" or submit an alternative compliance strategy.

FPE RAI 07 Attachment A, Table B-1, Section 3.3.5.3 of the LAR states three levels of compliance in the "Compliance Statement" column, but only defines the compliance basis for "complies with clarification" and "complies via previous NRC ApprovaL" Provide a specific description of what portion of this requirement is satisfied by the EEEE.

FPE RAI 08 Attachment A, Table B-1, Section 3.3.6 of the LAR indicates compliance by "clarification" and identifies compliance with an equivalent Auxiliary and Power Conversion Systems Branch's

- 8 Branch Technical Position 9.5-1 requirement (current licensing basis) as the clarification. The compliance is with a different standard than that listed in NFPA 805 and, therefore would need to be justified as a suitably equivalent standard to Class A of NFPA 256, "Standard Methods of Fire Tests of Roof Coverings." Provide sufficient justification regarding Class I, Factory Mutual System Approval Guide, as equivalent to Class A, NFPA 256.

FPE RAI 09 Attachment A, Table B-1, Section 3.3.7.1 of the LAR includes the compliance strategy regarding storage of flammable gas and states that "No flammable gases are stored in safety related buildings." However, the same compliance statement also states that "The bulk flammable gas stored in the Reactor Buildings, Diesel Generator Rooms, and [AOG Augmented Off Gas/Auxiliary Off-Gas] AOG Building, as approved in the SER, are still in use at BSEP

[Brunswick Steam Electric Plant]." Clarify this apparent contradiction and cite the SER section that approves the locations of this flammable gas. Additionally, the LAR references SER Section 6.3, "Control of Combustibles," as the previous approval for gas storage. This appears to be incompatible. Provide clarification regarding why SER Section 6.3 applies to flammable gas storage or identify the appropriate section(s).

FPE RAI10 Attachment A, Table B-1, Section 3.3.9 of the LAR was omitted. Revise the B-1 Table to include Section 3.3.9 "Transformers." Provide the appropriate information and compliance strategy for all applicable transformers. In providing the appropriate information for the compliance strategy, include an explanation of Table S-1, Item #2 of the LAR, to "provide a method to ensure the compliance with NFPA 805." Explain what this modification entails and how it relates to code compliance.

FPE RAI11 Attachment A, Table B-1, Section 3.5.5 of the LAR, identifies compliance with fire pump separation from each other and from the rest of the plant by rated fire barriers. Table B-1 of the LAR indicates "complies" with "no additional clarification." The referenced design basis, "DBD-62 , Water Based Suppression System" addresses the pump separation from each other in Section 3.3.5 as "flame impingement barriers." Describe whether this separation includes the pumps, controllers, and drivers. Describe whether this "flame impingement barrier" is fire rated as required by NFPA 805. If so, describe the rating that is provided. Provide a detailed description of the separation credited. Describe the bases for how the configuration meets the NFPA 805 requirement of separation by rated barriers.

FPE RAI12 Attachment A, Table B-1, Section 3.5.15 of the LAR, states compliance by "previous NRC approval." The 1977 SER indicated that the proposed extension of the loop to the Service Water Intake Structure (SWIS) required two additional hydrants for improved coverage. The LAR compliance strategy indicates that "in association with upgrades for the Service Water Intake Structure, a nearby yard hydrant will be installed" and stated that this was accomplished.

- 9 Describe the number of hydrants that were installed to meet the conditional approval of the 1977 SER Section 4.3.1 (3). Provide additional information to demonstrate the 1977 SER prerequisite was fully met.

FPE RAI13 LAR Attachment A, Table B-1, Section 3.6.5 was omitted. Revise the B-1 Table to include Section 3.6.5 "Seismic Hose Stations." Provide the appropriate information and compliance strategy.

FPE RAI14 Table B-1, Section 3.11.4 of the LAR identifies three compliance strategies, but there is nothing written in the compliance basis for "Complies via Previous NRC Approval," or "Complies via EEEE." Provide more detail regarding these two compliance strategies to clarify which portions of the requirements apply to which strategies.

FPE RAI15 Attachment A, Table B-1, Section 3.11.5 of the LAR includes ERFBS being identified as part of the compliance strategy. The compliance is achieved by "Complies" and "Complies via EEEE."

There is no attempt to differentiate the two in terms of compliance. Provide a detailed description of what portion of the requirement is satisfied by "Complies" and what portion of the requirement is satisfied by the "Complies via EEEE."

Specifically, for the Pyrocrete ERFBS in the Diesel Generator Building EDG Cell #1, it is not apparent in which compliance category this barrier falls. There is no referenced EEEE for Pyrocrete in Table B-1, Section 3.11.5 of the LAR, however BNP-PSA-080 Attachment 23 indicates there is an "adequate for the hazard" evaluation for the configuration even though it does not comply with Generic Letter 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Area" (Evaluation 85-125-0-10-F Revision 1). Provide clarification with regard to the compliance strategy for the Pyrocrete barrier credited as ERFBS in the FPRA.

FPE RAI16 Attachment A, Table B-1, Section 3.2.3 of the LAR and Attachment S, Table S-2, Implementation Item #5 of the LAR indicate the intent to use the performance-based frequencies from Electric Power Research Institute (EPRI) Technical Report TR-1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features." The adoption of the EPRI method as a performance-based alternative to the deterministic Chapter 3 element requires approval in accordance with 10 CFR 50.48(c)(2)(vii).

Address whether EPRI TR 1006756 is intended as an alternative, and, if so, provide the appropriate supporting information consistent with Section 50.48(c)(2)(vii).

- 10 FPE RAI17 Attachment C, Table B-3 of the LAR for Fire Areas RB1-6 and RB2-6 Mini Steam Tunnels, describes one sprinkler head placed over one safety related reactor core isolation cooling (RCIC) Steam Isolation Valve in a deterministically compliant fire area. Both of these areas are identified as compliant by deterministic Section 4.2.3 of NFPA 805. Provide more detail regarding the intent of fire protection separation scheme and a justification of deterministic compliance. Describe whether the single sprinkler head contributes in any way to deterministic compliance. The Fire Safety Analysis for RB1-6 in Section A6.1 DID indicates that fire detection and suppression will be credited and designated as DID. However, this area is deterministically compliant. Provide more information regarding this issue.

FPE RAI18 NFPA 805, Section 3.5.16, Dedicated Fire Protection Water, states: "The fire protection water supply system shall be dedicated for fire protection use only." Attachment L, Approval Request

1. 1 of the LAR identifies twelve uses of the fire water system other than for fire protection purposes. The evaluation needs to address the potential impact of each of these evolutions on the availability of the fire protection system being capable of meeting its primary function. If during the conduct of each of these alternative uses, there is the possibility of simultaneous demand for fire protection purposes, provide the following:

a. For each of these operations provide the estimated flow and pressure demand requirement for the system uses over and above the fire protection design demand if they were to be concurrent. Describe any of these operations that may be simultaneously performed. Include the design demand conditions required of the fire protection water systems.

b. Identify what restoration requirements (such as tank refilling including time restraints) are needed to restore the standby nature of the fire protection system(s). Describe the engineering design features, design controls, or alarm features that are in place to prevent these operations from impairing the ability of the fire protection systems to meet demand.

c. Describe the administrative controls, procedures, communications, equipment, training, and work control practices that are in place to preclude interference with the ability of the fire protection systems to meet demand.

d. Attachment L of the LAR states that the fire protection tank level shall be maintained with a minimum contained volume of 232,500 gallons (corresponding to a level of 24' 9-1/2"),

and the demineralized water tank, with a minimum contained volume of 90,000 gallons (corresponding to a level of 14' 0"). Describe the controls, alerts, and annunciators that are in place to prevent these requirements from being violated. Include the rate or how quickly the required levels can be restored. Describe whether the procedures and level instrumentation use the same units of measure (e.g., feet, or gallons).

e. Provide justification why the use of the fire protection water supply is allowed for normal evolutions. The use of the fire protection water supply for abnormal or emergency conditions when no other sufficient source is available seems reasonable, but using it for the purposes that follow will require further justification:

- 11

i. Residual Heat Removal (RHR) Service Water Shutdown and wet layup process.

ii. Flushing, filling, and venting RHR service water and heat exchangers.

iii. RHR Service Water System Operability Test.

iv. Flushing Radwaste Radiation Monitor.

v. Seal water to Storm Orain Collector Basin Pumps.

vi. Temporary Cooling Water Supply to Service Air Compressor 1(2) O.

vii. Transfer of Fire Protection System Water Supply to the make-up demineralizer Tank.

viii. Refill of standby gas treatment drain trough.

FPE RAI19 Attachment C, Table B-3 of the LAR identifies the "Required Regulatory Systems" for each applicable fire area. For fire areas with deterministic compliance strategies (4.2.3), there appear to be numerous cases where suppression systems and detection systems are identified as required systems for 010 performance-based compliance. For example; Fire Areas OG-2 identifies Flame detection as required for "0" 010 (4.2.4). Other cases include Fire Areas OG-3, OG-6, OG-9, OG-13, OG-19, OG-20, OG-21 , OG-22, MWT-1, RB1-6, and RB2-6. Provide clarification regarding the apparent discrepancy.

FPE RAI20 Attachment A, Table B-1, Section 3.3.1.3.1 of the LAR, indicates that the hot work process will be controlled by procedures including FIR-NGGC-0003 "Hot Work Permit." Section 3.16 of that procedure indicates that "roving Hot Work Fire Watches" are used during operating Modes 4 and 5. The roving fire watch is allowed to monitor "several hot work jobs in relatively close proximity to each other." Additionally, Section 4.8.9 of that procedure indicates that using a video camera and monitor is acceptable for viewing hot work activities. The NRC staff position is that neither of these practices are recognized by NFPA 5'1 B, "Standard for Fire Prevention during Welding, Cutting, and other Hot Work." Provide the bases why these practices are considered acceptable for compliance with NFPA 805, Section 3.3.1.3.1.

FPE RAI21 Attachment A, Table B-1, Section 3.4.1 (c) of the LAR requires that the brigade leader and at least two brigade members have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance. Oescribe the duties of the Fire Brigade Operations Advisor. If this advisor performs any other duties not in direct support of the fire brigade, provide an evaluation in compliance with the 10 CFR 50.48(c)(2)(vii) including 010 and safety margin that justifies any additional duties.

- 12 Probabilistic Risk Assessment (PRA) RAI 01 - FPRA Facts and Observations (F&Os)

Clarify the following dispositions to fire F&Os and supporting requirement (SR) assessments identified in Attachment V of the LAR that have the potential to impact the FPRA results and do appear to be fully resolved:

a) F&O 1-8 against ES-A1 (Not Met), ES-A2 (Cat-IIII/III), ES-A3 (Not Met), and FQ-A2 (Cat 11111111):

Attachment 8 of BNP-PSA-085 shows in a table whether, and in some cases, how internal event initiators were addressed in the FPRA. Describe how equipment, whose fire-induced failure could cause initiating events, was matched to the appropriate plant response models (i.e., internal events accident sequences). Given the cited sensitivity study results, justify treating the cited initiators as fire-induced failure of equipment following a plant trip rather than using the internal events plant response models associated with internal event initiators.

b) F&O 1-9 against ES-A1 (Not Met), ES-A4 (Cat-llll), and FQ-A2 (Cat 1111/111):

The disposition to this F&O indicates that the independence of High-Pressure Coolant Injection (HPCI) and RCIC is a source of uncertainty. Explain how the dependency between HPCI and RCIC was accounted for in the FPRA, including a discussion on uncertainty as appropriate.

c) F&O 1-14 against PRM-B4 (Cat-IIIIIIII):

This F&O indicates that cable tracing was not performed in some cases. In areas where cable tracing was not performed, identify the assumptions made about possible plant trips and fire induced failures. Was an "exclusionary approach" used that assumes cable failure in any areas where the presence of cable cannot be ruled out?

d) F&O 1-19 against FSS-A1 (Not Met):

The disposition for this F&O explains that the lOI associated with a 143 kilo-watt (kW) heat release rate (HRR) (75th percentile) transient fire was used in all fires areas, except the turbine building where a lOI for a 317 kW HRR (98th percentile) fire was used. The disposition provides the basis for this lower HRR as existing and planned administrative controls, plant experience, and insights from a bounding sensitivity study. Provide further justification for the use of 143 kW transient fires, given that both 143 kW and 317 kW are taken from the same HRR distribution. Include further description of the administrative controls used in the different areas for managing transient combustibles, the results of reviewing plant experience and records of violations of transient combustible controls, other key factors for this reduced fire size, and the results of the bounding sensitivity study referred to in the disposition. Also, confirm that 143 kWand 317 kW HRRs were the only transient fire sizes used in the FPRA.

- 13 e) F&O 1-20 against FSS-A1 (Not Met):

As stated in the disposition, Appendix H.2 of NUREG/CR-6850 recommends that vulnerability to transient fires be limited to cable vulnerability. However, Appendix H.2 also recommends that if sensitive electronics can be impacted, then ignition of such components should be considered. Describe how the impact on sensitive electronics from transient fires is modeled in the FPRA; as appropriate, refer to the draft FAQ under development on sensitive electronics. If this impact was not considered, provide a sensitivity study that estimates this impact on core damage frequency (CDF), large early release frequency (LERF), llCDF, and llLERF.

f) F&O 1-26 against HR-G1 (Cat 1), FSS-B2 (Cat II), and HRA-C1 (Cat II):

Describe how the Human Reliability Analysis (HRA) was performed for alternate shutdown following MCR abandonment. Include in this description:

i. Identification of events or conditions that prompt the decision to transfer command-and-control from the MCR to the alternate shutdown station. Clarify how the loss-of-control due to fires in the MCR or Cable Spreading Room was modeled.

ii. Explanation of how timing was established (i.e., total time available, time until a cue is reached, manipulation time, and time for decision-making) and which fire or fires were used as the basis for the timing. Include in the explanation the basis for any assumptions made about timing.

iii. Discussion of how different core damage end-states defined by the Abandonment HRA Event Trees presented in Attachment 10 of BNP-PSA-084 were incorporated into the FPRA, given that some sequences resulted in early and others resulted in late core damage.

iv. Description of how the feasibility of the operator actions supporting alternate shutdown was assessed.

v. Justification for assuming that continuous communication and coordination will occur during implementation of OASSD-02 by the different operators at their different locations. Include consideration of actions that require taking off headsets or the unavailability of phone systems.

vi. Description of how the impact of complexity on coordination of actions and operator performance in OASSD-01 and OASSD-02 was addressed.

vii. Description of the treatment of potential dependencies between individual actions, including discussion of operator actions that can impact the actions of other operators.

g) F&O 1-30 against FSS-A 1 (Not met):

Describe the approach and assumptions used to model fires in open and closed cabinets, and the sensitivity study on motor control centers (MCCs) presented in Section 4.8.3.1 of the LAR. Include in this description:

- 14

i. Confirmation that walkdowns were performed to determine open and closed cabinets.

ii. Given an MCC cubicle fire, identification of the cubicles in the MCC assumed to fail.

iii. Explanation of why the sensitivity study shows no impact on Unit 1 LERF and llLERF, and Unit 2 CDF, llCDF, LERF, and llLERF, while showing an increase in Unit 1 CDF and llCDF.

h) F&O 1-32 against FSS-C1 (Cat 1):

The disposition to this F&O states, based on footnotes to NUREG/CR-6850 Table G-1, that for the 98th percentile case, an HRR associated with motor fires (69 kW) was used for pump electrical fires rather than the pump electrical HRR of 211 kW that is recommended by NUREG/CR-6850, Table G-1. Provide a sensitivity study that shows that impact on CDF, LERF, llCDF, and llLERF of using the NUREG/CR-6850 recommended HRR of 211 kW as the 98th percentile HRR for pump electrical fires.

i) F&O 1-38 against LE-G2 (Not Met), LE-F3 (Not Met), UNC-A1 (Not Met), FQ-E1 (Not Met), FQ-F1 (Not Met) combined with F&O 4-18 against QU-E3 (Cat I), QU-A3, UNC-A4 (not Met), and FQ-A4 (Cat 11111111):

Explain how parametric data uncertainty was propagated and the state of knowledge correlation (SOKC) was evaluated for fire CDF and LERF. Identify fire-PRA-specific parameters (e.g., hot short probabilities, fire frequencies) that can appear in FPRA cutsets and how they were correlated. NUREG-1855 states that all basic events (regardless of system) must be correlated if their failure rates for a given failure mode are derived from the same data set. Therefore, if SOKC was applied only to basic events within the same system, provide a justification.

j) F&O 2-2 against CS-A 1 (Cat 11111111), CS-A3 (Cat 1/111111), CS-C1 (Not Met):

Document BNP-PSA-085 provides a description of component selection methodology and refers to cable selection methods, but provides no description of the cable selection and location methodology. Describe the cable selection methodology and identify where the methodology for the cable selection and location is documented.

k) F&O 2-14 against FSS-D7 (Cat 1):

Clarify whether information from the System Health Reporting and System Notebook processes, or other sources, shows data for more than 1 year to confirm that the Fire Detection and Suppression Systems have not experienced "outlier behavior." If only 1 year of data was used, justify why this is sufficient.

I) F&O 2-16 against FSS-D9 (Cat 1):

Provide additional justification for not postulating smoke damage. Address in this justification the specific types of components vulnerable to smoke damage and the potential damage mechanisms presented in Appendix T of NUREG 6850.

- 15 m) F&O 4-13 against FSS-D3 (Cat 1):

Capability Category II of Supporting Requirement, FSS-D3, as clarified by RG 1.200, Rev. 2, requires accurate characterization of significant contributors to fire risk. Further justify how this requirement is met, and identify the criteria used to determine which fire scenarios should be modeled in more detail. Also include identification and justification of physical analysis units and scenarios where fire modeling remains bounding rather than realistic.

n) F&O 4-14 against FSS-E3 (Cat I), FSS-H5 (Cat I), FSS-H9 (Cat 11111111), UNC-A2 (Cat 1/11/111):

Explain how uncertainty was treated with respect to CDF, LERF, b.CDF, and b.LERF.

Clarify the extent to which statistical quantification of uncertainty was used to evaluate fire CDF, LERF, b.CDF, and b.LERF. Identify significant fire scenarios where uncertainty was characterized qualitatively. For these scenarios, explain (per Supporting Requirement QU-E4) how the FPRA is affected by these sources of uncertainty.

0) F&O 5-13 against QU-D2 (Not Met), QU-F3 (Cat I), FQ-E1 (Not Met), and FQ-F1 (Not Met):

The disposition to this F&O indicates that a review was performed on FPRA modeling to confirm that no inconsistencies were created between sequence and system modeling, or between the FPRA and how the plant is operated. This discussion of this review is not apparent in the cited documentation (BNP-PSA-085). Describe this review and its conclusions, and identify where it is documented.

p) F&O 5-15 against QU-F2 (Cat 1/11/111), QU-F3 (Cat I), QU-D6 (Cat I), QU-D7 (Not Met),

FQ-E1 (Not Met), and FQ-F1 (Not Met) combined with F&O 5-16 against LE-F1 (Not Met), LE-F2 (Cat I), LE-G3 (Not Met), UNC-A1 (Not Met), FQ-E1 (Not Met), and FQ-F1 (Not Met):

Describe the assessment performed to determine the significant risk contributors and risk importance events and failures for CDF and LERF. Clarify how the insights from importance analysis were used to review the correctness and reasonableness of the FPRA modeling.

q) F&O 5-18 against LE-G2 (Not Met), LE-F3 (Not Met), LE-G4 (Not Met), UNC-A1 (Not Met), UNC-A2 (Cat 1111/111), FQ-E1 (Not Met), and FQ-F1 (Not Met):

These F&Os note that uncertainty and importance analysis was not performed for fire LERF. Describe the sources of uncertainty and results of im portance analyses of fire LERF.

- 16 r) F&O 6-1 against CS-B1 (Cat II) and CS-C4 (Not Met):

It is unclear from the documentation whether the breaker coordination studies for Brunswick Units 1 and 2 are complete. Section 3.3.1.7 of the LAR states that "short circuit and coordination calculations shall be updated as necessary" and it is noted that there are several breaker coordination change packages, and revised packages documented in BNP-PSA-080. Attachment 36 of BNP-PSA-080 states that three raceways could not be routed. In light of these observations:

L Clarify how the breaker coordination study assessed the three raceways that could not be routed, given that breaker coordination is assessed based on length of cable.

iL Clarify that all panels modeled in the FPRA have been evaluated and whether the breaker coordination study is complete.

PRA RAI 02 - Use of Unreviewed Analysis Methods (UAMs)

Other than the UAM identified in Section 4.8.3.1, were any other UAMs or deviations from NUREG/CR-68S0 used? If so, identify and describe those methods and clarify whether guidance from the June 21, 2012, memo from Joseph Giitter to Biff Bradley was used in applying those methods ("Recent FPRA Methods review Panel Decisions and EPRI 1022993,

'Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires"'). For identified deviations from NUREG/CR-68S0 that fall outside this guidance memo, provide a sensitivity study that estimates the impact of their removal on CDF, LERF, ~CDF, and ~LERF.

PRA RAI 03 - Transient Fire Frequency Calculation NUREG/CR-68S0 Section 6 and FAQ 12-0064 describe the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance:

a) Clarify that the methodology used to calculate hot work and transient fire frequencies applies influencing factors using NUREG/CR-68S0 guidance or FAQ 12-0064 guidance.

b) Clarify whether administrative controls are used to reduce transient fire frequency, and if so, describe and justify these controls.

c) Clarify the basis for assigning an influencing factor of "0" to Maintenance, Occupancy, or Storage for fire compartments FC296 and FC346 (Reactor Building Main Steam Isolation Valve Pit), FC30S (Reactor Building Control Rod Drive Repair Room), and FC3S6 (Reactor Building Skimmer Surge Tank Room Vault).

d) Given that a weighting factor of "SO" was not used in any fire area, provide a sensitivity study that assigns weighting factors of "SO" per the guidance in FAQ 12-0064.

PRA RAI 04- Transient Fire Placement at Pinch Points Per NUREG/CR-68S0 Section 11.S.1.6, transient fires should at a minimum be placed in locations within the plant physical analysis units (PAUs) where conditional core damage probabilities are highest for that PAU (Le., at "pinch points"). Pinch points include locations of redundant trains or the vicinity of other potentially risk-relevant equipment, including the cabling

- 17 associated with each. Transient fires should be placed at all appropriate locations in a PAU where they can threaten pinch points. Hot work should be assumed to occur in locations where hot work is a possibility, even if improbable, keeping in mind the same philosophy. Describe how transient and hot work fires are distributed within the PAUs at your plant. In particular, identify the criteria for your plant used to determine where an ignition source is placed within the PAUs. Also, if there are areas within a PAU where no transient or hot work fires are postulated because those areas are considered inaccessible, describe the criteria used to define "inaccessible." Note that an inaccessible area is not the same as a location where placement of a transient is simply unlikely. If there are "inaccessible" locations where hot work or transient fires are improbable and these locations are pinch points, provide a sensitivity study to determine the possible risk increase reflecting the possible size and frequency of fires in these locations.

PRA RAI 05 - Use of Incipient Detection in the MCR The sensitivity study presented in Section 4.8.3.6 of the LAR removes credit for incipient detection, also known as, the Very Early Warning Fire Detection System that will be installed in the MCR main control boards (MCBs). Explain why the sensitivity study results indicate no change (Le., 0%) in LlCDF but relatively significant change (i.e., +48%) in LlLERF.

PRA RAI 06 - MCR Fire Modeling of BNP-PSA-080 states that an MCR fire that does not result in a manual or automatic shutdown and is "contained" would be treated as a "non-event" by the FPRA. Explain how MCB and cabinet fires in the MCR, including the "back-panel" area, were modeled. Include in this explanation:

a) Discussion of how MCB or cabinet fire propagation was considered and which cabinet fires were considered "contained" b) Discussion and basis of placement of transient fires including how open-back panels were considered c) Clarification of credit taken for ionization smoke detectors mentioned in Attachment 6 PRA RAI 07 - Fire Induced Instrument Failure Fire-induced instrument failure should be addressed in the HRA per NUREG/CR-6850 and NUREG-1921. Describe how fire-induced instrumentfailure (including no readings, off-scale readings, and incorrect/misleading readings) is addressed in the fire HRA. Include discussion of instrumentation that was modeled explicitly in the fault trees, the success criteria assumed for this modeling, and how explicit modeling of instrumentation was done in the evaluation of Human Error Probabilities.

PRA RAI 08 - FPRA Modeling of HVAC Describe how heating, ventilation, and air conditioning (HVAC) modeling was performed to support the FPRA and whether HVAC cable tracing and fire modeling were performed to support this modeling. Confirm that additional operator actions are not needed for crediting HVAC. Heat load calculations performed for the internal events probabilistic risk assessment

- 18 (PRA) do not account for the additional heat load from fires. Confirm that heat loads from fires do not fail additional equipment in rooms that do not credit HVAC.

PRA RAI 09 - Wrapped or Embedded Cables Identify if any VFDRs in the LAR involved performance-based evaluations of wrapped or embedded cables. If applicable, describe how wrapped or embedded cables were modeled in the FPRA, including assumptions and insights on how these cables contribute to the VFDR delta-risk evaluations.

PRA RAI10 - Bases for Total Reported Plant CDF and LERF Attachment W of the LAR presents the total CDF and LERF for Units 1 and 2 and specifies the CDF from each of the following contributors: "Internal Events (including internal flooding),"

"External Flood," "High Wind," "Seismic," and "Fire." The seismic CDF (6.2 E-8/yrfor Unit 1 and 6.5E-8/yr for Unit 2) used in this estimate is low compared to the seismic CDF estimate (1.5E-5/yr) presented in a memorandum from NRC staff dated September 2010 providing updated results for Generic Issue 199 (memo titled: Safety/Risk Assessment Results for Generic Issue 199, Implication for Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United states on Existing Plants"). Also, the CDF provided for internal events (7.9E-6/yr) is much lower than the internal events CDF (4.2E-5/yr) reported in NUREG-1437, Supplement 25, dated 2006, for the BSEP license renewal environmental report. Identify the bases for the internal and seismic event CDFs and LERFs presented in the LAR, and justify the adequacy of these risk estimates for this application.

PRA RAI 11 - Risk of MCR Abandonment 6 of BNP-PSA-080 describes how the risk of MCR abandonment was calculated for fire in Fire Area CB-23E. Address the following:

a) No transient fire scenarios were postulated in the region of the MCR where operators manipulate controls, either for loss-of-control or for abandonment. The guidance in NUREG/CR-6850 is to evaluate transient fires in the MCR, including its potential contribution to abandonment. Please perform this evaluation and provide the results.

One approach is to provide a sensitivity analysis that assesses the impact of postulated transient fires in the MCR on CDF, LERF, L'lCDF, and L'lLERF.

b) The abandonment risk is highly sensitive to whether the MCR electrical cabinets are assumed to be single-bundle cables or multiple-bundle cables. Provide justification for the assumption that the MCR cabinets only contain single-bundle cables. If cabinets containing multiple-bundle cables are present in the MCR, provide the results of a sensitivity analysis accounting for the MCR cabinets that contain multiple-bundle cables.

PRA RAI 12 - Calculation of VFDR ACDF and ALERF Attachment W of the LAR provides the L'lCDF and L'lLERF for the VFDRs and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe how L'lCDF and L'lLERF or the additional risk of recovery actions were calculated. Describe the method(s) used

- 19 to determine the changes in risk reported in the Tables in Appendix W. The description should include:

a) A description of how the reported changes in risk were calculated. Include in this description any exceptions to the normal modeling mechanisms such as cases where not enough resolution exists in the PRA to model the VFDR. Also, clarify whether FAQ 08-0054 guidance was used, and describe the use of any data or methods that were not included in the FPRA Peer Review.

b) A separate description specific to how the llCDF and llLERF and additional risk of recovery actions were calculated for the MCR (Fire Area CB-23E). Include in the description how this calculation was performed for loss-of-control scenarios and for MCR abandonment scenarios (i.e., alternate shutdown).

PRA RAI13 - Scenario Results Asymmetry between Unit 1 and 2 Attachment W of the LAR presents fire scenario results for the top contributors. These results indicate an asymmetry of the CDF and LERF results between Unit 1 and 2 (e.g.,

FC210_4525_BFM, FC213_4522_B75, FC230_4801_B75, FC230_4801_B98, FC213_4621_B75, FC230_4718_B75, FC213_4617_B75, FC230_4731_B75, FC230_4811_B75, FC212_4608_B75, FC212_4607 _B75, FC210_4521_BFM). Explain the reason for this asymmetry of seemingly parallel scenarios for the two units. Also explain the asymmetries between MCR results for Unit 1 and 2.

PRA RAI14 - Table W Results for Fire Area CB-23E Explain how the additional risk of recovery actions was determined for abandonment scenarios.

PRA RAI 15 - Table W-4-1 & 2 Table Inconsistencies There appear to be a number of inconsistencies in Tables W-4-1 and 2 of the LAR Supplement.

Clarify the following:

a) Why "N/A" is reported in the additional risk of recovery actions column for fire areas where Recovery Actions are indicated (i.e., RB2-1, SW1-1, and TB-1).

b) Why a "below truncation" value is reported in the llCDF/LERF column for deterministic fire areas (i.e., AOG-1, CB-7, CB-8, DG-3, DG-4, DG-6, DG-10, DG-19, DG-20, DG-21 ,

DG-22, ISB, MWT-1, RB1-6, RB2-6, RMCSB, RPDC1, RPDC-2, RW-1, SERV, STORES, and STORM), as opposed to indicating "N/A."

c) Why a zero value is reported in the llCDF/LERF column for fire areas with VFDRs (i.e.,

DG-13, DG-14, DUCTBANK, TB1, and Yard).

PRA RAI 16 - Implementation Item Impact on Risk Estimates Identify any plant modifications (implementation items) in Attachment S of the LAR that have not been completed but that have been credited directly or indirectly in the change-in-risk estimates provided in Attachment W. When the effect of a plant modification has been included in the PRA before the modification has been completed, the models and values used in the PRA are necessarily estimates based on current plans. The as-built facility after the modification is

- 20 completed may be different than the plans. Please add an implementation item that verifies the validity of the report change in risk subsequent to completion of all PRA-credited implementation items. This item should include your plan of action should the as-built change-in-risk exceed the estimates reported in the LAR.

PRA RAI17- Model Changes and Focused Scope Reviews Since Full Peer Review American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS)

RA-Sa-2009 describes when changes to a PRA should be characterized as a "PRA upgrade.>>

Identify any such changes made to the internal events or FPRA subsequent to your most recent full-scope peer review. Also, address the following:

a} If any changes are characterized as a PRA upgrade, indicate if a focused-scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, and describe any findings and their resolution.

b) If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, describe what actions will be implemented to comply with the ASME/ANS standard.

PRA RAI 18 - Internal Events PRA F&Os Please clarify the following dispositions to internal events PRA F&Os identified in Attachment U of the LAR that have the potential to impact the FPRA results and do not appear to be fully resolved:

a} F&O 6-8 against SC-C2 (Cat III III II):

Identify software codes other than MAAP that were used to establish success criteria (e.g., GOTHIC), and describe any limitations of these codes to support success criteria used in the PRA.

b) F&O 4-5 against SY-A13 (Cat 11111111):

This F&O states that failure of feedwater check valve F032A or F0328 can lead to flow diversion that defeats HPCI or RCIC. The disposition to this F&O argues that these valves each have a failure probability two orders of magnitude lower than other HPCI or RCIC failures and therefore do not need to be modeled. Given that check valves fail at approximately 2E-4/demand, it is not clear why these failures can be dismissed per guidance in SR SY-A15. Provide further justification for dismissing these failures c) F&O 3-3 against HR-E3 (Cat I): F&O 3-4 against HR-E4 (Cat I):

Annex E4 of BNP-PSA-034 (Human Reliability Analysis) presents an "Operator Interview Worksheet" form and an "engineering review," but no operator interview results.

Describe how and where interviews with plant operators and training staff for the purpose of confirming procedure interpretation in support of the PRA modeling are documented. Likewise, describe where and how talk-throughs with plant operators or

- 21 simulator observations for the purpose of confirming the response models for the scenarios modeled in the PRA are documented. If these interview or talk-throughs do not exist as part of the FPRA documentation, provide the interview and talk-through results.

d) F&O 2-3 against HR-12 (Cat 111111):

Describe the Human Failure Events (HFEs) screening process. Explain how HFEs that were screened out of the internal events PRA but could impact FPRA results were evaluated.

e) F&O 2-2 against DA-C8 (Cat I):

The F&O states that plant specific data concerning standby time is not collected and used in the PRA. Explain how the requirement to determine component standby time (Le., DA-C8) using operational records is met. Alternatively, justify why meeting this requirement at Capability Category II is not needed.

f) F&O 6-12 against LE-G5 (Not Met):

It is not clear what was done to resolve this F&O. Characterization of LERF uncertainty is presented in BNP-PSA-075, but limitations in the LERF analysis do not appear to be provided in this document or elsewhere. Clarify what the specific limitations in the LERF analysis are for this application.

g) F&O 1-22 against IFSO-A4 thru IFQU-B2 (Many SRs are Not Met):

For nearly all internal flooding findings presented in Attachment U of the LAR, the dispositions state that internal flooding can have no impact on the FPRA. A number of scenarios listed in Tables W-2-1 and W-2-2 of the LAR supplement result in LOCAs. In general, spurious actuations have the potential to cause internal flooding. Clarify whether any fire event can result in internal flooding. If flooding can occur as a result of a fire event, then further justify why these F&Os and other internal flooding F&Os can have no impact on fire CDF, LERF, 8CDF, and 8LERF.

h) F&O 6-16 against IFSN-A6 (not Met) and F&O 1-33 against IFQU-A9 (Not Met)

Since spurious actuations have the potential to cause spray effects, clarify whether any fire event can result in spray effects impacting components modeled in the PRA. If so, justify why these F&Os can have no impact on fire CDF, LERF, 8CDF, and 8LERF.

Programmatic Question 01 Describe the specific documents that will comprise the post transition NFPA 805 fire protection program design basis.

- 22 Describe whether documents, analyses, designs, and engineering reviews prepared to support the NFPA 805 fire protection program are managed as controlled documents under the document control process.

Programmatic Question 02 Describe how the training program will be revised to support the NFPA 805 change evaluation process, including the training by plant position and how the training will be implemented (e.g., classroom, computer-based, reading program).

Programmatic Question 03 Describe how the various configuration control and change control procedures are implemented together to ensure compliance with NFPA 805 change evaluation and configuration control requirements.

Programmatic Question 04 Describe how the combustibles loading program will be administered to ensure that the Fire Probabilistic Risk Assessment assumptions regarding combustibles loading are met.

Programmatic Question 05 LAR Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 80S," does not indicate whether future NFPA 805 analyses will be conducted in accordance with the requirements of NFPA 80S, Section 2.7.3. Indicate whether future NFPA 805 analysis will be conducted in accordance with NFPA 80S, Section 2.7.3.

Programmatic Question 06 NEI 04-02 Section 4.6 indicates that the LAR should contain a "discussion of the changes to Updated Final Safety Analysis Report (UFSAR) necessitated by the license amendment and a statement that the changes will be made in accordance with 10 CFR 50.71 (e). n LAR Section 5.4 indicates that after approval of the LAR, the UFSAR will be revised consistent with NEI-04-02, however, there is no description of the changes that need to be made to the current UFSAR. Describe the changes that will to be made to the current UFSAR as a result of implementing NFPA 805. Alternatively, indicate whether the UFSAR will be updated following the guidance provided in Frequently Asked Question (FAQ) 12-0062 (ADAMS Accession No. ML121980557).

Programmatic Question 07 Describe how the plant specific requirements and configuration are incorporated when corporate or fleet wide procedures are implemented at the Brunswick plant.

- 23 Fire Modeling RAI 01 NFPA 805, Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction] ..." The NRC staff noted that fire modeling comprised the following:

The Generic Fire Modeling Treatments (GFMTs) approach was used to determine the ZOI for transient and oil spill fires in all fire areas throughout plant Fire Dynamics Tools (FDT's) were used for ZOI calculations of cabinet and cable tray fires throughout the plant The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to calculate control room abandonment times Fire Dynamics Simulator used for various fire hazard calculations Section 4.5.1.2, "FPRA" of the LAR states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V [verification and validation]," for a discussion of the acceptability of the fire models that were used.

Specifically regarding the acceptability of CFAST for the MCR abandonment times study:

a. Attachment 13 to BNP-PSA-083, Revision 2 presents a compilation of fire brigade response times from drills performed between 2002 and 2010. Of the 56 drills for which fire brigade response time data are given, 5 are for the Control Building. The response times for these drills were 20, 21, 25, 19, and 17 minutes. On page 18 of BNP-PSA-083, it is stated that the drill times are reduced by a factor of two. During the audit, it was discussed that the reduced drill times were used as the basis for the assumption in the MCR abandonment times study that the fire brigade is expected to arrive within 15 minutes. Describe the uncertainty associated with the 15-minute assumption, discuss possible adverse effects of not meeting this assumption on the results of the FPRA and explain how possible adverse effects will be mitigated.

b. Page 36 of the modeled domain section of the MCR abandonment times study, Revision 2, states "In spaces where the compartment height varies with position, the maximum height is assumed since this maximizes the entrainment." This assumption may not be conservative because, everything else (HRR, floor area, ventilation, etc.) being the same, the hot gas layer (HGL) generally will descend faster when the ceiling height is lower. Provide justification for the use of the maximum height in the CFAST analysis.

c. The sensitivity study in Appendix B of the MCR abandonment times study shows that poorly ventilated burning conditions result in a significant reduction of the MCR abandonment times in some scenarios. For instance, according to Table B-3, for a scenario involving a closed cabinet with multiple cable bundles and normal ventilation, poorly ventilated burning conditions result in a reduction of the abandonment time from 9.41 to 5.85 minutes. Explain how these abandonment time reductions affect the CDF, ~CDF, LERF and ~LERF; or provide justification for why these scenarios were not included in the FPRA calculations.

- 24 Specifically regarding the acceptability of the GFMTs approach:

d. Explain how the modification to the critical heat flux for a target that is immersed in a thermal plume described in Section 2.4 of the GFMTs document was used in the lOI determination.

Regarding the acceptability of the PSA approach, methods, and data in general:

e. Explain how the effect of the increased HRR from intervening combustibles (cable trays) on the lOI was accounted for, or provide justification for ignoring this effect.

f. Explain how wall and corner effects in the HGL calculations were accounted for, or provide a justification if these effects were not considered.

g. The FPRA Walkdown Instructions indicate that generally a 3' x 3' footprint was assumed for transient combustibles, and that the vertical lOI was measured from the floor (see page 8 of FPIP-200, Rev. 8). Actual transient fires may have a smaller area and their base may be elevated above the floor. Provide justification for the transient fire areas and elevations that were assumed during the walkdowns.

Explain how deviations from these assumptions (Le., smaller actual transient fire area and/or higher transient fire base elevation) affect the risk (CDF, ilCDF, LERF and ilLERF).

h. Address how it was assured that cables not credited in the PRA and non-target and non-cable intervening combustibles were not missed in all areas of the plant.

Provide information on how intervening combustibles were identified and accounted for in the fire modeling analyses.

i. Attachment 7 to BNP-PSA-080, Revision 2, discusses the mUlti-compartment analysis (MCA). Section 1.1.2 discusses some of the underlying assumptions in the MCA, which include that (1) an open surface area of approximately 9 fe is a general rule of thumb for the minimum area between compartments to transmit a HGL, and (2) the zone of influence is defined as approximately 5 ft vertical and 2 ft horizontal.

The licensee stated that this is appropriate, based on discussions with industry experts, which concluded that "the heat diffusion in the adjacent room would limit the HGL to a local area around the failed barrier." Provide additional information about how these two sets of criteria were specifically used in the MeA.

Fire Modeling RAI 02 NFPA 805, Section 2.5, requires damage thresholds be established to support the performance-based approach. Thermal impact(s) must be considered in determining the potential for thermal damage of structures, systems, or components. Appropriate temperature and critical heat flux criteria must be used in the analysis.

Section 3.1.1.b of the HGL Calculation (BNP-MECH-HGL-001, Rev. 1), states that "BNP predominantly has thermoset cables so the damage criteria associated with thermoset cables has been used in this analysis."

- 25 Provide the following information:

a. Describe how installed cabling in the power block was characterized, specifically with regard to the critical damage threshold temperatures and critical heat flux for thermoset and thermoplastic cables as described in NUREG/CR-6850. If thermoplastic cables are present, explain how raceways with a mixture of thermoset and thermoplastic cables were treated in terms of damage thresholds.

b. Section 2.0 of the GFMTs document provides a discussion of damage criteria for different types of targets. Section 2.1 of the GFMTs document states: "Damage to IEEE [Institute of Electrical and Electronics Engineers] 383 qualified cables is quantified as either an imposed incident heat flux of 11.4 kW/m2 (1 Btu/s-ff) or an immersion temperature of 329°C (625 OF) per Nuclear Regulatory Guidance

[NRC, 2005, NUREG 6850,2005]." Section 2.2 of the GFMTs document states:

"Damage to non-IEEE-383 qualified cables is quantified as either an imposed incident heat flux of 5.7 kW/m2 (0.5 Btu/s-ft2) or an immersion temperature of 204°C (400 OF) per Nuclear Regulatory Guidance [NRC, 2005, NUREG 6850, 2005]."

The above statements imply that in the GFMTs document, IEEE-383 qualified cables are assumed to be equivalent in terms of damage thresholds to "thermoset" cables as defined in Table 8-2 of NUREG/CR-6850. In addition, non-IEEE-383 qualified cables are assumed to be equivalent to "thermoplastic" cables as defined in Table 8-2 of NUREG/CR 6850. These assumptions mayor may not be correct. An IEEE-383 qualified cable mayor may not meet the criteria for a "thermoset cable" as defined in NUREG/CR-6850. It is also possible that a non-IEEE-383 qualified cable actually meets the NUREG/CR-6850 criteria for a "thermoset" cable.

For those areas that are assumed to have thermoset damage criteria, confirm that the cables are actually thermoset and that the potential confusion about IEEE-383/thermoset is not applicable.

c. Explain how the damage thresholds for non-cable components (i.e., pumps, valves, electrical cabinets, etc.) were determined. Identify any non-cable components that were assigned damage thresholds different from those for thermoset and thermoplastic cables, and provide a technical justification for these damage thresholds.

d. Describe the damage criteria that were used for exposed temperature-sensitive equipment. Explain how temperature-sensitive equipment inside an enclosure was treated, and provide a technical justification for these damage criteria.

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

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

26 Furthermore Section 4.7.3 "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805" of the LAR states "Calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c} were verified and validated as required by Section 2.7.3.2 of NFPA805."

Regarding the V&V of fire models:

a. Attachment J of the LAR does not provide the V&V basis for the following fire models and correlations that were used in the transition:

1. All models and correlations that were used in the development of the GFMTs.

2. Method to determine the "heat soak" time in the method of McCaffrey, Quintiere, and Harkleroad hot gas layer calculations for naturally vented compartments, described in Section 3.3.1.1 of BNP-MECH-HGL-0001, Revision 1.

3. Method to determine the "heat soak" time in Beyler's HGL calculations for closed compartments, described in Section 3.3.2.1 of BNP-MECH-HGL-0001 ,

Revision 1.

4. Method to account for the "cable thermal endurance duration" in the calculation of the thermal damage time of cables above a burning electrical cabinet, described in HNP-M/MECH-1194.

5. Method to determine the horizontal lOI described in HNP-M/MECH-1129, Revision O.

6. Method to determine the radiant energy target damage profile described in NED-M/MECH-1007, Revision O.

7. FDS, in particular the version(s) that were used in the analyses in support of the NFPA 805 transition.

Revise Attachment J to the LAR to include a discussion of the V&V basis of these models and correlations, and describe their application in the transition at BNP.

b. Provide technical details to demonstrate that fire models that were used in the NFPA 805 transition have been applied within the validated range of input parameters, or to justify the application of the model outside the validated range reported in NUREG-1824 or other V&V basis documents.

c. The discussion for the flame height calculation in Table J-1 states incorrectly that it "is used in both the CFAST and FDS." Revise the discussion to correct this error.

Fire Modeling RAI 04 NFPA 805, Section 2.7.3.3, "Limitations of Use," states: "Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method."

Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," of the Transition Report states that "Engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c} were applied appropriately as required by Section 2.7.3.3 of NFPA805."

- 27 Regarding the limitations of use:

a. The parts of the GFMTs that were used in the transition at BNP have an applicability section, which discusses the limitations of the treatments described in each part.

Explain how it was ensured that the GFMTs were used within their limits of applicability, or that uses of the GFMTs outside their limits of applicability were justified.

b. The following documents describe the lOI, HGL and MCA and related'calculations performed in support of the transition at BNP.

1. BNP-MECH-HGL-001, Rev. 1, "Hot Gas Layer Calculation"

2. BNP-PSA-080, Attachment 7, Rev. 1, "Multi-Compartment Analysis"

3. BNP-PSA-080, Attachment 19, Rev. 1, "Cable Tray Fire Propagation"

4. HNP-M/MECH-1129, Rev. 0, "Fire lone of Influence Calculation"

5. HNP-M/MECH-1194, Rev. 0, "Thermal Damage Time of Cables above a Burning Electrical Cabinet"

6. NED-M/MECH-1006, Rev. 0, "Generic Fire Modeling Treatments"

7. NED-M/MECH-1007, Rev. 0, "Radiant Energy Target Damage Profile" Each of these documents has a section that describes the assumptions and limitations of the calculations presented in the document. Explain how it was ensured that the results of these calculations were used within their limits of applicability, or that uses of these results outside their limits of applicability were justified.

c. Demonstrate that FDS was used within its scope, assumptions and limitations to evaluate the effect of tire-induced flow in an MCC (see HNP-M/MECH-1207, "Fire Induced Flow within a Motor Control Center," Rev. 0).

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

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

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

Duril1g the transition to 10 CFR 50.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g., fire modeling) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805 Section 2.7.3.4.

- 28 Post-transition, for personnel performing fire modeling or FPRA development and evaluation, Carolina Power & Light has developed and maintains qualification requirements for individuals assigned various tasks. Position-Specific Guides have been developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work. The following Training Guides have been developed and implemented.

ESG0089N - Fire Probabilistic Safety Assessment Engineer (Quantification)

ESG0093N - Fire Probabilistic Safety Assessment Engineer (Initial Development)

ESG0094N - Fire Probabilistic Safety Assessment Engineer (Data Development), and ESG0105N - Basic Fire Modeling Regarding qualifications of users of engineering analyses and numerical models:

a. Describe what constitutes the appropriate qualifications for the plant and corporate staff and consulting engineers to use and apply the methods and fire modeling tools included in the engineering analyses and numerical models.

b. Describe the process/procedures for ensuring the adequacy of the appropriate qualifications of the engineers/personnel performing the fire analyses and modeling activities.

c. Explain the communication process between the fire modeling analysts and PRA personnel to exchange the necessary information and any measures taken to assure the fire modeling was performed adequately and will continue to be performed adequately during post-transition.

Fire Modeling RAI 06 NFPA 805, Section 2.7.3.5, "Uncertainty Analysis," states: "An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," of the LAR states that "Uncertainty analyses were performed as required by 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and FPRA development."

Regarding the uncertainty analysis for fire modeling:

a. Describe how the uncertainty associated with the fire model input parameters was accounted for in the fire modeling analyses.

b. Describe how the "model" uncertainty was accounted for in the fire modeling analyses.

c. Describe how the "completeness" uncertainty was accounted for in the fire modeling analyses.

- 29 Radioactive Release Question 01 For areas where containment/confinement is relied upon, provide the qualitative/quantitative assessment.

a. For Liquids:

1) Identify where the capacities of sumps, tanks, transfer pumps, etc., is provided.

2) Identify any operator actions (e.g., to direct effluent flow/overflow with temporary measures (drain covers, etc.))

3) Identify if any of the sumps being relied upon, have auto pump out features (an automatic discharge/release at a certain sump level).

4) Identify if there are any plant features that may divert the effluent flow that were not taken into account (e.g., Auxiliary Building roll-up doors).

b. For Gaseous

1) Identify where filtering and monitoring of confined gaseous (smoke) effluent is addressed.

2) Identify any operator actions (e.g., "manual" ventilating fire areas to other ventilated areas)

3) Identify if there are plant features that can bypass the planned filtered/monitored ventilation pathway that have not been accounted for.

Radioactive Release Question 02 For areas where containment/confinement is not available, provide the quantitative assessment (liquid and/or gaseous as appropriate). Identify whether the assessment credits operator actions.

Radioactive Release Question 03 Indicate whether any of the operator actions identified in the assessments are addressed in the fire pre-plans and fire brigade training materials. Provide examples.

M. Annacone -2 gO-day Responses Section title Question Number(s)

Radiation Release 1,2,3 Fire Protection Engineering 3,4,5,6,7,8,9,10, 11,12, 13, 14, 15, 16, 17, 18, 19, 20,21 Safe Shutdown Analysis 1,2,5,9,11,13,14 Probabilistic Risk Assessment 1J, 1M, 2, 3, 6, 7, 11, 12, 13, 14, 15, 16 Fire Modeling 1B, 2C, 5, 5C 120-day Responses Section title Question Number(s}

Fire Protection Engineering 2 Safe Shutdown Analysis 15 I Probabilistic Risk Assessment 1 1E, 1L, 8 Fire Modeling 1C, 1D, 11, 2, 2D, 3,4,6 The NRC staffs information request is enclosed. Please note that review efforts on this task are continuing and additional RAls may be forthcoming.

If you have any questions regarding this letter, please feel free to contact me at (301) 415-1055.

Sincerely, IRA!

Christopher Graton, Senior Project Manager Plant Licensing Branch 11-2 Division of Operating Reactor licensing Office of Nuclear Reactor Regulation Docket Nos. 50-325 and 50-324

Enclosure:

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