BSEP 13-0083, Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805

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

ML13220B041
Person / Time
Site:	Brunswick
Issue date:	07/31/2013
From:	Hamrick G Carolina Power & Light Co, Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
BSEP 13-0083, TAC ME9623, TAC ME9624
Download: ML13220B041 (61)

Text

George T. Hamrick Vice President ENERGY, Brunswick Nuclear Plant P.O. Box 10429 Southport, NC 28461 o: 910.457.3698 July 31, 2013 Serial: BSEP 13-0083 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Brunswick Steam Electric Plant, Unit Nos. 1 and 2 Docket Nos. 50-325 and 50-324 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association Standard 805 (NRC TAC Nos. ME9623 and ME9624)

References:

1. Letter from Michael J. Annacone (Carolina Power & Light Company) to U.S.

Nuclear Regulatory Commission (Serial: BSEP 12-0106), License Amendment Request to Adopt NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water ReactorElectric GeneratingPlants (2001 Edition), dated September 25, 2012, ADAMS Accession Number ML12285A428

2. Letter from Michael J. Annacone (Carolina Power & Light Company) to U.S.

Nuclear Regulatory Commission (Serial: BSEP 12-0140), Additional Information Supporting License Amendment Request to Adopt NFPA 805 Performance-BasedStandardfor Fire Protectionfor Light Water Reactor Electric GeneratingPlants (2001 Edition), dated December 17, 2012, ADAMS Accession Number ML12362A284

3. Letter from Christopher Gratton (USNRC) to Michael J. Annacone (Carolina Power & Light Company), Request for Additional Information Regarding Voluntary Risk Initiative NationalFire ProtectionAssociation Standard 805 (TAC Nos. ME9623 and ME9624), dated May 15, 2013, ADAMS Accession Number ML13123A231 Ladies and Gentlemen:

By letter dated September 25, 2012 (i.e., Reference 1), as supplemented by letter dated December 17, 2012 (i.e., Reference 2), Duke Energy Progress, Inc., formerly known as Carolina Power & Light Company (CP&L), submitted a license amendment request to adopt a new risk-informed performance-based (RI-PB) fire protection licensing basis for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

During the week of April 8 through 12, 2013, the NRC conducted an audit at the Brunswick Plant to support development of questions regarding the license amendment request. On May 15, 2013 (i.e., Reference 3), the NRC provided a set of requests for additional information (RAIs) regarding the license amendment request. This letter divided these RAIs into 60-day, 90-ILA-~

U.S. Nuclear Regulatory Commission Page 2 of 3 day, and 120-day responses. In subsequent telephone calls with the NRC Project Manager for BSEP, the following modifications were agreed to regarding the RAI response schedule shown in the May 15, 2013, letter:

" The 60-day RAI responses will be submitted by July 1, 2013 (i.e., 60 days following the May 2, 2013, clarification call that was conducted with the NRC). These responses were submitted by letter dated June 28, 2013. Probabilistic Risk Assessment (PRA) RAIs 1A, 1B, 1C, 1D, 1F, 1G, 11, 1K, 1N, 10, 1P, 1Q, 1R, 4, 5, 9,10, 17, and 18, which were included in the set of 60-day RAIs, will be addressed in a separate submittal due by July 15, 2013 (i.e., 60 days following the date of the letter). These responses were submitted by letter dated July 15, 2013.

• The 90-day RAI responses will be submitted by July 31, 2013 (i.e., 90 days following the May 2, 2013, clarification call). Fire Protection Engineering RAI 1, which was included in the set of 60-day RAIs, will be addressed as part of the 90-day RAI responses.

• The 120-day RAI responses will be submitted by August 30, 2013 (i.e., 120 days following the May 2, 2013, clarification call). PRA RAI 1H will be addressed as part of the 120-day RAI responses, rather than with the 60-day RAI responses.

A tabulation of the individual RAIs and the planned response submittal dates is provided in . Duke Energy's responses to the set of 90-day RAIs are provided in Enclosure 2.

Also included is the response to RAI 1K, which was deferred to the 90-day RAI responses from the 60-day PRA-related responses.

This document contains no new regulatory commitments.

Please refer any questions regarding this submittal to Mr. Lee Grzeck, Manager - Regulatory Affairs, at (910) 457-2487.

I declare, under penalty of perjury, that the foregoing is true and correct. Executed on July 31, 2013.

Sincerely, George T. Hamrick

Enclosures:

1. Revised Response Schedule to NFPA 805 Request for Additional Information

2. Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association (NFPA) Standard 805

U.S. Nuclear Regulatory Commission Page 3 of 3 WRM/wrm cc (with enclosures):

U. S. Nuclear Regulatory Commission, Region II ATTN: Mr. Victor M. McCree, Regional Administrator 245 Peachtree Center Ave, NE, Suite 1200 Atlanta, GA 30303-1257 U. S. Nuclear Regulatory Commission ATTN: Mr. Christopher Gratton (Mail Stop OWFN 8G9A) 11555 Rockville Pike Rockville, MD 20852-2738 U. S. Nuclear Regulatory Commission ATTN: Ms. Michelle P. Catts, NRC Senior Resident Inspector 8470 River Road Southport, NC 28461-8869 Chair - North Carolina Utilities Commission P.O. Box 29510 Raleigh, NC 27626-0510 Mr. W. Lee Cox, III, Section Chief (Electronic Copy Only)

Radiation Protection Section North Carolina Department of Health and Human Services 1645 Mail Service Center Raleigh, NC 27699-1645 lee.cox@dhhs.nc.gov

Enclosure 1 Page 1 of 1 Revised Response Schedule to NFPA 805 Request for Additional Information Revised Response Schedule Section Title Questio nNumber(s) Submittal Date 60-Day Response - Non-PRA Programmatic 1, 2, 3, 4, 5, 6, 7 July 1, 2013 Safe Shutdown Analysis 3, 4, 6, 7, 8, 10, 12 (Complete Fire Modeling 1A, 1E, 1F, 1G, 1H, 2A, 2B, 5A, 5B June 28, 2013) 60-Day Response - PRA Probabilistic Risk 1A, 1B, IC, 1D, 1F, 1G, 11, 1N, 10, 1P, July 15, 2013 Assessment 1 Q, 1R, 4, 5, 9, 10, 17, 18 (Complete July 15, 2013) 90 Day Response _____... _... ___.

Radiation Release 1, 2, 3 Fire Protection Engineering 1, 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 July 31, 2013 Probabilistic Risk 1J, 1K, M, 2, 3, 6, 7, 11, 12, 13, 14, 15, Assessment 16 Fire Modeling 1B, 2C, 5C 120 Day'Response __........___........_"______: _____...... _"_..

Fire Protection Engineering 2 Safe Shutdown Analysis 15 Probabilistic Risk 1E, 1H, 1L, 8 August 30, 2013 Assessment Fire Modeling 1C, 1D, 11, 2D, 3, 4, 6

Enclosure 2 Page 1 of 57 Response to Request for Additional Information Regarding Voluntary Risk Initiative National Fire Protection Association (NFPA) Standard 805 By letter dated September 25, 2012 (i.e., ADAMS Accession Number ML12285A428), as supplemented by letter dated December 17, 2012 (i.e., ADAMS Accession Number ML12362A284), Duke Energy Progress, Inc., formerly known as Carolina Power & Light Company, submitted a license amendment request (LAR) to adopt a new risk-informed performance-based (RI-PB) fire protection licensing basis for the Brunswick Steam Electric Plant (BSEP), Unit Nos. 1 and 2.

During the week of April 8 through April 12, 2013, the NRC conducted an audit at BSEP to support development of questions regarding the license amendment request. On May 15, 2013, the NRC provided a set of requests for additional information (RAIs) regarding the license amendment request. That letter divided the RAIs into 60-day, 90-day, and 120-day responses.

In subsequent telephone calls with the NRC Project Manager for BSEP, the following modifications were agreed to regarding the RAI response schedule shown in the May 15, 2013, letter:

• The 60-day RAI responses will be submitted by July 1, 2013 (i.e., 60 days following the May 2, 2013, clarification call that was conducted with the NRC). These responses were submitted by letter dated June 28, 2013. Probabilistic Risk Assessment (PRA) RAls 1A, 1B, 1C, 1D, 1F, 1G, 11, 1K, 1N, 10, 1P, 1Q, 1R, 4, 5, 9, 10, 17, and 18, which were included in the set of 60-day RAls, and were addressed in a separate submittal dated July 15, 2013 (i.e., 60 days following the date of the letter).

• The 90-day RAI responses will be submitted by July 31, 2013 (i.e., 90 days following the May 2, 2013, clarification call). Fire Protection Engineering RAI 1, which was included in the set of 60-day RAls, will be addressed as part of the 90-day RAI responses.

• The 120-day RAI responses will be submitted by August 30, 2013 (i.e., 120 days following the May 2, 2013, clarification call). PRA RAI 1 H will be addressed as part of the 120-day RAI responses, rather than with the 60-day RAI responses.

Duke Energy's 90-day response to the RAls is provided below. Also included is the response to RAI 1 K, which was deferred from the 60-day PRA-related responses to the 90-day RAI responses.

Radiation Release Requests for Additional Information Radiation Release RAI I 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).

Enclosure 2 Page 2 of 57

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

Response

a. For Liquids:

Identify where the capacities of sumps, tanks, transfer pumps, etc., is provided BSEP UFSAR Section 11.2.2 System Description contains specific system capacity details that were considered during the panel reviews described below for water runoff from firefighting in radioactive areas collected in the Radwaste system. Additionally, Drawing D-02533, Sheet I and 2, shows tark capacities/pumps for power block buildings.

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

Qualitative discussion of liquid and gaseous effluent capabilities can be found in LAR Attachment "E" and Table E-1. No operator actions specific for control of radioactive release due to fire fighting operations were specified under the evaluation. General containment of run-off and ventilation discharge of smoke and combustion by-products are addressed on a precautionary level in the fire pre-plans.

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

None were identified during the qualitative review based on, BSEP UFSAR Section 11.2.2 System Description contains specific system capacity details that were considered during the panel reviews described below for water runoff from firefighting in radioactive areas collected in the Radwaste system. Additionally, Drawing D-02533, Sheet 1 and 2, shows tank capacities/pumps for power block buildings.

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)

None were identified during the qualitative review based on, BSEP UFSAR Section 11.2.2 System Description contains specific system capacity details that were considered during the panel reviews described below for water runoff from firefighting in radioactive areas collected in the Radwaste system. Additionally, Drawing D-02533, Sheet 1 and 2, shows tank capacities/pumps for power block buildings.

Enclosure 2 Page 3 of 57

b. For Gaseous Identify where filtering and monitoring of confined -gaseous(smoke) effluent is addressed.

Qualitative discussion of liquid and gaseous effluent capabilities can be found in LAR Attachment "E" and Table E-1 Identify any operator actions (e.g., "manual" ventilating fire areas to other ventilated areas)

No operator actions specific for control of radioactive release due to firefighting operations were specified under the evaluation. General containment of run-off and ventilation discharge of smoke and combustion by-products are addressed on a precautionary level in the fire pre-plans (i.e., the Precautions and Limitations section of the procedure).

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

None were identified during the qualitative review, based on the BSEP UFSAR Section 11.3 System Description, which contains specific system capacity details that were considered during the panel reviews described above for water runoff from firefighting in radioactive areas collected in the Radwaste system.

The following is provided to clarify the review and evaluation process utilized:

For areas where containment/confinement is relied upon, the methodology for radioactive release review consisted of a qualitative review on a building and fire area-by-fire area basis utilizing the fire pre-plans as a guide and documented in the LAR Attachment "E." Specifically, for BSEP, a review was conducted by a review panel to ensure specific steps are included for containment and monitoring of potentially contaminated materials so as to limit the potential for release of radioactive materials due to firefighting operations. The review panel consisted of representatives from Operations, Engineering (i.e., Fire Protection; Heating, Ventilation, and Air Conditioning Systems), Operations Fire Brigade Training, and Radiation Protection. A review of engineering controls was performed to ensure containment of gaseous and liquid effluents (i.e.,

smoke and fire fighting agents). This review included all plant operating modes (i.e., including full power and non-power conditions).

Fire pre-plans that address fire areas where there is no possibility of radioactive materials being present were screened from further review. All other fire pre-plans were reviewed to ascertain whether existing engineering controls are adequate to ensure that radioactive materials (i.e., contamination) generated as a direct result of fire suppression activities are contained and monitored before release to unrestricted areas, such that the release would meet the NFPA 805 radioactive release performance criteria.

The review determined that existing engineering controls, such as curbs and forced air ventilation, were qualitatively adequate, based on panel member expertise and plant knowledge, to meet the NFPA 805 radioactive release requirements. In addition, each of the fire pre-plans addressing fire areas where radioactive materials may be present were updated to include general guidance for containment and monitoring of smoke and fire suppression agent runoff should the effectiveness of the installed engineering controls be challenged or impacted by fire suppression activities.

The review panel determined existing engineering controls are adequate to ensure that radioactive materials (i.e., radiation) generated as a direct result of fire suppression activities is

Enclosure 2 Page 4 of 57 contained and monitored prior to release to unrestricted areas such that such release would be as low as reasonably achievable and would not exceed applicable 10 CFR Part 20 limits.

Engineering controls, such as use of forced air ventilation and damming for fire suppression agent run off, were considered during review of fire pre-plans, for areas in which this is the anticipated response identified in the pre-fire plan. No new engineering controls were identified or established as a result of this review and all present controls are as currently in place under the approved pre-transitional fire protection program.

Radiation Release RAI 2 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.

Response

For plant fire areas identified with the potential for radioactive release, none were identified where containment/confinement was not available through the means of fire pre-plan guidance or available installed engineering controls. This was based on panel reviews conducted of the various plant areas and pre-plans as described in Section 4.4 of the LAR and BSEP Change Package BNP-0228, Revision 0.

No operator actions specific for control of radioactive release due to fire fighting operations were specified under the evaluation. General containment of run-off and ventilation discharge of smoke and combustion by-products are addressed on a precautionary level in the fire pre-plans.

Radiation Release RAI 3 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.

Response

General containment of run-off and ventilation discharge of smoke and combustion by-products are addressed on a precautionary level in the fire pre-plans. Actions included in fire pre-plans are described in LAR Attachment "E," Table E-1. An example is shown below:

OPFP-MBPA, Miscellaneous Building Pre-Fire Plans - Protected Area Rev 18, 01130112

- Modify section 2.0 to include a statement to prompt containment and monitoring of potentially contaminated fire suppression agents and products of combustion and referencing OPFP-13.

- Modify Attachments 7, 8, 15, 18, 25 to accommodate precautions for containment and monitoring of potentially contaminated fire suppression agents and products of combustion.

The fire pre-plan OPFP-MBPA was previously provided on the SharePoint site.

Enclosure 2 Page 5 of 57 Fire Protection Engineering Requests for Additional Information Fire Protection Engineering RAI I 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.

Response

The very early warning fire detection system (VEWFDS), or incipient fire detection system, is in the approved design phase. The code of record being used in the development of the VEWFDS portion of BSEP Engineering Change, EC 50724, is National Fire Protection Association (NFPA) 72, National FireAlarm Code. The appropriate sections of NFPA 76, Standardfor the Fire Protection of Telecommunications Facilities,will be used for guidance to ensure that the VEWFDS meets the performance goals for proper credit in the Fire PRA. VEWFDS equipment planned for use will be consistent with that installed at other Duke Energy Progress facilities, including the Harris pilot plant. BSEP plans to use the SAFE air sampling incipient fire detection equipment, which utilizes a multi-zone, single detector with aspiration tubing provided for one or more cabinets, as it is the most feasible design.

Installation and testing will be in conformance with the Nuclear Generation Group (NGG) Fleet Engineering Change (EC) process and will be performed to both NFPA 72 and the original equipment manufacturer (OEM) requirements. Training and qualification of installation technicians associated with the installation of VEWFDS at BSEP will be in accordance with applicable fleet and site procedures for the conduct of maintenance and construction activities.

Training requirements will be finalized and documented during the EC Process. Vendor support, when provided, will be in accordance with applicable NGG material and contract service procedures in place, and under the direction of Duke Energy. Final installation and commissioning of the system, including acceptance testing, sensitivity and setpoint control(s),

will be performed by the OEM with assistance and support of site personnel. Regular and preventative maintenance will also be in accordance with the OEM, NFPA 72, NFPA 76, and BSEP's preventative maintenance program. The VEWFDS will include continuously monitored trouble annunciation consisting of a circuit supervisory signal for faults in the detector(s) or a failure of one of the system modules. Any detector or system fault condition would be annunciated, investigated immediately, and appropriate compensatory measures implemented until the fault condition is corrected. In addition to the continuously monitored supervisory trouble indication, the VEWFDS will receive quarterly surveillance testing and annual maintenance as recommended by the OEM. The VEWFD System will be integrated into the

Enclosure 2 Page 6 of 57 BSEP ESTTM fire detection system with Main Control Room annunciations "Trouble," "Pre-Fire,"

and "Fire" (i.e., equivalent to "Trouble," "Alert," and "Alarm" conditions). Control Room Operators will respond to these alarms in accordance with plant operating procedures.

Qualified on-shift Operations and/or Maintenance personnel will respond to investigate all alarms without delay and with the same response urgency, and will ensure that there is continuous attendance of the affected area until the condition is resolved. Responding personnel will have basic training sufficient to initiate early firefighting activities (i.e., use of portable fire extinguisher equipment) such that the expectation will be satisfied that if a developing fire is discovered, there will be an immediate intervention to suppress/control the fire.

Risk reduction credited for use of in-cabinet VEWFDS associated with this EC is in accordance with 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.

Fire Protection Engineering RAI 3 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.

Response

Based on current engineering design windows and outage schedules, the work is conservatively planned to be finished by Refueling Outage 22 for BSEP Unit 2, scheduled for April of 2015, which would be before the second cycle after the expected issuance of the Safety Evaluation.

Planning for an additional cycle for each unit will allow for any unscheduled plant or equipment issues during implementation. Appropriate compensatory measures will remain in place until those modifications are in place.

Fire Protection Engineering RAI 4 Attachment A, Table B-i, 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."

Response

The referenced evaluation, Calculation 2FP-0052, Unit 2 Thermo-Lag SeparationZone Evaluation, evaluates the acceptability of abandoning, in place, Thermo-Lag fire barrier materials which are installed within the identified 20 foot separation zones of Fire Area RB2-1 that is located in the Unit 2 Reactor Building. As such, the "complies via EEEE" compliance strategy applies only to Fire Area RB2-1. All other Fire Areas in Section 3.3, Requirement 2 (i.e.,

Section 3.3(2)) fall under the "complies" compliance strategy.

Fire Protection Engineering RAI 5 Attachment A, Table B-i, 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

Enclosure 2 Page 7 of 57 "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.

Response

The referenced evaluation, Calculation OFP-0033, StructuralSteel Fireproofing,evaluated the acceptability of not fire proofing exposed structural steel which is located in the control building elevator shaft. This evaluation is limited to the following fire zones: CB-6, CB-11, and CB-22, which are located within fire area CB-23E. The evaluation concluded that the steel columns installed in the west wall of the Control Building elevator shaft have adequate fire resistance for the worst case fires expected in either the elevator shaft or the men's restroom, if left unprotected. As such, the "complies via EEEE" compliance strategy applies only to Fire Area CB-23E. All other Fire Areas fall under the "complies" compliance strategy.

A revision to Attachment A, Table B-i, Section 3.3.2, will be submitted with the 120-day RAI responses.

Fire Protection Engineering RAI 6 Attachment A, Table B-I, 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.

Response

The compliance strategy "Complies via Previous NRC Approval" applies only to the following Control Building cable accessway fire zones: CB-01A, CB-01B, CB-02A, CB-02B, CB-1 2A, CB-12B, CB-13A and CB-13B for existing electrical raceway construction details which are located in Fire Areas CB-1 and CB-2. The "Complies with Clarification" compliance statement applies to all other plant fire areas and zones relative to Specification 048-001 requirements and the guidance of FAQ 06-0021. As such, the "Complies via Previous NRC Approval" compliance strategy applies only to Fire Areas CB-1 and CB-2. All other fire areas fall under the "Complies with Clarification" compliance strategy.

A revision to LAR Attachment A, Table B-i, Section 3.3.5.2, to include this clarification, will be submitted with the 120-day RAI responses.

Fire Protection Engineering RAI 7 Attachment A, Table B-i, 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.

Enclosure 2 Page 8 of 57

Response

The compliance strategy "Complies via EEEE" is not necessary. The "Complies with clarification" (i.e., using FAQ 06-0022) and the "Complies via previous NRC Approval" compliance statements are sufficient.

A revision to LAR Attachment A, Table B-i, Section 3.3.5.3, to include this clarification, will be submitted with the 120-day RAI responses.

Fire Protection Engineering RAI 8 Attachment A, Table B-i, Section 3.3.6 of the LAR indicates compliance by "clarification" and identifies compliance with an equivalent Auxiliary and Power Conversion Systems Branch's 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, StandardMethods of Fire Tests of Roof Coverings. Provide sufficient justification regarding Class I, Factory Mutual System Approval Guide, as equivalent to Class A, NFPA 256.

Response

NFPA 256 was withdrawn at the Annual NFPA meeting in 2008 because the material is found in ASTM E 108 and UL 790. ASTM E 108 is a measure of the relative fire characteristics of roof coverings under simulated fire originating outside the building. The Fire Model (FM) Approval Standard 4471 for Class I Panel Roofs sets performance requirements for panel roofs which include all components necessary for installation of the panel roof assembly, not just the roof coverings like ASTM E 108. The FM Approval Standard also requires the assembly to exhibit low fire spread below the panel as well as above the panel, unlike ASTM E 108 which only tests for fires originating outside the building. For the combustibility test from above the roof assembly, the FM Approval Standard references the ASTM E 108 fire test as the test method, but then also places further restriction on the flame spread test over the ASTM E 108 fire test (i.e., that the flame shall not be allowed to spread to more than one 'ateral edge of the exposed panel roof beyond 12 inches of the leading edge of the test sample).

The FM Approval Standard 4471 for Class I Panel Roofs fully encompasses ASTM E 108 and places further testing requirements on the full roof assembly and is, therefore, equivalent to Class A of ASTM E 108 (i.e., formerly NFPA 256).

Fire Protection Engineering RAI 9 Attachment A, Table B-i, 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 OffGas/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).

Enclosure 2 Page 9 of 57

Response

Attachment A, Table B-I, Section 3.3.7.1, of the LAR mistakenly stated that there was bulk flammable gas stored in the Reactor Buildings, Diesel Generator Rooms, and AOG Building.

The bulk gas storage allowed in the structures listed is not flammable.

A revised LAR, Attachment A, NEI 04-02 Table B-i, "Transition of Fundamental Fire Protection Program & Design Elements," will be submitted along with the 120-day RAI responses and will include Section 3.3.7.1 noting a compliance statement of, "Complies", and the reference to the SER will be removed.

Fire Protection Engineering RAI 10 Attachment A, Table B-i, 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.

Response

Attachment A, Table B-i, Section 3.3.9, of the LAR was omitted in error. A revised LAR, Attachment A, NEI 04-02 Table B-I, "Transition of Fundamental Fire Protection Program &

Design Elements," will be submitted along with the 120-day RAI responses and will include Section 3.3.9 noting a compliance statement of "Complies with Clarification," and the basis of "See Implementation Item pertinent to NFPA 805, Chapter 3, Section 3.3.9, compliance in Attachment S of the Transition Report."

Fire Protection Engineering RAI 11 Attachment A, Table B-i, 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.

Response

The flame impingement barrier described in DBD-62 separates the diesel engine driven fire pump and its controller from the motor driven fire pump and its controller. From the modification which installed the barrier (i.e., PM 79-230), the barrier consists of a 6-inch high concrete curb to contain a possible oil spill and a 9-foot high steel plate partition. The plating, 1/8-inch thick, is coated on both sides with% inch of "thermo-lag."

Enclosure 2 Page 10 of 57 Per the Fire Protection Safety Evaluation Report for the Brunswick Plant, dated November 22, 1977:

Both fire pumps and their controllers are located in the water treatment building, and could be subject to damage by a fire in that structure. To preclude such an event, the licensee has proposed to provide automatic sprinklers, and barriers, to prevent flame impingement between the pumps and between the pumps and the controllers, and three hour fire barriers between the building and the diesels fuel tank. A flow switch and cutoff valve to detect a rupture in the supply line and shut off fuel flow to the diesel driven fire will be provided. We conclude that, subject to the implementation of the above described modifications, the fire pumps satisfy the objectives identified in Section 2.1 of this report and are, therefore, acceptable.

A revised LAR, Attachment A, NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements will be submitted along with the 120-day responses and will include Section 3.5.5 noting a compliance statement of "Complies via Previous NRC Approval."

Fire Protection Engineering RAI 12 Attachment A, Table B-i, 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.

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.

Response

Two fire hydrants were installed on the fire loop to the service water intake structure (i.e.,

Hydrants 8 and 14) as stated on the 1977 SER, Section 4.3.1(3). These hydrants are shown on plant drawing D-02043, Sheet 1. Per a letter from C. E. Murphy (NRC) to J. A. Jones (CP&L) dated September 19, 1979, the Fire Protection SER Commitment Paragraph 3.1.8 relating to the installation of the two fire hydrants on the fire loop to the service water intake structure was closed by the NRC. Of the two fire hydrants installed on the fire loop, one nearby hydrant to the service water intake structure was installed (i.e., Hydrant 14) to satisfy Section 5.7.6 of the 1977 SER. Both the NRC letter and the plant drawing D-02043, Sheet 1, are posted on the SharePoint site.

Fire Protection Engineering RAI 13 LAR Attachment A, Table B-i, 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.

Response

Attachment A, Table B-i, Section 3.6.5, of the LAR was omitted in error. A revised LAR, Attachment A, NEI 04-02 Table B-i, Transition of Fundamental Fire Protection Program &

Design Elements, will be submitted along with the 120-day RAI responses and will include

Enclosure 2 Page 11 of 57 Section 3.6.5 noting a compliance statement of "N/A" and the basis of "There are no seismic required hose stations at BSEP. See Compliance Basis for Section 3.6.4."

Fire Protection Engineering RAI 14 Table B-A, 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.

Response

The "Complies via Previous NRC Approval" compliance strategy refers to the referenced Safety Evaluation, dated May 29, 1987, which stated the following:

The BSEP acceptance criteria developed for the penetration seals contains the following three major elements:

1. The test conditions will use the standard fire exposure curve as defined in ASTM E-1 19. This is the same requirement for all three referenced criteria (NRC, ANI, and IEEE);

2. The standard hose-stream test will be conducted as specified in ASTM E-1 19.

This is a stricter test than any of the three referenced criteria in that only a solid stream is allowed by ASTM; and

3. The temperature rise criteria as defined by ANI was selected with one additional consideration. When a recorded temperature exceeds the temperature rise limit of 325 °F, the situation will be analyzed and can be dispositioned if justified. This criterion is less strict than the NRC limits, but more strict than IEEE.

3.0 CONCLUSION

S After having reviewed the penetration seal program, the staff concludes that the acceptance criteria established by the licensee as well as the various seal installation configurations, are acceptable, and are suitable deviations from BTP ASB 9.5-1. We concur with the licensee that the additional 75 OF temperature rise allowed by the ANI criteria is not considered likely to significantly add to the risk of igniting material on the unexposed side of the barrier. Therefore, there is no need to add to the administrative controls already in place with respect to the control of combustibles inside the plants.

The Complies via EEEE compliance strategy refers to the referenced list of evaluations where the seal configurations were evaluated to be acceptable for the hazards in which they are installed. These various identified penetration seal locations are evaluated in each of the referenced EEEEs. The table below identifies the applicable Fire Zones/Fire Areas which apply to each of the referenced EEEEs.

A revision to Attachment A, Table B-I, Section 3.11.4 and Table identifying the Fire Areas/Zones that each of the referenced EEEEs apply to, will be submitted with the 120-day RAI responses.

Enclosure 2 Page 12 of 57 The referenced evaluation 704U-M-22/SI title is incorrect; it should be listed as 704U-M-22.

Reference 98-00429 should be deleted, as it does not involve an evaluation of the fire resistance capabilities of penetration seals. It involves justification for removing existing seals from the inspection program.

Enclosure 2 Page 13 of 57 Evaluation Title Fire Zone Fire Area 2FP-0036 Evaluation of the Penetration Seal in the Unit 2 Reactor RB2-6 RB2-6 Building ECCS Room RB2-01n RB2-1 2FP-0050 Evaluation of Control Building Selected Penetrations CB-01b CB-1 CB-5, CB-6, CB-21 CB-23E TB2-01a TB1 RW-01 b RW-1 TB1-01b TB1 U2 TB BLDG EAST TB1 HALLWAY Fire Zone not assigned part of Fire Area TBI 704U-M-28 Evaluation of Seal Design D1 (From Spec Waiver SWB- AOG-1 AOG-1 118-003-H) When Installed From One Side of a Barrier in the AOG Building OFP-1060 Fire Resistance Rating for Penetration Seals R2-2-021 RB2-04 RB2-1 through R2-2-027 TB2-01b TB1 85-125-0-12-F Evaluation of Control Building Flush Mounted CB-20 CB-23E Junction Boxes TB2-01a TB1 85-125-0-16-F Evaluation of Junction Box Penetration Seals within a DG-02 DG-2 Floor Slab in the Diesel Generator Bldg DG-03 DG-3 DG-13 DG-13 DG-14 DG-14 85-125-0-17-F Evaluation of Reactor Building PAM Tubing RB1-01g(SNV) RB1-1 Penetrations, Unit 1 & 2 TB1-01a TB1 RB2-01g(N/W) RB2-1 TB2-01b TB1 85-125-0-18-F Evaluation of Diesel Generator Building Penetration DG-20 DG-20 Seals DG-21 DG-21 DG-22 DG-22 85-125-0-27-F Evaluation of Bus Duct Seals in the Diesel Generator DG-1 1 DG-1 1 Building DG-12 DG-12

_DG-13 DG-13

Enclosure 2 Page 14 of 57 Evaluation Title Fire Zone Fire Area DG-14 DG-14 85-125-0-32-F Evaluation of Control Building Penetrations behind Pull CB-20 CB-23E Boxes Ul TB BLDG EAST TB1 HALLWAY Fire Zone not assigned part of Fire Area TB1 85-125-0-34-F Evaluation of Penetration Seal 2-FP-R2-3-008 RB2-01h (W/C) RB2-1 U2 TB BLDG EAST TB1 HALLWAY Fire Zone not assigned part of Fire Area TB1 85-125-0-38-F Evaluation of Penetrations 1-FP-R1-4-001 and RB1-01j RB1-1 2-FP-R2-4-001 RB2-01j RB2-1 TB1-12 & TB1-13 TB1 TB2-12 & TB2-13 TB1 85-125-0-42-F Evaluation of Control Building Penetration 0-FP-CB-2-257 CB-22 and CB-19 CB-23E TB2-12 & TB2-13 TB1 85-125-0-45-F Sealing Requirements for Penetration 0-FP-CB-2-277 CB-13a CB-2 CB-23 CB-23E 85-125-0-48-F Evaluation of Penetration Seal R2-1-012 RB2-01b RB2-1 RW1-01a RW-1 PT TB1 85-125-0-53-F Evaluation of Two Conduits in the Unit 2 Reactor RB2-6 RB2-6 Building ECCS Room RB2-01g (N/C) & RB2-01g RB2-1 (N/E)

OFP-0021 Downgrade of Rattle Space Wall-to-Sleeve Link-Seals to PT TB1 Non-Fire Rated Status TB2-01a TB1 RB1-01b, RB1-01a, RB1-01d RB1-1 RB2-01g (N/W), RB2-01b RB2-1 CB-02a, CB-02b, CB-013a, CB-2 CB-013b CB-01a, CB-01b, CB-012a, CB-1 CB-012b

Enclosure 2 Page 15 of 57 Evaluation Title Fire Zone Fire Area OFP-0026 Battery Room Penetration Seals CB-07 CB-7 CB-08 CB-8 CB-05 CB-23E CB-09 CB-9 CB-10 CB-10 704U-M-26 Evaluation of the Use of Nelson CLK and RSW for 3-Hour AOG-1 AOG-1 Fire Barrier Penetration Seals for Spare Open Conduits 84-0622 Penetration Evaluation - SWIS East Wall SWI-1 SWI-1 85-125-0-08-F Diesel Generator Bldg. Pyrocreted Pull Box Enclosures DG-04 DG-4 DG-05 DG-5 85-125-0-14-F PVC Pipe Penetration SW-3-031 in Service Water Intake SWI-1 SWI-1 Structure OUTDOORS 85-125-0-21-F Evaluation of Seals CB-1-262, 263, 264, 265, 270, 271, CB-06 CB-23E 272, & 273 TB1-01a, TB1-01b, TB2-01a TB1 and TB2-01b 85-125-0-23-F Turbine Bldg-Combination Link Seal and Additional TB1-01d & TB2-01d TB1 Moisture Seals No Fire Zone/Area assigned to rattle space 85-125-0-31-F Unit 1 & 2 Turbine Bldg/Reactor Wall Thru Pipe Link Seals TB1-Old & TB2-01d TB1 RB1-01a RB1-1 RB1-01b RBI-1 RB2-02a RB2-1 RB2-02b RB2-1 85-125-0-49-F Reactor Bldg.- Eccentric Link-Seal Design RB2-01a RB2-1 No Fire Zone/Area assigned to rattle space 85-125-0-51-F Existing Link Seal Evaluations RB1-01b RBI-1 RB2-01b RB2-1 RB1-01a RB1-1 RB1-01h RBI-1 89-0149 Evaluation of Service Water Building Penetration Seal SWI-1 SWI-1 SW-3-031 I I 95-00642 Alternative Repair to Fire Barrier Penetration Seals CB-05 CB-23E

Enclosure 2 Page 16 of 57 Evaluation Title Fire Zone Fire Area CB-06 CB-23E No Fire Zone/Area assigned to rattle space 95-01461 Evaluate Fire Rating of Penetrations R2-1-009, T2-1-002 RB2-1b RB2-1 and R2-1-018 CB-02a, CB-02b, CB-13a, CB- CB-2 13b TB1-01d TBI RB2-1a RB2-1 PT TB1 98-00054 Evaluation of Silicone Foam Fire Seals Containing Copper CB-23 CB-23E Pipe Control Building Roof OUTDOORS No fire zone assigned 704U-M-22/S1 Evaluation of Conduit Penetrations in Reactor Buildings CB-la, CB-1b, CB-12a, CB- CB-1 and Control Building 12b CB-2a, CB-2b, CB-13a, CB- CB-2 13b RB1-01g(SIN & S/C), RB1- RBI-1 01h(S/W), RBI-10(S)

RB2-01g(N/W & N/C), RB2- RB2-1 01h(N/W & N/C) 85-125-0-41-F Reactor Bldg.-RHR Rooms Penetration Seals RB1-01e & RB1-01f RB1-1 RB2-01e & RB2-01f RB2-1 704U-M-31 Sealing Requirements for Penetration AO-2-057 in AOG AOG-1 AOG-1 Bldg __

85-125-0-02-F U/1 & U/2 Reactor Building ECCS Room Minimum RBI-6 RB1-1 Embedment of Hilti Kwik Bolts for Boot Seal Penetration RBI-1 RB1-1 Fire Seals 85-125-0-05-F Control Building Pull/Junction Box Fire Stop Applications TB1-01a & TB2-01 a TB1 All fire zones of CB-23E CB-23E 85-125-0-10-F Diesel Gen. Bldg. Pyrocrete Enclosure Barriers of Pipe & All DG Fire Zones All DG Fire Conduit Areas 85-125-0-11-F Diesel Generator Building Evaluation of Thermo-Lag DG-06 DG-6 Installation DG-16 DG-16E

Enclosure 2 Page 17 of 57 Evaluation Title Fire Zone Fire Area DG-04 DG-4 DG-03 DG-3 85-125-0-22-F Deviation to Design "C" of Specification # 118-003 CB-04 CB-23E CB-06 CB-23E 704U-S-03 12" Diameter Grouted Sleeved Opening, Fire Seal GENERIC ALL FIRE ZONES ALL FIRE Evaluation AREAS 85-125-0-04-F Cellular Concrete Floor & Wall Blockout Electrical GENERIC ALL FIRE ZONES ALL FIRE Penetrations AREAS 98-00429 Engineering Walkdown of Penetration Seals Recommend this EEEE be deleted from the B-1 Table.

Evaluation involves verification that various seals are included in inspection program

Enclosure 2 Page 18 of 57 Fire Protection Engineering RAI 15 Attachment A, Table B-I, Section 3.11.5 of the LAR includes Electrical Raceway Fire Barrier System (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-i, 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 Fire Probabilistic Risk Assessment (FPRA).

Response

The compliance strategy "Complies via EEEE" will be deleted.

The pyrocrete barrier evaluated under Calculation 85-125-0-10-F, Diesel generatorBuilding Pyrocrete Enclosure Barriersof Pipe & Conduit, and described further in Calculation BNP-PSA-080, BNP Fire PRA - Quantification,is not considered to be an ERFBS. Calculation BNP-PSA-080 and BNP-0209 evaluated that the pyrocrete encapsulated conduits/circuits in Fire Area DG-5, Diesel Generator Cell One, as capable to be one hour rated. Specifically stated:

While the Pyrocrete arrangement is not a tested and approved ERFBS in accordance with GL 86-10, Supplement 1, as an Electrical Raceway Fire Barrier System (ERFBS),

the installed Pyrocrete provides the identified conduits/circuits adequate protection to maintain free of fire damage conditions for an extended period of time and was considered to be similar to that of conduits protected through embedment in concrete structures. Thus, it is qualitatively reasonable to assume the circuits will be unaffected by fire for in excess of 1-hour.

The pyrocrete configuration identified above is credited in the FPRA for risk reduction purposes.

Pyrocrete is not credited in the deterministic analysis (i.e., not in Safe Shutdown Analysis (SSA)/

Nuclear Safety Capability Assessment (NSCA).

The responses to SSA RAI 8 (i.e., submitted June 28, 2013) and PRA RAI 9 (i.e., submitted July 15, 2013) provided additional clarification that this configuration is not credited as an ERFBS. The response to PRA RAI 9 identifies that credited ERFBS configurations are planned via modification and Table S-1 items 5 and 7 provide additional details.

The referenced Evaluation 96-00640 involves an evaluation of a Thermo-Lag radiant energy shield which is not credited in the FPRA and will be deleted. The referenced procedure OPT-34.15.9.7 will also be deleted as this procedure does not detail configurations which are to be installed.

Enclosure 2 Page 19 of 57 The Compliance Strategy "Complies" will be clarified with a statement in the "Compliance Basis" column to indicate the required ERFBS identified in Table S-1 will be installed in accordance with the requirements specified.

The updated Attachment A of the LAR to include this clarification will be provided in conjunction with the 120-day set of RAI responses.

Fire Protection Engineering RAI 16 Attachment A, Table B-i, 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).

Response

Duke Energy submits the following approval request, in accordance with 10 CFR 50.48(c)(2)(vii), for use of Electric Power Research Institute (EPRI) Technical Report (TR) 1006756, Fire Protection Equipment Surveillance Optimization and Maintenance Guide, Final Report, July 2003:

NFPA 805 Section 3.2.3(1)

In accordance with 10 CFR 50.48(c)(2)(vii), "Performance-based methods," the fire protection program elements and minimum design requirements of Chapter 3 may be subject to the performance-based methods permitted elsewhere in the standard.

In accordance with NFPA 805, Section 2.2.8, the performance-based approach to satisfy the nuclear safety, radiation release, life safety, and property damage/business interruption performance criteria requires engineering analyses to evaluate whether the performance criteria are satisfied.

In accordance with 10 CFR 50.48(c)(2)(vii), the engineering analysis performed shall determine that the performance-based approach utilized to evaluate a variance from the requirements of NFPA 805 Chapter 3:

A. Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; B. Maintains safety margins; and C. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire nuclear safety capability).

Duke Energy requests formal approval of performance-based exception to the requirements in Chapter 3 of NFPA 805 as follows:

Enclosure 2 Page 20 of 57 NFPA 805, Section 3.2.3(1):

Procedures shall be established for implementation of the fire protection program. In addition to procedures that could be required by other sections of the standard, the procedures to accomplish the following shall be established:

Inspection, testing, and maintenance for fire protection systems and features credited by the fire protection program.

Duke Energy requests the ability to utilize performance-based methods to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805. Performance-based inspection, testing, and maintenance frequencies will be established as described in EPRI TR 1006756.

Basis for Request:

NFPA 805, Section 2.6, "Monitoring," requires that "A monitoring program shall be established to ensure that the availabilityand reliabilityof the fire protection systems and features are maintainedand to assess the performance of the fire protection program in meeting the performance criteria. Monitoringshall ensure that the assumptions in the engineeringanalysis remain valid."

NFPA 805 Section 2.6.1, "Availability,Reliability, and PerformanceLevels, " requires that

'Acceptable levels of availability,reliability, and performance shall be established."

NFPA 805 Section 2.6.2, "MonitoringAvailability, Reliability, and Performance," requires that "Methods to monitoravailability,reliability, and performance shall be established. The methods shall considerthe plant operatingexperience and industry operatingexperience."

The scope and frequency of the inspection, testing, and maintenance activities for fire protection systems and features required in the fire protection program have been established based on the previously approved Technical Specifications / License Controlled Documents and appropriate NFPA codes and standards. This request does not involve the use of the EPRI Technical Report TR-1006756, Fire ProtectionEquipment Surveillance Optimization and Maintenance Guide, to establish the scope of those activities as that is determined by the required systems review identified in Table 4-3, "NFPA 805, Ch 4 Required FP Systems/Features."

This request is specific to the use of EPRI Technical Report TR-1006756 to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features credited by the fire protection program. As stated in EPRI Technical Report TR-1006756, Section 10.1, "The goal of a performance-based surveillance program is to adjust test and inspection frequencies commensurate with equipment performance and desired reliability."

This goal is consistent with the stated requirements of NFPA 805, Section 2.6. The EPRI Technical Report TR-1006756 provides an accepted method to establish appropriate inspection, testing, and maintenance frequencies which ensure the required NFPA 805 availability, reliability, and performance goals are maintained.

Enclosure 2 Page 21 of 57 The target tests, inspections, and maintenance will be those activities for the NFPA 805 required fire protection systems and features. The reliability and frequency goals will be established to ensure the assumptions in the NFPA 805 engineering analysis remain valid. The failure criteria will be established based on the required fire protection systems and features credited functions and will ensure those functions are maintained. Data collection and analysis will follow the EPRI Technical Report TR-1006756 document guidance. The failure probability will be determined based on EPRI Technical Report TR-1006756 guidance and a 95%

confidence level will be utilized. The performance monitoring will be performed in conjunction with the Monitoring Program required by NFPA 805, Section 2.6, and it will ensure site specific operating experience is considered in the monitoring process. The following is a flow chart that identifies the basic process that will be utilized.

Enclosure 2 Page 22 of 57 Piro,gram, ..rameWon..'k

.. entifyTarget Tests and Inspecotins, Establish Reliability andFrequency Goals, Set Failure Criteria.

Assess Ucensing impact and Other Consiraints 1.

Data Collection and'Evaluation Establish DataCoiiection Guideines*.

Collect RequiredSurveillance Data

.Assemble Data in Spreadseet or Database.

Analyze Datalto Identity Failures, Reliability and, UncertaintyAnalysis comput- Failure Probabiiltiesl Compute Uncertainty Limitsr Confirm That Reliability.Supports Target Frequency Program Implementation Modify Program Documents Revse Surveillance Procedures Conduct Ongoing Performance Monitoring Refine and Modify Frequencies as Appropriate EPRI TR-1 006756 - Figure 10-1 Flowchart for Performance-Based Surveillance Program BSEP does not intend to revise any fire protection surveillance, test or inspection frequencies until after transitioning to NFPA 805. Existing fire protection surveillance, test and inspection will remain consistent with applicable station, Insurer, and NFPA Code requirements. BSEP's intent is to obtain approval via the NFPA 805 Safety Evaluation to use EPRI Technical Report TR-1006756 guideline in the future as opportunities arise. BSEP reserves the ability to evaluate fire protection features with the intent of using the EPRI performance-based methods to provide evidence of equipment performance beyond that achievable under traditional prescriptive maintenance practices to ensure optimal use of resources while maintaining reliability.

Enclosure 2 Page 23 of 57 Nuclear Safety and Radiological Release Performance Criteria:

Use of performance-based test frequencies established per EPRI Technical Report TR-1 006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to Nuclear Safety Performance Criteria by the use of the performance-based methods in EPRI Technical Report TR-1006756.

The radiological release performance criteria are satisfied based on the determination of limiting radioactive release. Fire Protection Systems and Features may be credited as part of that evaluation. Use of performance-based test frequencies established per the EPRI Technical Report TR-1006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited to meet the Radioactive Release performance criteria. Therefore, there is no adverse impact to radioactive release performance criteria.

Safety Margin and Defense-in-Depth:

Use of performance-based test frequencies established per EPRI Technical Report TR-1006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features are maintained to the levels assumed in the NFPA 805 engineering analysis which includes those assumptions credited in the Fire Risk Evaluation safety margin discussions. In addition, the use of these methods in no way invalidates the inherent safety margins contained in the codes and standards used for design and maintenance of fire protection systems and features. Therefore, the safety margin inherent and credited in the analysis has been preserved.

The three echelons of defense-in-depth described in NFPA 805, Section 1.2 are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, fire rated cable, success path remains free of fire damage, recovery actions).

Echelon 1 is not affected by the use of the EPRI Technical Report TR-1 006756 methods. Use of performance-based test frequencies established per EPRI Technical Report TR-1 006756 methods combined with NFPA 805, Section 2.6, Monitoring Program, will ensure that the availability and reliability of the fire protection systems and features credited for defense-in-depth are maintained to the levels assumed in the NFPA 805 engineering analysis. Therefore, there is no adverse impact to echelons 2 and 3 for defense-in-depth.

==

Conclusion:==

NRC approval is requested for use of the performance-based methods contained in the EPRI Technical Report TR-1 006756, Fire ProtectionEquipment Surveillance Optimization and Maintenance Guide, Final Report, July 2003, to establish the appropriate inspection, testing, and maintenance frequencies for fire protection systems and features required by NFPA 805.

As described above, this approach is considered acceptable because it:

Enclosure 2 Page 24 of 57 a) Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; b) Maintains safety margins; and c) Maintains fire protection defense-in-depth (i.e., fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Fire Protection Engineering RAI 17 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 RB11-6 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.

Response

Fire Areas RB1-6 and RB2-6 are deterministically compliant as stated in Attachment C of the LAR. In each fire area, the sprinkler head over the RCIC steam isolation valve is not needed for deterministic compliance, since RCIC is not credited for Reactor Coolant System (RCS) inventory control in these areas. It is understood that areas that are deterministically compliant do not require an assessment of defense in depth; however, such an assessment is not precluded if the plant chooses to provide additional defense-in-depth to a deterministically compliant area.

In the pre-transition Appendix R analysis, the sprinkler head provided part of the justification for an approved exemption for these areas to allow crediting of the RCIC System for Appendix R compliance. This exemption is not being transitioned under the new licensing basis. In the NFPA 805 NSCA, inventory control is provided from the Main Control Room (MCR) using the Core Spray System following plant depressurization with the Safety Relief Valves (SRVs).

The use of Core Spray for inventory control meets the deterministic requirements of NFPA 805.

The presence of the installed sprinkler head increases the likelihood that RCIC will also be available from the MCR, and this forms part of the basis for crediting the sprinkler head for defense in depth.

As discussed in RAI FPE-19, an evaluation of Defense-In-Depth (DID) was performed for all fire areas, as detailed in project procedure FPIP-129,NFPA 805 Fire Safety Analysis. This evaluation was performed regardless of whether NFPA 805 compliance was demonstrated using a performance based approach or a deterministic approach. If future events result in a need to change the licensing basis for RB1-6 and RB2-6 from the deterministic approach of Section 4.2.3 to the risk-informed, performance-based approach of Section 4.2.4, the existing DID evaluation will serve to facilitate that change.

The licensing basis for each fire area, on a unit basis, is provided in the Regulatory Basis section of Attachment C of the LAR. Nothing in the Required Regulatory Systems table should be viewed as overriding this designation.

Enclosure 2 Page 25 of 57 Fire Protection Engineering RAI 18 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 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:

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 Drain Collector Basin Pumps.

vi. Temporary Cooling Water Supply to Service Air Compressor 1 (2) 0.

Enclosure 2 Page 26 of 57 vii. Transfer of Fire Protection System Water Supply to the make-up demineralizer Tank.

viii. Refill of standby gas treatment drain trough.

Response (Fire Protection Engineering RAI 18, Part a)

Containment Heat Removal The Nuclear and Conventional Service Water Systems are used for removing heat from the RHR system, Diesel Generator, and Reactor Building Closed Cooling Water (RBCCW) systems.

The RHR system in turn is used for removing heat from the Primary Containment and for reactor core decay heat removal.

In the unlikely event of a complete and sustained loss of Nuclear and Conventional Service Water, Abnormal Procedures direct the operator to align water from the fire protection tank to the RHR Heat exchangers. This would be done only after attempts to restore service water flow from any one of five pumps are unsuccessful and if actions to isolate major service water system leaks are not successful. Concurrent flow demands from Emergency Use and Fire Suppression are highly unlikely. If concurrent flow demands were to occur, the fire water system will recover quickly. Simultaneous use with the other three emergency uses is not likely.

Estimated flow and pressure demand for the Electric and Diesel Fire Pumps is 2000 gpm each, at a normal system operating pressure of approximately 125 psig.

Reference:

OAOP-18.0 Alternate Coolant Iniection In the unlikely event that reactor water level cannot be restored and maintained using installed high and low pressure injections systems, the Emergency Operating Procedures direct the operator to restore reactor coolant level using all of the following systems: Standby Liquid Control, Heater Drains, Service Water, Deminerialized Water, and Fire Protection Water.

Concurrent flow demands from Emergency Use and Fire Suppression are highly unlikely. If concurrent flow demands were to occur, the fire water system will recover quickly. Simultaneous use with the other three emergency uses is not likely. Estimated flow and pressure demand for the Electric and Diesel Fire Pumps are 2000 gpm each, at a normal system operating pressure of approximately 125 psig.

Reference:

OEOP-01-LEP-01 Alternate Boron Iniection Emergency Operating Procedures direct the operator to inject boron if it has been determined that the reactor will not remain shutdown under all conditions without boron and following a reactor scram. The operator is directed to inject boron with one or more of the following systems: Control Rod Drive, HPCI, RCIC, and RWCU. If RWCU System is used, the emergency procedure directs the operator to use a 1-1/2 inch fire hose to fill the system precoat tank to pre mix boron for injection. Estimated flow from the fire house should be less than 100 gpm, at a normal system operating pressure of approximately 125 psig. Filling a completely empty precoat tank should require not more than 300 gallons. There is adequate margin for this use concurrent with fire suppression. Simultaneous use with the other three emergency uses is not likely.

Reference:

OEOP-01-LEP-03.

Enclosure 2 Page 27 of 57 Fuel Pool Cooling The spent fuel pool decay heat removal systems consist of Fuel Pool Cooling and Alternate Decay Heat Removal System. Additionally the Division II RHR system is capable of providing fuel pool cooling assist. If an abnormal event occurs that results in decreasing water level or increasing fuel pool temperature, the Demineralized Water and the Fire Protection Water Systems will be aligned to provide make up and cooling. In the unlikely event of a complete and sustained loss of these systems, Abnormal Operating Procedures direct the operator to add water from the Demineralized Water System and the Fire Protection Water System. Fire hoses will be used to direct as much water as necessary to restore and maintain water level in the spent fuel pool. Flow for three fire hoses is estimated at 250 gpm, at a normal system operating pressure of approximately 125 psig. Concurrent flow demands from Emergency Use and Fire Suppression are highly unlikely. If concurrent flow demands were to occur, the fire water system will recover quickly. Simultaneous use with the other three emergency uses is not likely.

Response (Fire Protection Engineering RAI 18, Part b)

Following the use of fire protection water for Containment Heat Removal, Coolant Injection, Alternate Boron Injection, or Fuel Pool Cooling, restoration requirements would be to realign the system to standby and refill the Fire Protection Tank from the Brunswick County Water Supply System. Restoration is described in plant operating procedures. The Fire Protection Water Tank is filled and maintained through 1.5-inch and 4-inch air-operated fill valves. Level switches on the tank control automatic makeup, and a low level alarm is provided in the MCR. A manual bypass valve may also be used to refill the tank. Fill water is supplied by a County Water Storage Tank, with two parallel pumps supplying flow. The County Water Storage Tank is maintained full by the Brunswick County Water Main. Additionally, a design feature is to manually align the Electric Motor and Diesel driven fire pump suctions to the Demineralized Water Storage Tank, which will allow time for the Fire Protection Storage Tank to refill.

Response (Fire Protection Engineering RAI 18, Part c)

Use of fire protection water for Containment Heat Removal, Coolant Injection, Alternate Boron Injection, and Fuel Pool Cooling are strictly controlled by Emergency Operating and Abnormal Operating procedures. Operators are trained regularly on these procedures and the equipment.

Communications are by the plant Public Address (PA) System, sound powered phones, and operations radio systems.

Response (Fire Protection Engineering RAI 18, Part d)

The Fire Protection Storage tank level is controlled by two automatic makeup valves from the Brunswick County Water Supply system. These 1.5-inch and 4-inch valves are controlled by level switches on the tank. A manual bypass valve is also provided. Routine surveillance checks by plant operators using a local tank level indicator verify that the tank level is kept above the minimum level. Operators in the control room are alerted by annunciator ifthe tank level is drawn down. Two alarms are provided at a low and a low-low level alarm set point.

Automatic and makeup supply is provided from a 15,000 gallon on site County Water Storage Tank. This tank is filled from the county water main by two pumps delivering 200 gpm each.

Level instrumentation is in feet above the tank bottom. The Demineralized Water Storage Tank is checked regularly by Operations, in the same manner as the Fire Protection Water Storage Tank, to verify it is 14 foot above the tank bottom.

Enclosure 2 Page 28 of 57 Response (Fire Protection Engineering RAI 18, Part e)

Based on flow rates and volumes used from the fire protection water supply explained in normal evolutions i thru viii below, the margin available in the fire protection water supply system is adequate. The largest water demand for a safety related area, Unit 2 South RHR, and the largest water demand for a non-safety related area, the Main Transformer, were both calculated to be within the capacity of a single fire pump. These flow demands are discussed in DBD-62, Section 3.3.4.

Estimated flow, pressure, and expected frequency, as applicable, are discussed in the paragraphs below. Each normal evolution is performed under procedural controls. Annunciators for low tank level or local monitoring will alert the operator such that minimum tank level will not be violated. Alerts are provided by tank low and low-low alarms, along with Electric Motor and Diesel Driven Fire Pump running annunciators in the MCR.

i. Residual Heat Removal (RHR) Service Water Shutdown and Wet Layup Process Usage of fire protection water for RHR Service Water (RHRSW) wet layup should be allowed because there is no appreciable flow of fire water from the Fire Water Storage tank. Wet layup following RHRSW system shutdown does not place a significant drain on the Fire Protection system. The RHRSW automatic valve controls and operating procedures will isolate valve 1(2)SW-V143 if the RHR Service Water system is placed in service to the RHR heat exchangers. While in a static wet layup alignment any RHR Service Water system leakage should not exceed the capacity of the county water make up to the Fire Protection Water storage tank nor the capacity of the two fire pumps.

ii. Flushing, filling, and venting RHR Service Water and Heat Exchangers Usage should be allowed because procedural controls prevent the operator from lowering tank level below the low level alarm set point. When performing the flush, operating procedures require an operator to be stationed to continuously monitor tank level locally and to maintain direct communications with the MCR by plant PA or radio.

Procedures require usage of not more than one-half foot tank level for each flush evolution. This amount of usage is well within the capacity of the county water makeup flow.

iii. RHR Service Water System Operability Test Quarterly testing is performed on each RHRSW system. There are two divisions on each unit. Following each test the system is flushed per operating procedures. Not more than one-half foot of Fire Protection Storage Tank level is used for each flush evolution. Usage is well within the Fire Protection Water Storage Tank makeup capacity and the volume stored in the County Water Storage Tank is more than enough for immediate use. Usage should be allowed because procedural controls prevent the operator from lowering tank level below the low alarm set point and immediate makeup is available.

Enclosure 2 Page 29 of 57 iv. Flushing Radwaste Radiation Monitor Periodic flushing of the Radwaste Radiation Monitor is performed at an estimated flow rate of 200 gpm. The expected frequency of this normal evolution is 212 flushes per year. Operators are in direct control of this evolution and procedural controls require that not more than 2000 gallons be used for each evolution. Gallons used is indicated in the Radwaste Control Room. This amount of flow and volume is well within the makeup capacity of the automatic valves and pumps that provide flow to the Fire Protection Water Storage Tank.

v. Seal water to Storm Drain Collector Basin Pumps Usage of seal water from the Fire Main to Storm Drain Basin pumps should be allowed because it is a small amount of flow. There are three pumps, and seal flow for all is not more than 40 gpm. Seal pressure is regulated to approximately 15 psig. Seal flow is used only when the pump is placed in service to lower Storm Drain Basin level. The expected frequency is dependent on rain fall. Usage should be allowed because demand from these pumps is well below the capacity of the two fire pumps, which are sized to deliver 2000 gpm each. This is an insignificant amount when compared to the large volume of the Fire Protection Water Storage.

vi. Temporary Cooling Water Supply to Service Air Compressor 1(2)D Usage of temporary cooling to Service Air Compressors should be allowed because this alignment is used typically once each refueling outage, and flow rates are well within the system capacity. Procedures require the affected unit to be in Mode 4 (i.e.,

Cold Shutdown) or Mode 5 (i.e., Refueling). Procedures direct that cooling flow to the air compressor be connected by a hose with pressure and flow regulated. Pressure and flow are estimated at 51 gpm and 44 psig. There is adequate margin in the capacity of the two fire pumps of 2000 gpm each and makeup capacity to the fire tank thru a 1.5-inch and 4-inch automatic make-up valve. Low tank level alarms are provided in the MCR. The Control Room Supervisor is in direct control of this procedure.

vii. Transfer of Fire Protection System Water Supply to the Make-Up Demineralizer Tank Transfer of Fire Protection System Water Supply to the Make-Up Demineralizer Tank does not use fire water. This item refers to procedure OOP-41, which is designed to realign the fire pumps suction path from the Fire Protection Tank to the Demineralized Water Storage Tank.

viii. Refill of Standby Gas Treatment Drain Trough Usage of installed fire protection piping and valves should be infrequently allowed to fill the Standby Gas Treatment (SGT) drain trough because the flow and volume are insignificant when compared to the Fire Protection Tank volume. The purpose of the trough is to ensure loop-seals can prevent by-pass leakage from the SGT filter compartments. Water level is checked regularly by plant operators and a small amount is added if needed to replenish evaporative loses. Expected frequency is dependent on evaporation rate normally less that twice per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Total volume of each trough is approximately 60 gallons. To completely fill the trough from a normally closed

Enclosure 2 Page 30 of 57 1/2-inch valve is insignificant compared to the volume of the Fire Protection Water Storage Tank and the design flow of the two fire pumps.

Fire Protection Engineering RAI 19 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 DID 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.

Response

An evaluation of Defense-In-Depth (DID) was performed for all fire areas as detailed in project procedure FPIP-1 29, NFPA 805 Fire Safety Analysis. This evaluation was performed for all areas, regardless of whether NFPA 805 compliance was demonstrated using a performance based approach or a deterministic approach.

Although a discussion of DID features is not strictly required for areas that are deterministically compliant, the decision to include the evaluation for such areas was based on two factors. First, it was seen as a way of enhancing the overall approach to providing the plant's desired level of fire protection to that area. Second, if future changes to deterministic areas dictate that a performance based approach may provide benefit, then including these features as credited DID features now will facilitate that transition.

The licensing basis for each fire area, on a unit basis, is provided in the Regulatory Basis section of Attachment C of the LAR. Nothing in the Required Regulatory Systems table should be viewed as overriding this designation.

It was noted in preparing this response that the Required Regulatory Systems table for Fire Area MWT-1 is incorrect. The Fire Safety Analyses for this area (i.e., 1 FP-1 108, 2FP-1 178) concluded that the installed suppression and detection systems were not needed for defense-in-depth.

Additional information related to the methodology used in performing the DID evaluation is provided in the response to RAI SSA-09.

Fire Protection Engineering RAI 20 Attachment A, Table B-i, 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 51 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.

Enclosure 2 Page 31 of 57

Response

Regarding use of a roving Hot Work Fire Watch:

FIR-NGGC-0003, Section 3.16 states:

A fire watch may be designated to rove through an area to cover several hot work jobs in relatively close proximity to each other. [HNP and RNP - During Modes 5 or 6] [BNP -

During Modes 4 or 5] [CR3 - A fire watch may be designated to rove through an area to cover several hot work jobs that are within close proximity and line of sight of each other.] In all cases, all provisions of the Fire Watch Responsibilities must be controlled in their entirety at all times for all areas covered.

NFPA 51B compliance is evaluated in fleet calculation NED-M/BMRK-0001, Code Compliance Evaluation for NFPA 51B, Standard for Fire Prevention during Welding, Cutting, and Other Hot Work- 1999 Edition, and OFP-1051, Code Compliance Evaluation NFPA 51 B, Standard for Fire Protection in Use of Cutting and Welding Processes. The NFPA standard is silent regards use of a single fire watch for multiple hot work jobs. This accompanied by the administrative control to ensure hot work jobs are within close proximity of each other and all provisions of the Fire Watch Responsibilities must be controlled in their entirety at all times for all areas covered, the fire watch requirements of NFPA 51 B are met.

Regarding use of video camera monitoring for fire watch applications:

FIR-NGGC-00003, Section 4.8.9 states:

As applicable, ensure an appropriate type portable fire extinguisher is positioned at the Hot Work location for Hot Work activities that are to be viewed by video camera and monitor.

Additional administrative requirements are applied to the use of video surveillance for hot work fire watch activities. These are specified in BSEP site procedure OFPP-005, Fire Watch Program. This procedure stipulates the following additional controls to be applied for use of video equipment. Use of the equipment is typically limited to ALARA applications and those specifically identified in the procedure.

6.2.2 For ALARA reasons, it may be beneficial to use a video camera and monitor to support or replace a posted fire watch. If this option is used, the following requirements apply:

1. In order for a video camera and monitor to be used to support or replace a posted fire watch, approval of the activity must be obtained from the Fire Protection Program Manager and the Manager-Operations, or designee, and documented on Attachment 4 of this procedure.

2. When approved, Attachment 4 shall be attached to the Hot Work Permit (FIR-NGGC-0003) or the Fire Watch Standers Log (OFPP-005), as applicable.

3. When a video camera and monitor are used, the fire watch shall be responsible for the following additional actions:

Enclosure 2 Page 32 of 57

a. The fire watch shall be briefed, dressed, and equipped to meet RWP and posted requirements for quick entry into the affected area if it becomes necessary.

b. If the fire watch would normally be assigned to continuously monitor the area or activity, then the fire watch shall maintain an unobstructed view of the video monitor for the duration of the assignment.

c. If the activity involves hot work viewed by video camera or monitor, the fire watch shall ensure that a portable fire extinguisher is staged at the Hot Work site.

d. If the activity involves hot work, a means of communication with personnel performing the hot work activity shall be established. If the video camera or monitor becomes inoperable, all hot work shall be stopped by the fire watch until either the video camera or monitor is returned to service or a fire watch is posted in the area of the hot work.

4. Use of video monitoring is approved for MSF [Members of the Security Force]

hourly rounds of the following areas. MSF should ensure that camera coverage of the area is unobstructed each time it is used (e.g., no equipment staged in front of camera, no camera functional failure resulting in loss of view or clarity).

a. Unit 1 HPCI roof

b. Unit 2 HPCI roof

5. When video monitoring is no longer required, Attachment 4 will be forwarded to the Fire Protection Coordinator, who will retain it for a period of 30 days.

Fire Protection Engineering RAI 21 Attachment A, Table B-i, 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. Describe 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 DID and safety margin that justifies any additional duties.

Response

BSEP is not using a Fire Brigade Operations Advisor. Instead, BSEP will utilize a fire brigade where, during every shift, the brigade leader and at least two brigade members shall have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance. This is consistent with NFPA 805, Chapter 3 requirements and fleet procedure FIR-NGGC-0007, NFPA 805 Fire Brigade Training Program.

Enclosure 2 Page 33 of 57 Safe Shutdown Analysis Requests for Additional Information SSA RAI I 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 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.

Response

In the event of a MCR enclosure fire, prompt detection will provide operations personnel with the necessary time to limit the fire damage by performing the requisite initial response to an In-Cabinet Incipient Detection alarm as prescribed in plant annunciator procedures with the time necessary to limit or delay the onset of fire damage. The VEWFDS will be integrated into the BSEP ESTTM fire detection system with Main Control Room annunciations "Trouble," "Pre.-F',re,"

and "Fire" (i.e., equivalent to "Trouble," "Alert," and "Alarm" conditions). Control room operators will respond to these alarms in accordance with plant operating procedures. Qualified on-shift Operations and/or Maintenance personnel will respond to investigate all alarms without delay and with the same response urgency, and will ensure that there is continuous attendance of the affected area until the condition is resolved. Responding personnel will have basic training sufficient to initiate early fire fighting activities (i.e., use of portable fire extinguisher equipment) such that the expectation will be satisfied that if a developing fire is discovered, there will be an immediate intervention to suppress/control the fire. Future operator training will include specific alarm procedures and fire protection system operability procedures and will be accomplished when the installation has been completed and the new procedures are approved. In order to remain in the MCR post fire, operators will base decisions concerning mitigating actions using damage assessments derived from communications with first responders, knowledge of what systems are associated with the degrading components, an evaluation of plant control based upon available indications, and the prevailing environmental conditions in the MCR. Also, the incipient alarm itself will provide insight regarding potentially fire affected equipment and assist in anticipating both spurious operations and other possible plant responses to the fire damage.

By ultimately limiting both the duration and the size of the fire, use of this VEWFDS provides the operator with an extremely early warning of a potential fire event and thereby minimizes the probability of an evacuation.

SSA RAI 2 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.

Enclosure 2 Page 34 of 57

Response

The intent of the modifications is to provide a delay in fire damage for risk reduction. Inasmuch as the redundant cables are within the same fire area, the intent of these modifications is not to meet the requirements of Section 4.2.3.3 of NFPA 805, but rather to reduce CDF and/or LERF.

For Item #5, the protection scheme will take the form of 1-hour fire rated ERFBS wrap on conduit 161L1/BA to the extent required to protect it within the zone of influence (ZOI) of the ignition source listed in the LAR (4521 "1-1L - 480V UNIT SUBSTATION 1L"), thereby preventing or delaying fire damage to that portion of the cable that supplies power to MCC 1-3A.

The power supply supports the operation of SRVs, as well as instrumentation and circuits supporting Emergency Core Cooling Systems.

Similarly, the modification described in Item #7 will protect the cables listed in the LAR by providing separation from ignition sources or 1-hour of fire rated protection. Again, the purpose is to delay fire induced cable damage rather than crediting separation to assure that the cable or its redundant is free of fire damage. The LERF reduction will be realized by reducing the likelihood of fire induced damage resulting in containment bypass from the main steam isolation valves. At this time, several options are still being considered.

SSA RAI 5 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.

Enclosure 2 Page 35 of 57

Response

Although safe and stable conditions were not included as an end state in the FPRA, BSEP qualitatively evaluated the actions and activities required to maintain those conditions and concluded that they represent relatively low risk evolutions. The factors considered were related to simplicity, ensured equipment availability, and the routine nature of replenishing commodities.

Following the establishment of safe and stable conditions, the ability to control reactor pressure, inventory, and temperature requires limited operator involvement as the actions are characterized by simple manipulations of valves and/or pump controls, and process instrumentation is readily available at appropriate locations. Additionally, the NSCA has shown that the equipment required to achieve safe and stable conditions for all fire scenarios should be free of fire damage, and thus operable either remotely from the MCR for fires involving Fire Area CB-23E or locally when the MCR (i.e., CB-23E) becomes uninhabitable.

Longer term, the risk associated with the replenishment of on-site commodities such as nitrogen and diesel fuel oil is very low based on the routine nature of ordering and taking delivery of them. Moreover, assistance is available long term for the plant operators not assigned to the shift in the form of trained emergency response personnel.

The systems, evolutions, and resources required to maintain safe and stable conditions are discussed below:

a. The actions necessary to align the Emergency Electrical Distribution, Service Water, RHR, RCIC, and/or High Pressure Coolant Injection (HPCI) Systems are completed when establishing hot shutdown. Therefore, actions required to maintain safe and stable conditions are limited to simple control activities such as adjusting service water and/or RHR flow. Longer term, no additional operator interventions are required other than the replenishment of on-site commodities, such as nitrogen and fuel oil. For the reasons discussed in paragraph b below, the risk associated with these activities is very low.

b. In establishing and maintaining the safe and stable conditions defined above, on-site storage capacity limitations on diesel engine fuel and SRV nitrogen supplies will affect two performance goals, Vital Auxiliaries (i.e., Electrical) and Decay Heat Removal.

Condensate Storage Tank (CST) water capacity is not limiting since sufficient suppression pool volume is available to the RCIC or HPCI pump suctions to permit adequate RCS Inventory Control, and the additional water volume of the CST is not required to control suppression pool temperatures. Calculation ORNA0001, Instrument Air Nitrogen Backup System Volume Requirements, shows that on-site supplies of nitrogen are sufficient for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, after which bulk liquid nitrogen is available long term from a local vendor or vendors who supply the liquefied gas under contract during normal plant operations. If one or more of the EDG's are operated to supply AC power, delivery of additional fuel oil would be required only if the normal supply approaches exhaustion after seven days. Should this occur, ordering and accepting delivery of fuel is a routine activity for BSEP that can be anticipated in ample time to ensure its continuous availability long term.

c. For safe and stable conditions to be established and maintained long term, the shift staffing minimums are adequate to provide all of the operators required to perform recovery actions or recovery actions - defense in depth. This can be accomplished while maintaining the minimum MCR personnel specified in the Technical Specifications for all

Enclosure 2 Page 36 of 57 fire scenarios including Control Room abandonment. In the most limiting case, additional operators are available on shift to perform recovery actions or recovery actions -

defense in depth while the MCR remains staffed. To achieve and maintain safe and stable conditions for all fire scenarios at BSEP, the time critical post fire events are to first establish RCS injection and then suppression pool cooling. For the MCR abandonment scenario, an additional action exists related to restoration of power to the charger for the credited station battery. In all cases, where these recovery actions are required, they have been shown to be feasible using specified minimum shift staffing in the BSEP manual action feasibility study, BNP-E-9.007.

There are no fire scenarios in which "off-shift" personnel are required to perform any actions in order to achieve and maintain safe and stable conditions within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. For fire initiated events where the Emergency Response Organization is activated, personnel are available to ensure that the longer term steps required to replenish on-site commodities can be reliably accomplished.

d. The risk associated with the failure of operator actions required to sustain safe and stable conditions is extremely low given the nature of the required actions, as described in paragraphs b and c above. Supplemental resources are available for actions required to replenish commodities for the long term.

SSA RAI 9 Section 4.5.2.2, 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] 50.48(c)," Revision 2, and the general strategy of looking for substantial imbalance in the echelons is described at a high level manner in Section 4.5.2.2 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 805 performance-based Section 4.2.4 compliance strategy.

Response

Duke Energy uses a project instruction to provide additional guidance when performing the Defense-in-Depth and Safety Margin on an area by area basis. Considerations for defense in depth included both recovery actions and fire protection systems and features.

Attachment C of the LAR will be revised to include a summary description of the methodology and criteria used to determine which recovery actions and other fire protection features were retained as defense in depth. A similar discussion of how the safety margin determination was evaluated will also be included. The revision to Attachment C of the LAR will be submitted with the 120-day RAI responses.

Enclosure 2 Page 37 of 57 SSA RAI 11 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."

Response

a. A listing of components (including power supplies) added, that were not included in the at-power analysis and of those at-power components that have a different functional requirement for NPO are shown below:

1(2)-E21-CO01A-M A Core Spray Pump Motor 1 (2)-E21-FO01A-MO 1(2) A Core Spray Torus Suction Valve 1(2)-E21-FO04A-MO 1(2) A Core Spray Outboard Injection Valve 1 (2)-E21-FO05A-MO 1(2) A Core Spray Inboard Injection Valve 1 (2)-E21-FO15A-MO 1(2) A Core Spray Full Flow Test BYP Valve 1(2)-E21-FO31A-MO 1(2) A Core Spray Min Flow Bypass Valve 1 (2)-E21-FI-R601A 1(2) A Core Spray Flow Indicator 1 (2)-E21-PI-R600A 1(2) A Core Spray Pressure Indicator 1(2)-SW-V101-MO-OC 1(2) Conventional SW Supply 1 (2)-SW-V105-MO-OC 1(2) Nuclear SW Supply 1(2)-SW-V13-MO-OC 1(2) CSW Pump A Discharge Valve 1(2)-SW-V14-MO-OC 1(2) NSW Pump A Discharge Valve 1(2)-SW-V15-MO-OC 1(2) CSW Pump B Discharge Valve 1(2)-SW-V16-MO-OC 1(2) NSW Pump A Discharge Valve

Enclosure 2 Page 38 of 57 1 (2)-SW-V17-MO-OC 1(2) CSW Pump C Discharge Valve 1(2)-SW-V18-MO-OC 1(2) CSW Pump Discharge to Nuclear Header Valve 1(2)B-ABO-52 Switchgear 1(2)B Incoming Line from UAT Circuit Breaker 1(2)B21-FO16-MO Main Steam Line Drain Inboard Isolation Valve Motor Operator 1 (2)-B21-F019-MO B21-F019 Motor Operator 1(2)-B21-LI-R605A(B) Reactor Water Level Indicator 1(2)-B21-SV-F003 B21-F003 Pilot Solenoid Valve 1(2)-B21-SV-F004 B21-F004 Pilot Solenoid Valve 1-1 C-AC5-52 Switchgear 1C Incoming Line from UAT Circuit Breaker 1-1D-AD7-52 Switchgear 1D Incoming Line from UAT Circuit Breaker 1 (2)-MPT-MAIN-XFMR-A,B,C UNIT 1(2) Main Power Transformer Phase A,B,C 1-PCB-20A, B Circuit Breaker for Castle Hayne East Feeder #20 1-PCB-22A, B Circuit Breaker for Main Transformer #1 1(2)-UAT-UNIT-AUX-XFMR Unit Auxiliary Transformer 1 2-2C-AC4-52 4KV SWGR 2C Incoming Line from UAT Circuit Breaker 2-2D-AD6-52 Circuit Breaker at Switchgear 2D Incoming Line from UAT 2-PCB-28A, B Circuit Breaker for Wallace Feeder #28 2-PCB-29A, B Circuit Breaker for Main Transformer #2

b. Calculation BNP-E-9.01 1, Revision 0, Attachment 1, shows all unavailable KSF paths and pinch points by fire area.

Calculation BNP-E-9.01 1, Revision 0, Section 5.1, provides a summary level identification of unavailable paths and pinch points as shown below. BNP-E-9.01 1, Revision 0, is included on the SharePoint website.

RHR System - Suction Valves, Discharge Valves and Flow-Path Valves can potentially fail during a fire event resulting in loss of shutdown cooling or inventory.

o Pinch point areas are:

• Battery Rooms

• DG Switchgear Rooms

• DG Engine Rooms

• Reactor Buildings

• Drywell

• Turbine Buildings

" Service Water System - Boundary Valves or pumps can potentially fail during a fire event resulting in a loss of service water flow or pump operation.

o Pinch point areas are:

• Service Water Building

• Reactor Buildings

• Turbine Buildings Electrical System - Power Supply, Control Power, Batteries, and Switchgear Ventilation can potentially be lost or limited due to a fire event

Enclosure 2 Page 39 of 57 o Pinch Point areas are:

• Reactor Buildings

• Turbine Buildings M Service Water Building

c. Although the NPO analysis takes credit for the at-power pre-fire rackout of the RHR shutdown cooling suction valves prior to establishing shutdown cooling, the proposed NPO procedures will not prescribe any actions to specifically mitigate the impact of fire-induced spurious actuations, either pre or post fire. However, in order to allow for unforeseen shutdown alignments for which spurious operation of a valve may cause the loss of a KSF, it may be allowable to de-energize one or more valves, subject to an engineering evaluation, as part of the strategy, to mitigate the potential loss of that KSF.

d. No locations exist at BSEP where NPO associated KSFs are achieved solely through Recovery Actions. There are certain instances where recovery actions are identified as possibly being needed if a KSF path is not available, as noted in Calculation BNP-E-9.01 1, Revision 0, and described in response to RAI SSA-1 lb above. Validation of these variable in-plant procedures uses a procedure review and validation process for AOPs, APPs, etc., for feasibility evaluation of any actions (PRO-NGGC-0204, Procedure Review and Approval).

e. No new, changed or deleted actions have been identified due to the NPO analysis.

SSA RAI 13 Table G-l, 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.

Enclosure 2 Page 40 of 57

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.

Response

a. The design of the BSEP electrical distribution system features cross unit powered components within the same train. This means that for any single fire, the same train of electrical power supports safe shutdown for both units, and some power supplies may be relied upon by both units to achieve safe and stable conditions, if a dual unit shutdown is required.

There is a fire scenario in the Unit 1 E6 switchgear room crediting Train A power for both units wherein Train B of the non-fire affected Unit 2 electrical distribution powers Unit 1 Train A. However, the defense-in-depth operator actions taken do not result in an alignment that is considered a traditional cross-connect because Train B of the Unit 2 emergency electrical distribution system is not required to power Train B components in Unit 1.

b. Although not strictly considered a cross-tie scenario, for the fire described in paragraph a., above, there are no cross unit recovery actions necessary to achieve safe and stable conditions that require Unit 2 operators.

c. For the reasons discussed in paragraphs a and b, above, there is no operational impact on the non-fire affected unit affecting its ability to achieve and maintain safe and stable conditions for any fire scenario

d. The FPRA separately analyzes the impact of all MCR fires on each unit. However, there is a risk impact for a particular unit only if the fire damages PRA components for that unit. If no PRA component on a particular unit is damaged by fire, the associated risk of a shutdown is assumed to be already considered in the internal events analysis.

To simplify the development of the FPRA, each unit is assumed to shut down for any MCR fire. This is a conservative assumption since not all fires will cause an automatic shutdown or damage sufficient equipment to require a manual shutdown. This assumption also encompasses shutdowns that might occur due to operator discretion in exercising conservative decision making. With the exception of MCR abandonment due to habitability, no attempt was made to establish any criteria for when a shutdown is actually required for either unit. For Main Control Room Abandonment (MCRA) due to habitability concerns, a detailed Human Reliability Analysis (HRA) was performed for each unit, as described in Attachment 10 of Calculation BNP-PSA-084, BNP Fire PRA -

Human ReliabilityAnalysis. This HRA was based on OASSD-02, Control Building, which provides guidance for responding to fires that require a MCRA and includes actions on both units. As described in Attachment 16 of Calculation BNP-PSA-080, BNP Fire PRA -

Quantification,the CDF and LERF for MCRA were estimated to be about 1 E-6 and 1E-7, respectively. These values are applicable to both units.

Enclosure 2 Page 41 of 57

e. BSEP Technical Specifications 3.8.7 and 3.8.8 simply prescribe divisional E-bus operability and do not address any particular electrical system alignment.

SSA RAI 14 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.

Response

a. BSEP believes that neither the 1984 ASCA, nor its associated Safety Evaluation (SE),

contain sufficient detail to conclude that any actions, either primary control station (PCS) or recovery action (RA), were specifically and individually reviewed and approved by the NRC at the time. The installation and features of the remote shutdown panel (RSDP) are only broadly described in the SE, and a merely general reference is made to BSEP's detailed submittal of its alternative shutdown capability in the ASCA. Regardless of the facts that the SE notes the existence of an alternative shutdown capability, the ASCA notes a few of the required actions, and that the ASCA was reviewed by the NRC, specific actions are neither discussed nor listed in the SE. Consequently, there is no evidence that explicit approvals of any actions by the NRC were provided.

These realities dictated that BSEP proceed without presupposing approval of any actions, so those designated as "RA" in Attachment G were subject to a risk evaluation and included in the LAR for approval under the NFPA 805 transition process. The actions listed in Attachment G as "PCS" make use of the controls installed on the remote shutdown panel located in the southeast corner of the 20 foot level of the Unit 1 and Unit 2 Reactor Buildings that have complete isolation capability from the MCR. BSEP omitted these actions from the risk evaluation because they are taken at the primary control station, as defined in Section 2.4(b) of Regulatory Guide (RG) 1.205, Revision 1.

In both cases, BSEP made no assumptions regarding the pre-approval of any recovery action.

b. For fire scenarios that require MCR abandonment, only the remote shutdown panel becomes a primary control station as defined in Section 2.4 of RG 1.205, and actions taken at that location are coded as "PCS" in Attachment G. For all other fire areas, actions taken at the Main Control Room panels located on the 49 foot level of the

Enclosure 2 Page 42 of 57 Control Building for Units 1 and 2 are MCR actions and any activities occurring at the RSDP are considered recovery actions.

c. The actions meeting the guidance of RG 1.205, Section 2.4(b), are listed in Attachment G as "PCS." Brunswick does not have "dedicated" shutdown systems or equipment meeting the definition of RG 1.205, Section 2.4(a).

Actions designated as "RA" or "RA-DID" in Attachment G are actions performed at locations outside of the main control room or primary control stations per Revision 1 of RG 1.205 and therefore, were evaluated for additional risk.

Enclosure 2 Page 43 of 57 Probabilistic Risk Assessment (PRA) Requests for Additional Information PRA RAI IJ 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

[not] appear to be fully resolved:

j) F&O 2-2 against CS-A 1 (Cat I/Il/Ill), CS-A3 (Cat I/Il/Ill), CS-Cl (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.

Response

Calculation BNP-PSA-085, BNP Fire PRA - Component Selection, concerns FPRA component selection and identifies the resulting FPRA Component List as an input to the FPRA Cable Selection task, Cable selection is performed under administrative controls previously developed for safe shutdown analysis because the FPRA uses the safe shutdown database to correlate fire-damaged circuits to basic events representing particular equipment failure modes.

The methodology for safe shutdown cable selection is documented in the fleet procedure, FIR-NGGC-0101, Fire ProtectionNuclear Safety CapabilityAssessment (NSCA).

FIR-NGGC-01 01, Section 9.3, discusses the methodology for cable selection and circuit analysis. The methodology implemented in this section of the procedure considers the required end state for the component as a function of the nuclear safety performance, non-power operation or Fire PRA criteria, and is based upon the guidance provided in NEI 00-01. When appropriate, detailed analysis of selected on and off scheme cables was used that considered the cable failure probabilities in Task 10 of NUREG/CR-6850, Volume 2. For the FPRA, FIR-NGGC-0101, Section 9.3.5, provides clarifications on cable selection and circuit analysis methodologies that might be different from those for safe shutdown or non-power operation components.

The cable location methodology makes use of software known as the Fire Safe Shutdown Program Manager Database (FSSPMD). This database is maintained using applicable quality controls and contains cable to raceway data and correlates the raceways to their respective fire zone locations.

Each required circuit was evaluated to determine which cables are necessary to support the post-fire function of the component. If a fire induced fault of the cable can place the component in a position/condition other than the desired position/condition due to activation of the circuit, the cable was identified as a spurious (S) cable. In situations where a fire induced cable fault could prevent a component from performing its intended function, that cable was assigned a code representing failure (F). If a fire induced failure of the cable cannot spuriously reposition or prevent the desired operation of the component, the cable was not considered a required post-fire cable. The identification of required cables not only provides a list of each cable, but it also establishes a link to the associated component and to the cable's routing and location within the plant. These relationships provide the basis for identifying potential equipment functional failures at a raceway, fire zone, and fire area level.

Enclosure 2 Page 44 of 57 The cable selection and location methodology and data were developed and stored together.

This fulfilled the requirements of CS-Al, CS-A3, and CS-Cl and obviated the need for a separate PRA notebook.

PRA RAI 1 K k) F&O 2-14 against FSS-D7 (Cat I):

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.

Response

Procedure EGR-NGGC-0010, System & Component Trending Program and System Notebooks, provides instructions for trending and monitoring structures, systems, and components to permit early detection of problems. These sources also provide essential support to several other regulatory required programs, including maintenance rule, mitigating system performance index (MSPI), and license renewal aging management. The trending program and system notebooks act as a historical performance trending database and were reviewed for "outlier behavior" of the Fire Detection and Suppression Systems.

To support the FPRA Peer Review a review of the then-current one year period (i.e., 2011) of System Health Report information was performed and indicated no "outlier behavior" for the Fire Detection and Suppression Systems. This focus on a 1-year period provided an overview of the most current system performance, measured against specific parameter/attributes, and helped to confirm the effectiveness of preventative and corrective maintenance. To further support the response to this RAI, the most recent three years of System Health Report information were also reviewed and found to show sustained acceptable performance levels, again with no "outlier behavior" noted for the Fire Detection and Suppression Systems.

PRA RAI IM 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.

Response

As clarified by Regulatory Guide (RG) 1.200, Rev. 2, Supporting Requirement FSS-D3 is considered met because detailed fire modeling was performed for all ignition sources and the development of the FPRA included the selection and application of either computational or non-computational fire modeling tools consistent with the guidance in Section 11.5.1.7.1 of NUREG/CR-6850. While there may be conservatisms associated with the selection and application of particular fire modeling tools, the fire modeling tools described in

Enclosure 2 Page 45 of 57 NUREG/CR-6850 are sufficiently accurate, despite any associated conservatisms, for the fire modeling of all physical analysis units and scenarios to be sufficiently realistic rather than bounding.

With regard to whether fire modeling is bounding rather than realistic, the fundamental contention is whether the fire modeling tools in NUREG/CR-6850 accurately characterize the fire risk or only produce bounding results. Whereas the Peer Review Team suggested that NUREG/CR-6850 represents overly conservative methods that can produce only bounding results, BSEP asserts that the fire modeling tools described in NUREG/CR-6850 are sufficiently accurate, despite any associated conservatisms, to be used to meet Capability Category II. This assertion is based on the premise that NUREG/CR-6850 is intended to provide acceptable guidance for satisfying regulatory requirements, and a Capability Category II is typically required for a PRA used in a regulatory application. Consequently, the fire modeling of all physical analysis units and scenarios is considered sufficiently realistic.

The selection and application of fire modeling tools was part of a larger iterative process for evaluating high risk contributors. This process involved a team of knowledgeable individuals with diverse expertise, including fire modeling, circuit analysis, PRA, and plant operations. As described in Attachment 39 of Calculation BNP-PSA-080, BNP Fire PRA - Quantification,fire scenarios were evaluated based on total CDF impact and importance of individual fire cutsets or groupings of fire cutsets. More detailed fire modeling was one of several approaches (e.g.,

crediting conditional circuit failure probabilities or operator recovery actions) available to remove excessive conservatism and thereby to achieve more realistic risk results. The determination to select and apply more detailed fire modeling tools was based on expert opinion of the expected resulting improvement. If the chosen approach resulted in a less-conservative, more-realistic risk, then the risk contributions of that scenario diminished and the next most important cutset could be subjected to the process. The process also provided for possible plant modification to reduce the risk for the scenario to acceptable levels when conservatism was not either readily apparent or easily removed. Although repeated iterations tend to produce diminishing results, this process stopped upon meeting the goal of a FPRA model that provides realistic risk estimates with some reasonable margin to the requirements of RG 1.174.

PRA RAI 2 Other than the UAM identified in Section 4.8.3.1, were any other UAMs or deviations from NUREG/CR-6850 used? If so, identify and describe those methods and clarify whether guidance from the June 21, 2012, memo from Joseph Glitter 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-6850 that fall outside this guidance memo, provide a sensitivity study that estimates the impact of their removal on CDF, LERF, ACDF, and ALERF.

Response

The only identified UAM is the UAM detailed in Section 4.8.3.1 of the LAR.

The only items identified as possible deviations from NUREG/CR-6850 and NUREG/CR-6850, Supplement 1 are shown in the following table, along with the reference for the associated sensitivity study:

Enclosure 2 Page 46 of 57 Identified Deviation Sensitivity Study Addressed In Treatment of smoke damage Response to PRA RAI 11 Treatment of sensitive electronics Response to PRA RAI le Credit taken for incipient detection Section 4.8.3.6 of LAR supplement Implementation of weighting factors in ignition Response to PRA RAI 3D frequency calculations (no 50s used)

The guidance from the June 21, 2012, memo from Joseph Glitter to Biff Bradley was not used.

PRA RAI 3 NUREG/CR-6850 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-6850 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), FC305 (Reactor Building Control Rod Drive Repair Room), and FC356 (Reactor Building Skimmer Surge Tank Room Vault).

d) Given that a weighting factor of "50" was not used in any fire area, provide a sensitivity study that assigns weighting factors of "50" per the guidance in FAQ 12-0064.

Response

a. The method used to calculate hot work and transient fire frequencies used the guidance in Section 6.5.7 of NUREG/CR-6850 to calculate influence factors, as described in Section 3.4.6.2 of BNP-PSA-083, and apply those factors to ignition frequencies, as described in Section 9.4.7 of BNP-PSA-086. When assignment of a maintenance weighting factor of "50" for a "very high" maintenance area was considered, no area with "significantly higher-than-average" maintenance was clearly evident.

b. Administrative controls were not used to reduce transient fire frequency.

c. As described in Section 3.4.6.2 of BNP-PSA-083, zero was assigned to all three influence factors for the MSIV pit in each Reactor Building (i.e., Fire Compartments FC296 and FC346) because, as a result of the plant design, the associated radiation levels preclude entrance while at power. As listed on Attachment 4 for BNP-PSA-083, three was assigned to all three influence factors for the Reactor Building Control Rod Drive Repair Room (i.e.,

FC305). However, zero was assigned to both the Maintenance and Storage influence

Enclosure 2 Page 47 of 57 factors for the Skimmer Surge Tank Vault in each Reactor Building (i.e., FC306 and FC356) because, by design, the vaults contain no equipment requiring maintenance and are sealed with a concrete plugs for radiation concerns (i.e., Attachments 2 and 17 of BNP-PSA-083).

d. Included with the other sensitivity studies in the 120-day RAI responses will be a sensitivity study that assigns a weighting factor of 50 for an area of "very high" maintenance or hotwork as described in Table 6-3 of FAQ 12-0064.

PRA RAI 6 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 Response of BNP-PSA-080 contains information about the contents of the Main Control Board (MCB) and MCR panels, including the "back-panel" area. Although this information was collected for use in the quantification effort, the statement "Any fire that does not result in an automatic or manual shutdown and that is contained within a panel would be treated as a non-event by the FPRA .... " was not used, as described in Section 3.2.4.3 of BNP-PSA-080.

Instead, all MCR fires were assumed to cause an automatic or manual shutdown.

MCR cabinets, including cabinets in the "back-panel" area, were modeled by splitting each of the cabinet fires into a "self' scenario, in which only the contents of the cabinet are damaged, and ZOI scenarios, in which both the content of the cabinet and external targets are damaged.

Scenarios which did not damage PRA components were not evaluated.

Similarly, MCBs which do not have incipient detection were modeled by splitting the fire into both "self' and ZOI scenarios. For MCBs with incipient detection, only the ZOI scenarios were modeled and evaluated, because the "self' scenarios were assumed to be negligible.

a) A fire in an MCR cabinet that was identified as being "closed" was considered to be "contained" and not to propagate to adjacent cabinets or external targets. None of the MCBs were considered to be "closed." PRA RAI 1G provides more detail on "closed" cabinets. MCB and MCR cabinets, in general, were assumed not to propagate to adjacent cabinets due to the construction of the cabinets, having an air gap separating the wall of one cabinet from the wall of the next. The exception to this is when cabinets were identified in walkdowns as having significant communicative space between the cabinets as documented in Attachment 17 of Calculation BNP-PSA-086, Fire Scenario Data. In these cases, the cabinet fires were assumed to propagate to all the adjacent cabinets which had significant communicative space between them.

Enclosure 2 Page 48 of 57 b) The response to PRA RAI 4 describes the methodology used for placing transient fires.

This methodology was applied in the MCR. The MCR does not contain any open-backed panels.

c) All of the in-cabinet ionization smoke detectors identified in Attachment 6 will be replaced with incipient detectors, as described by Item #1 in Table S-1 of the License Amendment Request. No credit was taken for the ionization smoke detectors in the baseline results, as credit was instead taken for the to-be-installed incipient detectors.

Credit was taken only for the ionization smoke detectors in the sensitivity study for the credit of incipient detectors.

PRA RAI 7 Fire-induced instrument failure should be addressed in the HRA per NUREG/CR-6850 and NUREG-1921. Describe how fire-induced instrument failure (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.

Response

Fire induced instrument faults were modeled for failure of indication required by operators for performing human actions credited in the FPRA. As described in Attachments 4 and 9 of Calculation BNP-PSA-084, BNP Fire PRA - Human ReliabilityAnalysis, instruments that are required for successful performance of each operator action credited in the PRA were subjected to a detailed analysis. Collectively, the required instruments provide diverse and redundant indications of plant conditions. Various reactor pressure vessel level and pressure indications are included to address inventory control and pressure control actions. Likewise, various containment pressure and temperature indications are included to address decay heat removal and containment pressure control actions. Also, system flow and pressure indications are included to address the need to start additional trains of standby equipment. Because fault tree modeling of this instrumentation is used to determine whether the minimum set of required indications is unaffected by fire, the HRA can credit the associated action with a Human Error Probability determined by shaping factors unrelated to reduced redundancy in required indication.

As described in Attachments 9 and 12 of Calculation BNP-PSA-085, BNP Fire PRA -

Component Selection, instrumentation identified as required in the HRA was included in component selection and modeled explicitly in the fault tree. This included the related power supplies. The associated cables were routed. However, no circuit analysis was performed because the various possible instrument failure modes (e.g., no reading, off-scale reading, incorrect/misleading reading) are not differentiated in the fault tree model.

No credit is given for instrumentation affected by fire, regardless of failure mode. If an associated cable is determined to be damaged by a fire scenario, the instrument is considered to be failed and the associated basic event is assigned a failure probability of 1.0.

The fault tree model includes basic events for credited operator actions under an OR gate with the appropriate logic for the applicable minimal required instrumentation. Therefore, for a fire

Enclosure 2 Page 49 of 57 scenario that fails more than the minimum required indication, the instrument failures show up in the minimal cutset, and the operator action is not credited.

PRA RAI 11 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, ACDF, and ALERF.

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.

Response

a) The method used to locate transient fires is described in the response to PRA RAI 4 and encompasses the guidance provided in Section 11.5.1.6 of NUREG/CR-6850. Based on this method, no transient fire was postulated as being located in the region of the MCR where operators manipulate controls on the MCBs because no target (i.e., cables) was identified within the postulated zone of influence. However, four transient fires were postulated to be located in the MCR, and these four transient locations were included in the evaluation of MCRA times for habitability concerns, as described in Attachment 16 of Calculation BNP-PSA-080, BNP Fire PRA - Quantification.The FPRA does not credit MCRA for loss of control. During the evaluation of the risk for MCRA, these four transient fires were not considered to contribute measurably to the overall risk, as described in Section 3.0 of Attachment 16 to Calculation BNP-PSA-080. This simplification was based on a comparison of the total ignition frequency of about 1.7E-6 yr-1 (i.e., Attachment 6 of Calculation BNP-PSA-080) for the four transient sources to the total ignition frequency of about 7.5E-3 yr 1 for the cabinets (i.e., Tables D, E, and F of Attachment 16 of Calculation BNP-PSA-080) in the MCR.

Based on this analysis of ignition frequencies, the effect of the transient fires on the estimated risk associated with MCRA is not considered to be measureable, and no further sensitivity analysis is considered necessary.

b) As detailed in Attachment 16 of Calculation BNP-PSA-080, whether single-bundle or multi-bundle was used for MCR cabinets was determined from the source walkdowns for all cabinets in the MCR. However, as a special case, the MCBs were treated as a single-bundle, rather than a multi-bundle, because the evaluation of the risk for MCRA is relatively simplistic and was considered to produce overly conservative results for the MCBs. A more realistic treatment would include credit for the close and constant proximity of reactor operators, the expected slow growth resulting from low energy ignition sources inside the cabinets, and a high probability of fire detection by the control staff such that early

Enclosure 2 Page 50 of 57 suppression will occur before a large multi-bundle fire could develop. Consequently, to achieve more realistic results within the parameters of the analysis described in Attachment 6, the MCBs were approximated as single-bundles.

A sensitivity study which evaluates the treatment of the MCBs as multi-bundle fires will be submitted in the 120-day responses, with the results of other sensitivity studies.

PRA RAI 12 Attachment W of the LAR provides the ACDF and ALERF for the VFDRs and the additional risk of recovery actions for each of the fire areas, but the LAR does not describe how ACDF and ALERF or the additional risk of recovery actions were calculated. Describe the method(s) used to determine the changes in risk reported in 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 ACDF and ALERF 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).

Response

The scope of the FPRA Peer Review was the base FPRA model and included none of the model manipulations described in FAQ 08-0054 for evaluating the changes in risk reported in Tables W-4-1 and W-4-2. No exception to normal modeling mechanisms was used to calculate the changes to risk reported in Tables W-4-1 and W-4-2.

Section W.2.1 of the LAR describes, in general, the method used to determine the changes in risk reported in Tables W-4-1 and W-4-2. As described in Calculation BNP-PSA-082, BNP Fire PRA - NFPA 805 Transition Support, and consistent with the guidance provided in FAQ 08-0054, the method assumes that compliance with the deterministic requirements eliminates the associated failure/deficiency, such that the change in risk is always a risk reduction. The aggregate for each fire area was reported in Tables W-4-1 and W-4-2.

VFDRs were evaluated by manipulating the fire PRA model to remove the potential fire induced failure associated with the VFDRs. VFDRs associated with cables were evaluated by removing associated cables from the fire damage sets. VFDRs related to a lack of automatic area wide suppression were evaluated by adding suppression to the required fire areas. VFDRs related to a lack of train separation were evaluated by assuming all train "A" safe shutdown cables are cable VFDRs. There was no significance in picking train "A" cables over train "B" cables.

The MCR is encompassed within Fire Area CB-23E. The calculation of ACDF and ALERF in the MCR differs from the description above only with respect to the treatment of MCRA. MCRA was evaluated with an external assessment from the FPRA model. The risk of recovery actions associated with MCRA was conservatively added to the total risk of transition. As described in

Enclosure 2 Page 51 of 57 the response to PRA RAI 15A, Tables W-4-1 and W-4-2 report "N/A" for the additional risk of recovery actions for all fire areas except CB-23E because the only recovery actions credited in the FPRA are those associated with MCRA for habitability concerns. Since no recovery action is credited by the FPRA for loss-of-control scenarios, no corresponding calculation of ACDF and ALERF was performed. The calculation of the risk for the recovery actions associated with MCRA is described in the response to PRA RAI 14.

PRA RAI 13 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.

Response

Asymmetry of CDF and LERF results exists between Unit 1 and Unit 2 primarily because the FPRA is very sensitive to the spatial relationships between the sources and targets, and that degree of spatial symmetry does not exist between the two units.

The cable routings are not symmetrical between the two units. Raceway routings are not symmetrical between the two units. Control room panels that provide similar functions are not always symmetrically located. Some shared functions are designated as Unit 2 systems. This leads to seemingly parallel scenarios having different equipment damage sets and, by extension, different risk results.

Consequently there is no expectation for symmetry in the FPRA results between the units.

PRA RAI 14 Explain how the additional risk of recovery actions was determined for abandonment scenarios.

Response

As described in Attachment 14 of Calculation BNP-PSA-082, BNP Fire PRA - NFPA 805 Transition Support, the additional risk of recovery actions was estimated as the risk reduction associated with assuming 100% success of the recovery actions in a detailed Human Reliability Analysis for Main Control Room Abandonment (MCRA). This HRA is described further in the response to PRA RAI 1 F. The HRA for MCRA estimated a Human Error Probability that represents the Conditional Core Damage Probability (CCDP) following MCRA and is described in Attachment 10 of Calculation BNP-PSA-084, BNP Fire PRA - Human ReliabilityAnalysis. In 4 of Calculation BNP-PSA-082, Conditional Large Early Release Probability (CLERP) was estimated as 10% of CCDP.

To estimate the frequency for MCRA, the sum of ignition frequencies for sources in the control room was multiplied by the applicable probability distribution for non-suppression of fires resulting in MCRA. This probability distribution was determined by applying the guidance for non-suppression probability in Appendix P of NUREG/CR-6850 to the MCRA times for fire size

Enclosure 2 Page 52 of 57 distributions determined in an evaluation of MCRA times for habitability. This evaluation of MCRA times considered many possible configurations of control room heating, ventilation, and air conditioning (HVAC) and boundary door status. The determination of frequency for MCRA conservatively selected a time distribution based on the control room boundary doors remaining closed during the fire and assumed 1% unavailability for the control room HVAC system.

MCRA due to loss of control was not credited in the FPRA. Therefore, only MCRA for habitability concerns was considered when estimating the additional risk of recovery actions.

The HRA for MCRA includes several actions, of which only those actions that meet the following three requirements were considered recovery actions:

1) Not performed in the main control room in preparation for control room abandonment,

2) Not performed while in transit to one of the four stations that support the Remote Shutdown Panel, and

3) Not performed at the Remote Shutdown Panel.

The CCDP and CLERP for MCRA were re-evaluated with the failure probabilities for the recovery actions set to zero. The new lower CCDP and CLERP were then used to determine the delta risk as the additional risk of the recovery actions.

PRA RAI 15 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, SWI-1, and TB-1).

b) Why a "below truncation" value is reported in the ACDF/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 ACDF/LERF column for fire areas with VFDRs (i.e.,

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

Response

a. The FPRA credited recovery actions only for those that reduce the risk associated with certain fire scenarios in Fire Area CB-23E and, in particular, for those fire scenarios associated with MCR abandonment for habitability concerns. For all other fire areas, Tables W-4-1 and W-4-2 report "N/A" for the additional risk of recovery actions, because possible risk reductions for other recovery actions were not quantified. The listing of recovery actions in Attachment G of the LAR includes both those credited for risk reduction and those not credited for risk reduction.

b. For deterministic fire areas, there is no reason to report a "below truncation" value rather than "N/A" in the ACDF/LERF column.

As part of the 120-day RAI responses, Tables W-4-1 and W-4-2 will be updated to show "N/A" in the Fire Risk Evaluation column when "No" is listed in the VFDR column.

Enclosure 2 Page 53 of 57

c. For fire areas with VFDRs, a zero value in the ACDF/LERF column means that cutset results were obtained for at least some associated fire scenarios for both the base FPRA model and the VFDR model but that, at the truncation level of the analysis, there was no difference between the total CDF/LERF.

PRA RAI 16 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 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.

Response

The following plant modifications are identified in Table S-1 as in the FPRA, but have not been completed. These modifications have been directly, or indirectly, credited for certain Fire Areas in the change-in-risk estimates provided in Table W-4-1 and Table W-4-2.

I Item I Proposed Modification Include specified in-cabinet incipient detection.

1-XU-1 2-XU-1 1-XU-2 2-XU-2 1-XU-3 2-XU-3 1-XU-4 2-XU-4 1-XU-51 2-XU-51 1-XU-80 2-XU-80 2-XU-69 1-H12-P601 2-H12-P601 1-H1 2-P603 2-H1 2-P603 5 Provide ERFBS Wrap for 61L1/BA above source 4521 (1-1L1).

Provide modification to change the valves to a failsafe condition of closed on loss of valve support for Hotwell Makeup Valves: 1-CO-LV-1-2, 2-CO-LV-1-2 7 Provide separation / protection of the following cables in the Main Control Room:

1JG8-WI3, 1AQ6-14B, 1AQ6-IJ1, 2AQ6-HZ3 The remaining plant modifications in Table S-1 have been identified as not in the FPRA and have been neither directly, nor indirectly, credited in the change-in-risk estimates provided in Tables W-4-1 and W-4-2. Those plant modifications that are identified in Table S-1 as not in the FPRA are considered to have no adverse impact of the functionality of the equipment. Item #11 and Item #12 described modifications that might have the potential for some risk reduction and, if incorporated into the FPRA, would not invalidate the change-in-risk estimates provided in Tables W-4-1 and W-4-2.

Enclosure 2 Page 54 of 57 The following implementation item will be added to Table S-2 in the LAR:

Item Unit Description LAR Section / Source Subsequent to the completion of modifications which have been identified in Table S-1 as in the FPRA, the change-in-risk estimates provided in Tables W-4-1 and W-4-2 will be validated.

9 1, 2 If the as-built change-in-risk for a Fire Area exceeds Attachment W the estimates reported in Table W-4-1 or Table W-4-2, the responsible feature will be identified and evaluated per the post transition change process per Section 4.7.2 of the LAR.

Enclosure 2 Page 55 of 57 Fire Modeling Requests for Additional Information Fire Modeling RAI 1B 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:

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.

Response

The statement cited in the RAI, "In spaces where the compartment height varies with position, the maximum height is assumed since this maximizes the entrainment," was not intended to suggest that relatively large portions of the Control Room or separate rooms within the Control Room complex have different or varying ceiling heights. Instead, the statement was intended to clarify that in situations where localized irregularities project lower than the room ceiling height are present, such as beams, HVAC ducts, etc., the ceiling height was used as the CFAST input.

In other words, fires in CFAST are not modeled under localized irregularities, such as ducts or beams. For the purposes of MCR abandonment calculations, these irregularities are minimal, and would not affect the results.

Calculation BNP-PSA-082, Attachment 14 (i.e., Evaluation of Main Control Room Abandonment Times at the Brunswick Nuclear Plant,Revision 2), lists the ceiling heights for each of the enclosures modeled in CFAST (i.e., see Table 2-1). The source for these heights is drawing BNP General Arrangement drawing F-07009. Sections A-A and C-C in this drawing show that there are no relatively large portions of the Control Room with different ceiling heights. That is, there are no prominent obstructions or irregularities in the ceiling height affecting the fire modeling in CFAST as performed and documented in the calculation.

Enclosure 2 Page 56 of 57 Fire Modeling RAI 2C 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."

Provide the following information:

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.

Response

Section 3.2.3.4.2 of Calculation BNP-PSA-080, Revision 3, BNP Fire PRA - Quantification, states the following:

Only cable trays and conduits were used for the quantification, the equipment targets were not explicitly failed for the following reasons. While relay cabinets and some sensitive equipment may have lower damage thresholds resulting in non-conservative damage sets, the temperatures related to electrical equipment failures are generally based on long term environmental qualification and not short term functionality. In addition, when equipment is in the ZOI, often the cables to or from that equipment is also in the ZOI such that the equipment failure is captured via the cable tray and conduit targets.

Consistent with this discussion, the damage thresholds assigned to cable tray and conduits are those listed in Appendix H of NUREG/CR-6850 for thermoset cables as described in responses to 60-day RAIs Fire Modeling RAI 2A and Fire Modeling RAI 2B. In addition, the BSEP Fire PRA did not assign different damage thresholds.

Damage to plant components other than cables, i.e., components in the FPRA equipment list that can be damaged by fire, are classified in two groups:

" Active components (i.e., mostly electrical components): Examples of these components include electrical cabinets, valves, pumps, etc. Per guidance in Appendix H of NUREG/CR-6850, it is assumed that the vulnerability of these components is governed by the cables connecting to them. Therefore, it is assumed that the components will fail consistent with the damage criteria for thermoset cables.

• Passive components (i.e., check valves, tanks, etc.): Per guidance in Appendix H of NUREG/CR-6850, these components are assumed not to be damaged by fire and no damage thresholds have been assigned to them.

Enclosure 2 Page 57 of 57 Fire Modeling RAI 5C Regarding the qualification of users of engineering analyses and numerical models:

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.

Response

Throughout the NFPA 805 transition process, the fire protection engineers who conducted the fire modeling and the PRA engineers have maintained frequent communications and worked together developing the necessary data, documentation, and quantification infrastructure.

Specifically:

• The fire PRA project instructions (e.g., FPIP-0150, Rev 1; FPIP-0200, Rev. 8; FPIP-0208, Rev. 5) relevant to fire modeling tasks were developed and are maintained jointly by the Fire Protection and PRA groups.

• Based on the above referenced set of project instructions, fire protection and fire modeling personnel conducted walk downs and populated the data tables associated with ignition source information and plant fire protection system and features information.

This information served as input to PRA engineers to perform with risk quantification activities.

• Fire modeling personnel also prepared and maintained the calculations supporting the fire PRA analysis quantification workbooks. Frequent discussions between Fire Protection and PRA, and the availability of a consistent set of project instructions on the specific details of the quantification workbooks, ensured that the information necessary to support the quantification process is provided.

• Both fire modeling/fire protection engineers and the PRA engineers participated in the cut-set review meetings during the development of the fire PRA. Participation in cut-set reviews facilitated the identification of fire modeling refinements (i.e., detailed analyses) necessary for supporting risk quantification activities.

The above described process will continue during transition implementation and future established activities as it is based on procedures and a systematic fire PRA methodology that is consistently applied throughout the fleet of nuclear plants.