CNS-15-001, License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants, 75-Day Response to NRC Request for Additional...

License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants, 75-Day Response to NRC Request for Additional...

ML15015A409
Person / Time
Site:	Catawba
Issue date:	01/13/2015
From:	Henderson K Duke Energy Carolinas, Duke Energy Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNS-15-001, TAC MF2936, TAC MF2937
Download: ML15015A409 (43)

Text

Kelvin Henderson Vice President Catawba Nuclear Station tENERGY.

Duke Energy CNO1VP 1 4800 Concord Road York, SC 29745 CNS-15-001 o: 803.701.4251 f: 803.701.3221 January 13, 2015 10 CFR 50.90 U.S. Nuclear Regulatory Commission (NRC)

Attention: Document Control Desk Washington, D.C. 20555

Subject:

Duke Energy Carolinas, LLC (Duke Energy)

Catawba Nuclear. Station, Units 1 and 2 Docket Numbers 50-413 and 50-414 License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants 75-Day Response to NRC Request for Additional Information (RAI)

(TAC Nos. MF2936 and MF2937)

References:

1. Letter from Duke Energy to the NRC, "License Amendment Request (LAR) to Adopt National Fire Protection Association (NFPA) 805 Performance-Based Standard for Fire Protection for Light-Water Reactor Generating Plants", dated September 25, 2013 (ADAMS Accession Number ML13276A503)

2. Letter from the NRC to Duke Energy, "Catawba Nuclear Station, Units 1 and 2: Request for Additional Information Regarding License Amendment Request to Implement a Risk-Informed, Performance-Based Fire Protection Program (TAC Nos. MF2936 and MF2937)",

dated November 20, 2014 (ADAMS Accession Number ML14308A037)

The Reference 1 letter requested NRC review and approval for adoption of a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants",

Revision 1, dated December 2009. This LAR was developed in accordance with the guidance contained in Nuclear Energy Institute (NEI) 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)",

Revision 2.

The Reference 2 letter transmitted RAIs necessary for the NRC to continue its review of the Reference 1 LAR.

14 (Do ýO www.duke-energy.com

Document Control Desk Page 2 January 13, 2015 These RAls were discussed with the NRC in draft form during the NRC's audit of the subject LAR in Duke Energy's corporate office on October 27-30, 2014. During the audit, each RAI was characterized as requiring a 60-day, 90-day, or 120-day response.

Because the due date for the 60-day responses would have fallen in late December 2014, Duke Energy and the NRC verbally agreed that these responses would be provided by January 13, 2015 (75 days). The purpose of this letter is to provide the docketed response to the 75-day RAIs. The enclosure to this letter provides this response. The format of the enclosure is to restate each RAI question, followed by its associated response.

Note that any LAR revisions necessary as a result of any RAI responses (75-day, 90-day, or 120-day) have been entered into the Catawba corrective action program and will be included in the submittal providing the 120-day RAI responses.

The conclusions of the No Significant Hazards Consideration and the Environmental Consideration contained in the Reference 1 letter are unaffected by this RAI response.

There are no regulatory commitments contained in this letter or its enclosure.

Pursuant to 10 CFR 50.91, a copy of this LAR supplement is being sent to the appropriate State of South Carolina official.

Inquiries on this matter should be directed to L.J. Rudy at (803) 701-3084.

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

Executed on January 13, 2015.

Very truly yours, Kelvin Henderson Vice President, Catawba Nuclear Station LJR/s Enclosure

I Document Control Desk Page 3 January 13, 2015 xc (with enclosure):

V.M. McCree Regional Administrator U.S. Nuclear Regulatory Commission - Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, GA 30303-1257 G.A. Hutto, III Senior Resident Inspector (Catawba)

U.S. Nuclear Regulatory Commission Catawba Nuclear Station G.E. Miller (addressee only)

NRC Project Manager (Catawba)

U.S. Nuclear Regulatory Commission One White Flint North, Mail Stop 8 G9A 11555 Rockville Pike Rockville, MD 20852-2738 S.E. Jenkins Manager Radioactive and Infectious Waste Management Division of Waste Management South Carolina Department of Health and Environmental Control 2600 Bull St.

Columbia, SC 29201

Enclosure Response to 75-Day NRC RAIs 1

REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414 By letter dated September 25, 2013, (Agencywide Documents Access and Management System (ADAMS) Accession. No. ML13276A503), Duke Energy Carolinas (Duke) submitted a license amendment request (LAR) to change its fire protection program to one based on the National Fire Protection Association (NFPA) Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, as incorporated into Title 10 of the Code of FederalRegulations (10 CFR), Part 50, Section 50.48(c). In order for the NRC staff to complete its review of the LAR, the following additional information is requested.

Fire Protection Engineering (FPE) Request for Additional Information (RAI) 01 LAR Attachment A, Section 3.4.1(c) states compliance with NFPA 805 Section 3.4.1(c) which requires that the fire 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 criteria.

Provide additional detail regarding the training provided to the fire brigade leader and members that addresses their ability to assess the effects of fire and fire suppressants on NFPA-805 nuclear safety performance criteria. Include the justification for how the training meets NFPA-805 Section 3.4.1.

Duke Energy Response:

Catawba utilizes 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 (Section 3.4.1(c)) and procedure AD-EG-ALL-1530, Fire Brigade Training. Note that the information in NSD 112 as referenced in LAR Attachment A, Section 3.4.1(c) has been superseded by AD-EG-ALL-1530.

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An equivalent knowledge of plant systems is provided for under procedure AD-EG-ALL-1530, Fire Brigade Training, Section 5.5, 5.6 and Attachment 3. Attachment 3 specifies the plant systems for either a PWR or BWR that represent the minimum plant knowledge for a Non-Licensed Operator (NLO) fire brigade member or leader to understand the effects of fire and fire suppressants on nuclear safety performance criteria (ref. NFPA 805 Section 3.4.1(c)). Specifics from AD-EG-ALL-1530 are as follows:

5.5 Fire Brigade Leader Training

1. The Fire Brigade Leader is required to complete:

• Fire Brigade Leader training

• An in-plant fire drill as fire brigade team leader

• Fire Brigade Leader qualification

2. In addition to the fire brigade member training and drill requirements (defined in this procedure), Fire Brigade Leader training should address and emphasize the following objectives:

• Command structure

" Roles of Fire Brigade Leader and Fire Chief

• Command of a fire brigade - Concepts of Incident Command

• Incident Safety Officer Responsibility

• Organizational setup for emergency response

• Coordination of off-site fire company using ICS

• Managing resources

• Pre-incident Information

• Communications

" Site Fire Brigade Standard Operating Guidelines

" Fire brigade checklist and drill review

" Firefighting survival, strategy, and tactics

" Site fire strategies

• Fighting of fires in RCAs and RCZs and potential Radioactive Material Releases

• Case studies of fire incidents (OE)

" Fire Brigade Leader practical to include special problems with fire suppression and leadership 5.6 Safety Systems Training

1. The FBL and at least two other members of the brigade shall have knowledge, training, and understand the effects of fire and fire suppressants on safe shutdown equipment. This training and knowledge may be satisfied by:

a. Completing and maintaining a fire brigade qualification and meeting one of the following requirements:

(1) Be a licensed Senior Reactor Operator.

(2) Be a licensed Reactor Operator.

(3) Have successfully completed a reactor operator certification program.

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W (4) Have successfully completed training on the plant systems listed in Attachment 3, Fire Brigade Safety Systems Individual System Knowledge List.

From Attachment 3:

Fire Brigade Safety Systems Individual System Knowledge List PWR BWR Reactor Coolant System 2035 - Core Spray System Steam Generator System 2040 - Standby Liquid Control System Auxiliary Feed System 2045 - Residual Heat Removal System Charging and Volume Control System 2070 - Containment Atmosphere Control Residual Heat Removal System 2095 - High Pressure Coolant Injection Safety Injection System 2100 - Reactor Core Isolation Cooling Containment Spray System 4060 - Service Water System Component Cooling Water System 4070 - Rx Bldg Closed Cooling Water Emergency Service Water System 5095 - Diesel Generator System Electrical System Overview - AC & DC 5097 - Supplemental DG System Emergency Core Cooling Systems 5098 - SAMG Diesel Generator System 5100 - Diesel Generator Fuel Oil System 5135 - 230KV Switchyard 5145 - SAT, UAT Transformer System 5170 - 4KV AC Distribution 5175 - 480V AC Distribution 5230 - 250V DC Distribution 6175 - Fire Protection System 6195 - Fire Protection C02 System 6205 - Halon Supply System 7071 - Standby Gas Treatment System 7110 - Fuel Pool Cooling System 8075 - Diesel Building HVAC 8185 - Reactor Building HVAC 8220 - Control Building HVAC 8232 - Service Water Building HVAC 4

FPE RAI 03 LAR Section 4.1.3 and LAR Attachment I, "Definition of Power Block" states that structures required to meet the radioactive release criteria described in Section 1.5 of NFPA-805 but not required to meet the nuclear safety criteria are not defined as power block. Currently, the endorsed guidance of NEI 04-02 states that, where used in Chapter 3, "power block" and "plant" refers to structures that have equipment required for nuclear plant operations, such as containment, auxiliary building, service building, control building, fuel building, radiological waste, water treatment, turbine building, and intake structure, or structures that are identified in the facility's current licensing basis. As currently described in the LAR Attachment E, the following compartments are not described as part of the power block in Attachment I, Table I-1:

" Radioactive Material Containers Area

" Radiography Vault

" Radiation Materials Control Building (#7767)

"Tents Containing Radioactive Materials

• Mixed Waste Storage Provide clarification whether these structures listed are accounted for as either within or not within the power block, as described in LAR Section 4.1.3.

Duke Energy Response:

The Catawba definition of power block was developed based on the guidance provided in FAQ 06-0019 Revision 4 (ML073060545) and the NRC closure memo to FAQ 06-0019 (ML080510224). The Catawba definition of power block was developed with respect to those structures required to meet the NFPA 805 nuclear safety performance criteria. This includes structures with the potential to affect power plant operations, the potential to affect equipment important to nuclear safety, and the potential to affect the ability to safely shutdown the plant in the event of a fire.

The NRC closure memo includes a chronological history of the development of FAQ 06-0019 that supports this definition. The NRC Staff Comments on FAQ 06-0019, Revision 3 states:

"The letter and the intent of the NFPA 805 definition for "powerblock" and "plant"

("Structures that have equipment requiredfor nuclearplant operations")includes all equipment needed to generate electricity (main turbine, feedwater, circulating water, service water, main steam, etc.) as well as that equipment needed to mitigate accidents requiredby the Technical Specifications (safety injection, emergency diesel generators,containment spray, emergency service water, etc.)."

The Catawba definition of Power Block as found in LAR Attachment I, Table I-1 is consistent with the NRC Staff Comments on the definition of power block.

Referring to the structures listed in the guidance (FAQ 06-0019), the Catawba power block includes containment (reactor buildings), auxiliary building (including control 5

t.

complex and fuel buildings), service building, diesel generator building, turbine building, intake (and discharge) structures, in addition to structures specific to Catawba - the doghouses, Standby Shutdown Facility (SSF), and yard areas where equipment required to meet the nuclear safety performance criteria goals is located.

Radiological material/waste areas are not included in the Catawba power block definition.

There are a series of structures identified in LAR Attachment E, Radioactive Release Transition, which are not identified in the Catawba power block. The Catawba radiological release areas are not included based on the guidance in the NRC Closure memo in that the areas are not needed to generate electricity or needed to mitigate accidents as required by the Technical Specifications. These specific Catawba areas include:

" Radioactive Material Containers Area (located in the Yard)

" Radiography Vault (located in the Yard)

" Radiation Materials Control Building (Building #7767 - located in the Yard)

" Tents Containing Radioactive Materials (located in the Yard)

" Mixed Waste Storage (located in the Yard)

Scoping of Radioactive Release compartments used a different criteria than the Power Block scoping criteria. The Radioactive Release compartment scoping criteria was based on all plant areas with radiologically controlled areas.

FPE RAI 04 LAR Attachment A, Section 3.11.5 states that Catawba Nuclear Station (CNS) does not utilize Electrical Raceway Fire Barrier Systems (ERFBS) for Chapter 4 compliance. ERFBS are fire barrier materials such as Thermo-Lag, 3M Interam, Hemyc, MT, or Darmatt systems. In LAR Attachment C, Table B-3, Hemyc is cited in a number of applications used as an ERFBS. On June 7, 2006, the licensee submitted their response to Generic Letter 2006-03, "Potentially Nonconforming Hemyc and MT Fire Barrier Configurations," (GL 06-03) and committed to bring their Hemyc issues into full compliance through their NFPA-805 transition process, however, the LAR does not provide a summary of this resolution.

a. Provide a description of how CNS resolved the GL 06-03 issue through the NFPA-805 transition process, including any proposed plant modifications. If performance-based methods are used, include a discussion of the risk, safety margin, and defense-in-depth (DID) considered in the evaluation.

Duke Energy Response:

Catawba originally relied upon Hemyc to protect safe shutdown circuits in the Unit 1 and Unit 2 Auxiliary Feedwater Pump rooms (Fire Areas 02 and 03). Catawba resolved GL 06-03 by no longer crediting the Hemyc wrap in these areas as part of the NFPA 805 analysis. The Catawba Fire Risk Evaluation report states "No credit is taken for HEMYC wrap on any cables in the Fire PRA." The VFDR's that discuss Hemyc are identified in LAR Attachment C, Table C-1 (NEI 04-02 Table B-3) in Fire 6

Areas 02 and 03. Two VFDR's (2-VFDR-08 and 3-VFDR-07) cite Hemyc and an accompanying modification. The non-credit of the Hemyc results in potentially non-coordinated loads. The associated modifications are for resolution of the non-coordinated loads and not specifically related to resolution of the Hemyc ERFBS. The other VFDR's that involve circuits previously protected by Hemyc were demonstrated to satisfy risk, DID, and safety margin with no further actions in the Catawba Fire Risk Evaluations. In summary, no Hemyc is required by NFPA 805 analysis thereby resolving the GL 06-03 issue.

b. Provide a description of any credited Hemyc fire barriers used for the Nuclear Safety Capability Assessment (NSCA). Provide the basis for barriers' credited rating as an ERFBS or any other credited uses.

Duke Energy Response:

There are no credited ERFBS applications being carried forward in the NFPA 805 analysis.

FPE RAI 05 In LAR Attachment A, Table B-i, Section 3.3.5.3, the LAR indicates that electrical cables comply with IEEE-383 flame propagation testing (Institute of Electrical and Electronics Engineers Standard 383, "IEEE Standard for Type Test of Class 1 E Electric Cables, Field Splices, and Connections for Nuclear Power Generating Stations").

During the audit, the licensee explained that the armored cable control circuit test program conducted by the licensee in 2006 included tests on cable qualified for use in Duke nuclear power plants. The 2006 test program involved comparative tests on cables with and without an outer PVC jacket. The comparative tests indicated that removing the outer PVC jacket may accelerate the horizontal flame spread over the cable. Describe how the requirements of NFPA-805 Section 3.3.5.3 are met.

Duke Energy Response:

The unjacketed cable used at Catawba exhibits flame spread and fire propagation characteristics consistent with cable types considered IEEE 383 equivalent. There were some unanticipated observations made during a series of cable tests performed in 2006, at the Intertek Testing Services NA, Inc. facility in Elmendorf, Texas (formerly Omega Point Laboratories, Inc.). One of these unexpected observations is the "flame propagation concern." It is important to note that this series of tests was conducted to evaluate fire-induced circuit failure, not flame propagation. Therefore, these tests were performed under conditions that were significantly more severe than testing required to meet IEEE 383 (IEEE Standard for Qualifying Class 1E Electric Cables and Field Splices for Nuclear Power Generating Stations) and IEEE 1202 (IEEE Standard for Flame-Propagation Testing of Wire and Cable).

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Calculation DPC-1435.00-00-0009, "Performance Characteristics of Duke Armored Cables Under Fire Exposure", is the AREVA Engineering Information Record document EIR 51-9160514-000, an analysis of the same name. As stated in this calculation:

In the fall of 2007, Duke requested that General Cable Corporation(GCC), the manufacturerof the Duke-specific armoredcable that was used in the Duke Armored Cable Control CircuitTests conducted in 2006, provide IEEE-1202 test results for both jacketed and unjacketed versions of the same armoredcable type used in some of the Duke Armored Cable Control Circuit Tests.

... These standardizedtest results indicate that both jacketed and unjacketed version of the armored 8-conductor#12AWG cable passed the IEEE-1202 cable flame propagationtest...

... The test results also indicate that there is essentiallyno difference in the flame propagationperformance of jacketed and unjacketed applicationsof this armored cable type when tested in accordancewith the IEEE 1202 standard.

Additionally, the IEEE 1202 test reportfor the jacketed version of cable type tested on 09/17/2007 (Duke GSI Armored 8-conductor#12 1KV-FRXLPE) noted that flame did in fact "shoot"out the bottom of the sample during the second half of the 20-minute flame exposure test. This indicates a similarphenomenon to what occurredin a few of the Duke Armored Cable Control CircuitTests conducted at Intertek Laboratoriesin 2006...

This calculation concludes:

... the "shootingflame" condition is not a result of flame propagationinternalto the armored cable. Fire testing shows that when flames have occurredat the ends of armoredcables, the mechanism that causes flame to occurat this location is attributedto a series of events and conditions that lead to hot gases and vapors traveling inside the armoredshielding from the vicinity of the fire exposure to the open ends of the cable where these gases ignite when mixed with availableair...

... Therefore, no additionalguidance or conservatism needs to be added to the FPRA based on the use of armored cable and the potentialfor the "shooting flame" condition to occur...

... In summary, the "shooting flame" phenomenon is not new to armoredcables.

Fire testing as far back as 1978 suggests that this condition has occurred under certain test parameters. Evidence of this phenomenon has also occurredat times during standardizedtests for determiningthe flame propagationcharacteristicsof armoredcables. In no instance has a testing authority ever deemed the results of a standardizedflame propagationtest for armoredcables unacceptabledue to this phenomenon.

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V Many of the armoredcable types use[d] at Duke nuclearpower generating stations have undergone standardizedtesting to determine theirflame spread propagationcharacteristics. These tests were typically performed in accordance with the IEEE 383 and/or IEEE 1202 test standards. The results of this testing confirmed that the armored cable types tested are IEEE 383 and/orIEEE 1202 "qualified"...

... There are standardizedtest methods for measuringand determininga cable's flame propagationqualities. IEEE 383 and IEEE 1202 are two such standards that have been deemed acceptable to the NRC. The armoredcable types used at Duke nuclearpower generatingstations have either been qualified to one of these standardsor are consideredequivalent by comparison.

Therefore, based on the information presented in the report that reviewed the Duke Energy fire tests, Catawba considers the unjacketed cables non-propagating/equivalent to IEEE 383 qualified cable and meet the requirements of NFPA 805 Section 3.3.5.3.

A revision to LAR Attachment A, Section 3.3.5.3 is planned to be submitted with the 120-day RAI responses to clarify Catawba cable is IEEE 383 or equivalent in accordance with flame propagation tests as outlined in FAQ 06-0022.

FPE RAI 07 NFPA-805 Section 3.11.3 requires that penetrations in fire barriers be provided with listed fire-rated door assemblies having a resistance rating consistent with the designated fire resistance rating of the barriers and that fire doors shall conform to NFPA-805, "Standard for Fire Doors and Fire Windows." LAR Attachment A, Table B-1 states that CNS complies with previous NRC approvals and the compliance bases only describe un-labeled and modified doors and pressure doors, as well as bullet- and missile-resistant doors. However, LAR Attachment K, Licensing Action 02 discusses the use of hollow metal doors in Fire Area 35 that are not rated and hollow metal doors with louvers in radiological areas. These specific doors are not discussed in LAR Attachment A, Table B-1 compliance bases. Provide justification for the apparent discrepancy between LAR Attachment A, Section 3.11.3 and LAR Attachment K Licensing Action 02.

Duke Energy Response:

The complete excerpt of the SER approval was not included in LAR Attachment A, Section 3.11.3. This was an oversight. The entire context of the SER approval includes the discussion of the hollow metals doors in Fire Area 35 which should have been included in LAR Attachment A, Section 3.11.3. LAR Attachment A, Section 3.11.3 will be revised to include both paragraphs from the SER identified in Licensing Action 02.

A revision to LAR Attachment A for this item is planned to be submitted with the 120-day RAI responses that will include this clarification.

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W.

Safe Shutdown Analysis (SSA) RAI 01 LAR Attachment B, Table B-2, identifies certain attributes of NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis" Revision 1, as "Aligns with Intent." For the following attributes, the alignment basis does not fully explain why CNS deviates from the recommendations of the attribute.

For each attribute, provide a more in-depth justification as to what specifically does not align. These include:

a. 3.1C Spurious Operation Duke Energy Response:

Catawba aligns with the guidance of considering spurious operations in both the selection of safe shutdown functions and systems as well as the cabling associated with the components relied upon to achieve those functions. Catawba aligns with the intent of the guidance for a loss of reactor pressure vessel/reactor coolant inventory in excess of the safe shutdown makeup capability for high/low pressure interfaces.

High/low pressure interfaces are limited to meeting the latest guidance in NEI 00-01, Revision 2. NEI 00-01 Revision I states that high/low pressure interfaces result in a LOCA. RG 1.189 Revision 2 Section 5.3.2.c endorses NEI 00-01 Revision 2, which expands the high/low pressure interface definition to a LOCA outside containment.

Catawba analyzed high/low pressure interfaces resulting in a LOCA outside containment.

b. 3.1.1.3 Use of Pressurizer Heaters Duke Energy Response:

For a main control room (MCR) safe shutdown no credit is taken for the use of pressurizer heaters as specified in GL 86-10, Enclosure 2, Section 5.3.5. Control of auxiliary feedwater and steam release is analyzed and assured. A shutdown from the SSF does not have all the systems as a MCR shutdown. The availability of one sub-bank of pressurizer heaters is desirable to assure subcooling is maintained and a bubble is not formed in the reactor vessel. This sub-bank of pressurizer heaters was analyzed as part of the equipment for the SSF shutdown areas and shown to be free of fire damage for fires in all areas requiring SSF shutdown. The analysis exceeds the guidance. Therefore, Catawba aligns with NEI 00-01.

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c. 3.1.1.7 Offsite power Duke Energy Response:

Catawba does not credit offsite power (OSP) in the deterministic analysis and therefore does not demonstrate it to be free of fire damage. Catawba analyzed safe shutdown success paths as shown on the functional logic diagrams. The success logics incorporate OSP to the extent that if a component can spuriously operate/maloperate with the availability of OSP, then OSP was assumed to exist. This includes Normal shutdown and Alternate shutdown. Thus, Catawba aligns with the guidance of considering OSP during the analysis.

d. 3.1.1.11 Multiple units Duke Energy Response:

Catawba analyzed the effects on both units from a fire in each area. Since multiple unit effects of fires were analyzed, Catawba aligns with the guidance.

e. 3.1.2.5 Process Monitoring Duke Energy Response:

Process monitoring is provided for safe shutdown from the MCR or SSF. Therefore, Catawba aligns with the guidance.

f. 3.3.1.3 Isolation Devices Duke Energy Response:

Isolation devices were analyzed for the credited train in a given FA; therefore, Catawba aligns with the guidance.

g. 3.3.1.6 Auto Initiation Logic Duke Energy Response:

In addition to the guidance, the analysis was not limited to "...the fire-induced failure of automatic logic circuits.." The automatic interlock signals were analyzed and assumed to occur/not occur in the worst case situation unless specifically analyzed not to do so. This approach exceeds the guidance. Thus, Catawba aligns with the guidance.

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h. 3.3.1.7 Circuit Coordination Duke Energy Response:

Fault coordination impacts were analyzed for the credited trains/busses. Thus, Catawba aligns with the guidance in that a breaker coordination analysis was performed. Coordination issues found during this review were reviewed for risk insights. This resulted in several required modifications to remove the non-coordination issues. The following committed modifications will be done to resolve these coordination issues (see LAR Attachment S, Table S-2b, Items 02, 03, 04, 05, and 06 for details):

1. Remove the incoming breaker and connect wiring directly to the MCC bus for the following MCCs: 1EMXA, IEMXB, 1EMXC, 1EMXD, 1EMXI, IEMXJ, IEMXK, &

1EMXL.

2. Remove the incoming breaker and connect wiring directly to the MCC bus for the following MCCs: 2EMXA, 2EMXB, 2EMXC, 2EMXD, 2EMXI, 2EMXJ, 2EMXK, &

2EMXL.

3. Install fuses on the load side of 1(2)EDE and 1(2)EDF breakers.

4. Remove the fuse from the Motor Operator Heater circuit for 1CAVA0050A and 2CAVA0050A.

5. Reroute new cables for the normally energized circuits on IWLLS5900 and 2WLLS5900.

i. 3.5.1.3 Duration of Circuit Failures Duke Energy Response:

Catawba did not take credit to clear (i.e., the duration of the hot short was not limited) spurious operations in the deterministic analysis. Thus, Catawba aligns with this guidance.

j. 3.5.2.1 Circuit Failures Due to an Open Circuit Duke Energy Response:

Catawba analyzed open circuits per the NEI 00-01 guidance, including potential high voltage CT secondary damage as itemized. Thus, Catawba aligns with the guidance.

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k. 3.5.2.3 A/B Circuit Failures Due to a Hot Short Duke Energy Response:

Catawba evaluated fire-induced failure modes such as hot shorts, open circuits, and shorts to ground per the guidance in NEI 00-01. Shorts to ground included grounded and ungrounded circuits. In the case of a grounded circuit, a short to ground on any part of the circuit would present a concern for tripping the circuit isolation device thereby causing a loss of control power. In the case of an ungrounded circuit, postulating only a single short to ground on any part of the circuit may not result in tripping the circuit isolation device; another short to ground on the circuit or another circuit from the same source would need to exist to cause a loss of control power to the circuit. There were no limits on the number of shorts to ground that could be caused by the fire. Thus, Catawba aligns with the guidance.

I. 3.4.1.4 Manual Actions Duke Energy Response:

VFDR resolutions in the performance based FREs included recovery actions as potential mitigating actions to maintain a safe and stable condition for the operational effects of fire damage. Recovery actions are demonstrated to be feasible in accordance with the NRC endorsed requirements in FAQ 07-0030. These requirements included demonstrating action can be performed, required systems and indications are available, necessary communications are available, emergency lighting, as necessary, is available, any tools, equipment, or keys required are provided, written procedures are provided, sufficient staff is available, actions in the fire area, where necessary, can be performed, actions can be performed in the required time, and training/drills are provided. Thus, Catawba aligns with the guidance.

m. 3.4.1.7 Additional Equipment Duke Energy Response:

Catawba analyzed the effect of fires on all of the plant's safe shutdown equipment to determine the VFDRs affecting a success path. For the train with the least impact due to the fire, each VFDR was then dispositioned sufficiently to assure a safe shutdown with that train. Thus, Catawba aligns with the guidance.

Changes to the LAR Attachment B, Table B-2 associated with this RAI response will be provided with the 120-day RAI responses.

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SSA RAI 03 NPFA-805 Sections 2.4.2.4 and 4.2.4.1.5 requires that a fire area assessment be performed to determine the effects of fire or fire suppression activities on the ability to achieve the nuclear safety performance criteria of Section 1.5, and that fire suppression activities shall not prevent the ability to achieve the nuclear safety performance criteria. LAR Attachment C, Table C-1 describes the results of the fire suppression activities effect on Nuclear Safety Performance Criteria as "safe and stable conditions can mostly be achieved and maintained utilizing equipment and cables outside the area of fire suppression activity." This statement appears to be used generically in all fire areas.

a. Provide a more detailed explanation regarding the extent of "can mostly be achieved."

Duke Energy Response:

The statement in LAR Attachment C, Table C-1 for "Fire Suppression Activities Effect on Nuclear Safety Performance Criteria" in each fire area did not provide a proper response. The term "mostly" should not have been used. Plant walkdowns for drainage of fire suppression water did not identify any areas where flooding or runoff could affect achieving and maintaining nuclear safety. The statement in Table C-1 describing the fire suppression activities effect on Nuclear Safety Performance Criteria for every fire area will be revised to delete "mostly."

b. For fire areas where safe and stable is not achievable by utilizing only equipment and cables outside the area, provide a description of the equipment and functions that may be affected by fire suppression activities, and a description of how suppression effects are controlled or mitigated so that the nuclear safety performance criteria is achievable.

Duke Energy Response:

Based on the response to part a., part b. is not applicable.

Changes to the LAR Attachment C, Table C-1 associated with this RAI response will be provided with the 120-day RAI responses.

SSA RAI 04 Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, (ADAMS Accession No. ML092730314),

Section 2.3 allows the use of existing Appendix R deviations to demonstrate compliance with design requirements of NFPA-805 Chapter 4. Additionally, NFPA-805 Section 4.1 requires that fire protection system or features that are required to achieve the nuclear safety performance criteria shall have its design and qualification meet the applicable requirements of Chapter 3. LAR Attachment K, Licensing Action 07, identifies fire protection features that are credited in the licensing action review for Fire Area 01, such as the fixed suppression system in 14

the Reactor Building annulus area (elevations 561-ft, 604-ft and 664-ft) and line-type heat detectors on all six levels of the annulus area. However, these fire protection systems and features are not listed in LAR Attachment C, Tables C-1 and C-2, for Fire Area 01 as required fire protection features (suppression and detection).

a. Clarify whether the fire protection systems or features identified in Licensing Action 07 are required to meet the nuclear safety performance criteria of NFPA-805 Chapter 4.

b. If the subject fire protection systems or features are required, then confirm that they are designed and qualified to meet the applicable requirements of NFPA-805 Chapter 3.

Duke Energy Response:

Licensing Action 07 discusses two distinct and independent issues.

• The first issue deals with the commitment to install suppression and detection in the Unit I and Unit 2 Reactor Building Annulus (Fire Areas RB1 and RB2).

" The second issue addresses a commitment to ensure the Fire Brigade can respond with extra hose to the pipe tunnel area adjacent to the ND (RHR) and NS (Containment Spray) pump area at elevation 522+0 of the Auxiliary Building (Fire Area 01).

The suppression and detection in the licensing action discussion is pertinent to the Annulus Areas (RB1 and RB2) only. The pipe tunnel area is in Fire Area 01. Fire Area 01 does not have any required suppression and/or detection systems. The commitment regarding Fire Area 01 is that additional lengths of fire hose will be stored at the fire brigade locker. The suppression and detection systems in Annulus areas of Fire Area RBI and RB2 are identified as required in LAR Attachment C, Tables C-1 and C-2 by Licensing Action 07 as identified by an "E".

SSA RAI 05 LAR Attachment D describes the methods and results for non-power operations (NPO) transition. Provide the following additional information:

a. 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.

Duke Energy Response:

No additional actions beyond normal operating procedures for initial system alignments are credited for Non-Power Operations (NPO). Existing operating procedures, in some cases, require de-energization of key components in the shutdown cooling flowpath to preclude a loss of shutdown cooling event (i.e., the Residual Heat Removal suction valves 1/2ND VA0001B, 1/2ND VA0002A, 1/2ND 15

VA0036B, and 1/2ND VA0037A from the Reactor Coolant System) and these could be considered pre-emptive actions. However, this procedure requirement was not used to exclude a pinch point in the analysis. Those locations where a fire-induced spurious operation could impact a Key Safety Function (KSF) are identified, and where complete loss of a KSF occurs, the location is identified as a pinch point for application of additional fire risk management actions during designated Higher Risk Plant Operating States.

b. Identify those recovery actions relied upon in NPO and describe how recovery action feasibility is evaluated.

Duke Energy Response:

No recovery actions are required to support the Non-Power Operation analysis assumptions or are used to restore a Key Safety Function following a potential fire event during NPO conditions. Evaluation of feasibility is not required.

SSA RAI 06 LAR Attachment B, Table B-2, Section 3.2.1.2, "Fire Damage to Mechanical Components,"

states that heat sensitive piping materials, including tubing with brazed or soldered joints are not included in the assumption of no mechanical damage. It appears that the NSCA analysis does not address instrument air and instrument sensing lines. Describe how the analysis addressed "heat sensitive piping," including tubing with brazed or soldered joints.

Duke Energy Response:

Catawba's NSCA instruments do not use soldered or brazed connections. Process instrument lines, including instrument air lines, use stainless tubing and/or copper tubing. These use compression fittings in accordance with procedure SI101AI50901001 "Tube Fitting and Tubing Installation" and are assumed to remain intact. In other applications, any brazed and soldered lines are assumed damaged in the event of a fire (See - CNC-1435.00-00-0071 / AREVA Document 51-9183972-002, Catawba "NFPA 805 Transition - Deterministic Safe Shutdown Analysis", Section 5.4 - #8).

The Deterministic Safe Shutdown Analysis, Section 6.5.2.4,indicates that the Instrument Air System has not been analyzed for Safe Shutdown; therefore, Instrument Air is assumed to be lost. Safe Shutdown components that would require Instrument Air to perform their safe shutdown function were considered and identified as VFDRs due to the effects of a fire, if applicable, and dispositioned in the Fire Risk Evaluations (FREs).

Therefore, Catawba NSCA aligns with this guidance.

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Fire Modeling (FM) RAI 01 NFPA-805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. The NRC staff noted that the fire modeling analysis comprised the following:

- The Generic Fire Modeling Treatments (GFMTs) approach was used to determine the Zone of Influence (ZOI) for ignition sources and the time to Hot Gas Layer (HGL) conditions in all fire areas throughout CNS, Units 1 and 2.

- The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to assess the main control room (MCR) abandonment time calculations.

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

Regarding the acceptability of CFAST for the MCR abandonment time calculations:

b. In the MCR abandonment time analysis, the licensee assumed that the external doors of the MCR open at 15 minutes based on an estimated fire brigade arrival time. State whether 15-minute fire brigade response time for the MCR is used in the PRA and provide the technical justification for the time used.

Duke Energy Response:

The Fire PRA selects the most adverse fire scenario case (regardless of if/when doors are opened) from the MCR abandonment time analysis and incorporates those results into the Fire PRA.

The Catawba Fire Brigade average time in 2014 from initial fire alarm signal to fire brigade arrival at the scene for all plant locations is 16.35 minutes. While the fire brigade does not conduct drills specifically for fires in the MCR, the dress-out area is located in close proximity to the MCR. Additionally, the Unit Supervisors offices are adjacent to the MCR and all Unit Supervisors are trained as Fire Brigade Leaders.

Furthermore, there are several factors to consider that cause drill response times to be greater than actual response times during actual fire events:

• It is industry and Catawba practice that fire brigade drills incorporate a degree of training while performing an overall evaluation of the fire brigade, firefighting equipment performance, and plant administrative controls.

• Fire brigade drills may vary in types of response, speed of response, and use of equipment. A level of proficiency and safety is desired above simply speed of completion during drills.

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• Another factor affecting response times during drills is delays required for compliance with security, administrative controls, radiological controls, and other barriers. During actual fire events, access/egress routes and administrative procedures are expedited for the fire brigade, further decreasing response time.

• Industry experience also indicates fire brigade response will quicken based on the human behavioral stimuli provided by an actual event.

• Finally, the time required for drill controllers to verbally describe and the fire brigade to visualize the fire conditions adds considerable time to the drill process that would not be present during an actual event.

Specifically regarding the acceptability of the GFMTs approach:

g. The GFMTs approach describes the critical heat flux for a target that is immersed in a thermal plume. Explain how the modification to the critical heat flux was used in the ZOI and HGL timing determinations.

Duke Energy Response:

The modified critical heat flux is a means of accounting for both elevated temperatures and flame heat fluxes and was implemented using either a two or a three point (i.e., temperature) treatment in the fire PRA. When the modified heat flux is used to establish an ignition source Zone of Influence (ZOI) in an enclosure with an elevated temperature, the ZOI is larger than an ambient temperature based ZOI. Most plant areas use the two point treatment of the modified critical heat flux. The first point corresponds to temperature conditions between ambient and 80°C (1760 F) and represents the temperature interval in which the Zones of Influence (ZOls) such as those documented in the Generic Fire Modeling Treatments (GFMTs) are applicable.

The second point corresponds to temperature conditions greater than 80°C (176°F) and is conservatively characterized in the fire PRA as a full-room burnout. This applies to both targets located in the thermal plume region and to targets that are located outside the thermal plume region.

The Hot Gas Layer (HGL) review for ZOI impact utilizes a three point treatment for greater resolution on the risk characterization. The first point corresponds to temperature conditions between ambient and 80 0 C (1760F) and represents the temperature interval in which the ZOls for thermoset cable targets are applicable. The second point corresponds to temperature conditions greater than 80 0 C (176°F) but less than 2200 C (4280 F) and represents the region where the HGL can produce a heat flux up to 5.7 kW/m 2 (0.50 Btu/s 2). The ZOls for thermoplastic cable targets, which have a heat flux threshold of 5.7 kW/m 2 (0.50 Btu/s 2 ) are applicable in this temperature range when used to identify thermoset cable targets because the total heat flux at the ZOI boundary is 11.4 kW/m 2 (1.0 Btu/s 2), the generic threshold for thermoset cables per NUREG/CR-6850. The third point corresponds to temperature conditions greater than 220°C (428°F) and is conservatively characterized in the fire PRA as a full-room burnout. This applies to both targets located in the thermal plume region and to targets that are located outside the thermal plume region.

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Specifically regarding the acceptability of the PRA approach, methods, and data:

h. Identify whether any fire modeling tools and methods have been used in the development of the LAR that are not discussed in Attachment J of the LAR. One example would be a methodology used to convert damage times for targets in Appendix H of NUREG/CR-6850 to percent damage as a function of heat flux and time.

Duke Energy Response:

No other fire modeling tool or method was used outside the Generic Fire Modeling Treatments (GFMTs) or MCR abandonment calculation.

j. Regarding fires in the proximity of a corner or walls, explain how the GFMTs approach was applied. Explain how wall and corner affects the ZOI and HGL timing calculations were accounted for, or provide technical justification if these effects were not considered.

Duke Energy Response:

The following methodology was used for wall and corner effects in the Zone of Influence (ZOI) evaluation:

Regarding separation distances, Calculation DPC-1535.00-00-0024, Rev. 0, "Generic Fire Modeling Treatments (GFMT)", Section 3.3.7 (Guidance for Fuel Packages Positioned in a Corner and Wall) states:

1. If the fuel package is within 0.6 m (2 ft) of a wall, then double the heat release rate and assume that the fire is centered at the fuel package edge adjacent to the wall.

2. If the fuel package is within 0.6 m (2 ft) of a corner, then quadruple the heat release rate and assume that the fire is centered at the fuel package corner nearest the wall corner.

This GFMT is reflected in the Catawba Fire PRA (FPRA) Scenario Development Report, CNC-1535.00-00-0110, Rev. 1, Section 9.3 (Location Factor). This section states:

The location of an ignition source relative to a wall or a corner may impact the zone of influence [ZOI]. Fixed and transient ignition sources located in corners or against walls were identified during the scenario walkdown and annotated as such in Appendix A. Some of the ventilated cabinets in the Battery Rooms (Fire Areas 9

& 10) were confirmed to be located against an interior wall. However, the location of these cabinets against the wall had minimal impact on the heat release rate given the height of the internal walls relative to the top of the cabinet (the location of the ventilated opening) and the overall battery room ceiling. Therefore, the 19

impact of the location on the zone of influence for these fixed ignition sources (all of which were equipped with a top mounted deflector shield) has been addressed in the scope of assumed target damage. Similarly for transients, if the postulated transient location ... was along a wall or in a corner, the zone of influence was adjusted accordingly.

A review of the transient scenarios indicates that a wall effect was applied to a limited number of scenarios in the Battery Rooms and the Train B Switchgear Rooms. The wall effect resulted in doubling the HRR for the 142 kW transient fire; ZOI values for this HRR condition are provided in Appendix A to the GFMTs. While several transient scenarios were assumed to result in room burnout, it was not necessary to apply wall or corner effects to the HRR. The hypothetical transient fuel packages were placed where targets such as cable trays or risers would be impacted. If the target damage could be achieved by placement of the ignition source away from the wall or corner (i.e., an open location transient fuel package), then no further adjustments were applied.

Wall and corner effects were not applied to the HGL screening analysis. Because the overall heat input to the room is not increased by placement near a wall or corner, in order to address the initial change in rate, the room volumes (and ventilation parameters) would also be doubled and quadrupled, accordingly. The net impact on the HGL room burnout calculation is, therefore, considered negligible.

k. Regarding high energy arcing fault (HEAF) generated fires, describe the criteria used to decide whether a cable tray in the vicinity of an electrical cabinet will ignite following a HEAF event in the cabinet. Explain how the ignited area was determined and subsequent fire propagation was calculated. Describe the effect of cable tray covers and fire-resistant wraps on HEAF induced cable tray ignition and subsequent fire propagation.

Duke Energy Response:

As stated in the fire scenario report, in addition to assuming loss of power at the switchgear/load center, the zone of influence for an HEAF should include:

• The first adjoining cubicle (if adjoining cubicle is empty, damage to first non-empty adjoining cubicle is to be assumed).

" Targets, including cable trays, within 5' vertical distance of the top of the panel.

• Targets within 3' horizontal distance from the front and back of the panel, below the top of the panel and 1' horizontal distance above the top of the panel.

Any resulting heat release rate (HRR) following the HEAF event is subsumed by the Bin 15 HRR which is used for the determination of the target damage set. No credit is 20

taken for "time to peak HRR" for the ZOI/target damage associated with the modeled HEAF events.

Fire propagation was treated such that the target trays within the ZOI are assumed to ignite with propagation up to the ceiling. Exceptions were limited to configurations where there is significant distance between trays such that the fire would not propagate to the next tray. Flame spread is further discussed in response to FM RAI 01.1, which is planned for submittal with the 90-day responses.

Tray covers and fire resistant wraps were not credited for HEAF fires.

FM RAI 02 American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications," Part 4, requires damage thresholds be established to support the FPRA. The standard further states that thermal impact(s) must be considered in determining the potential for thermal damage of systems, structures, and components (SSCs) and appropriate temperature and critical heat flux criteria must be used in the analysis.

Provide the following information:

c. Describe how cable tray covers, conduits and wraps affect the damage thresholds that were used in the fire modeling analyses.

Duke Energy Response:

No Fire PRA credit was taken for cable tray covers, conduits, or wraps.

d. 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.

Duke Energy Response:

In accordance with Appendix H.2 of NUREG/CR-6850, for major components such as motors, valves, etc., the fire vulnerability was assumed to be limited by the vulnerability of the power, control, and/or instrument cables supporting the component. As stated in the scenario development calculation, "In some fire scenarios, the target set may include another cabinet in which case application of the cable damage threshold would tend to be conservative since no credit would be taken for the protective nature of the enclosure." In other words, electrical cabinets are subject to the same damage thresholds as the cables in the analysis. No other non-21

0-cable components were assigned a damage threshold different from that which was used for cables.

e. Explain how exposed temperature-sensitive equipment was treated, and provide a technical justification for the damage criteria that were used.

Duke Energy Response:

The sensitive electronics treatment at Catawba is consistent with many aspects of the FPRA FAQ 13-0004. For example, the damage criteria used for temperature-sensitive electronic equipment inside of electrical cabinets was the same as that for thermoset cables. However, Catawba has yet to officially incorporate FPRA FAQ 13-0004 since it was not approved when the FPRA was developed. The current sensitive electronics treatment in the Catawba FPRA does not fully address the caveats in FPRA FAQ 13-0004 regarding sensitive electronics mounted on the surface of cabinets and the presence of louvers or vents. These caveats will be addressed in further detail in the response to PRA RAI 14, which will be included in the 120-day responses.

FM RAI 03 NFPA-805, Section 2.7.3.2, states that each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models.

LAR Section 4.5.1.2, states that fire modeling was performed as part of the FPRA development (NFPA-805 Section 4.2.4.2). Reference is made to LAR Attachment J for a discussion of the verification and validation (V&V) of the fire models that were used. Furthermore, LAR Section 4.7.3 states that "calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805."

a. Regarding the V&V of fire models, for any fire modeling tool or method that was used in the development of the LAR or that is identified in the responses to the above fire modeling RAIs, provide the V&V basis if it is not already explicitly provided in the LAR (for example in LAR Attachment J).

Duke Energy Response:

No other fire modeling tool or method was used outside the Generic Fire Modeling Treatments (GFMTs) or MCR abandonment calculation which are already explicitly provided in the LAR.

FM RAI 05 NFPA-805, Section 2.7.3.4, "Qualification of Users," states: "Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be 22

competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations."

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

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

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

Post-transition, cognizant personnel who use and apply engineering analysis and numerical models shall be competent in this field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations. Duke Energy will develop and maintain qualification requirements for individuals assigned various tasks. Individuals will be qualified to appropriate job performance requirements per ACAD 98-004. Engineering training guidelines will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA-805 Section 2.7.3.4 to perform assigned work.

Regarding qualifications of users of engineering analyses and numerical models:

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

Duke Energy Response:

Duke Energy considers the following to be appropriate qualifications for Fire Protection Engineers and contractors to perform and review fire modeling analyses using fire modeling tools and methods:

" The INPO accredited training program, or recognized equivalent, will be used to ensure that individuals are qualified to perform the applicable task.

• The training program will include activities such as complete reading assignments of task instructions (i.e., calculation procedures) for the relevant work that will be 23

performed. This requirement also includes completing independent studies for relevant industry methodology and/or guidance documents such as NUREG/CR-6850, NUREG-1934, NUREG-1805, and other applicable fire modeling user's guide documents, etc.

Education on the subject of combustion, fire dynamics, and/or fire modeling.

Examples of education activities meeting this requirement include:

o Academic training in fire analysis (e.g., fire modeling, fire dynamics, etc.)

o Demonstration of comprehension and proficiency in fire modeling For the specific case of Duke Energy contractors, the contractor's quality assurance process ensures that the personnel performing the fire modeling are qualified and trained. The contractor's qualifications are maintained by the contracting company quality assurance manager who ensures that the education credentials, appropriate quality assurance training, and reading assignments are completed before the tasks are performed.

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

Duke Energy Response:

Fire modeling calculations are required to be performed by a Fire Protection Engineer who meets the qualification requirements of Section 2.7.3.4 of NFPA 805. The qualification process is based on the following programs, which provides the minimum training necessary to perform calculations and analyses:

• Fire Protection Plant Change Impact Review

• Fire Protection Engineer

" Basic Fire Modeling The requirement in NFPA 805 listed above will continue to be met and adhered to through Duke Energy procedures and project management of contractor support staff.

For personnel performing fire modeling or fire PRA development and evaluation, Duke Energy maintains qualifications. The qualifications are developed in accordance with Duke Energy's accredited training program. The qualifications identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805, Section 2.7.3.4 to perform assigned work.

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c. Describe who performed the walk-downs of the MCR and other fire areas in the plant.

Describe whether these were the same people who performed the fire modeling analysis.

Duke Energy Response:

Hughes personnel performed the walkdowns and fire modeling analysis for the MCR.

For the other fire areas, ERIN personnel performed the initial walkdowns and then applied the Hughes developed Generic Fire Modeling Treatments (GFMTs). Duke Energy personnel conducted subsequent walkdowns.

d. Explain the communication process between the CNS fire modeling analysts, PRA personnel, consulting engineers and CNS 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.

Duke Energy Response:

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

Currently, knowledge transfer between the consulting engineers and the station and fleet personnel has been in progress through the development of fire risk insights, the Request for Additional Information (RAI) process, updates to the analysis based on plant modifications/changes, and use of the fire PRA analysis to support plant activities such as other regulatory activities where the fire PRA is required to support.

• The fire modeling/fire PRA qualification relevant to fire modeling tasks will be developed and maintained jointly by the Fire Protection and PRA groups.

• Fire modeling personnel will also prepare and maintain the calculations supporting the fire PRA analysis.

It should also be noted that portions of the GFMTs calculation and the MCR abandonment calculation methodology are used at the other Duke Energy nuclear plants.

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I Radiation Release (RR) RAI 07 LAR Section 4.4.2 states that the fire protection program will be compliant with the requirements of NFPA-805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the Implementation Items identified in LAR Attachment E. Should this be LAR Attachment S instead of LAR Attachment E?

Duke Energy Response:

Upon further review, LAR section 4.4.2 was determined to contain an editorial typographical error. The last paragraph of this LAR section will be revised to read, "The radioactive release review determined the fire protection program will be compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the Implementation Items identified in Attachment S."

The corrected wording will appear in the revised compiled LAR planned for submittal with the 120-day RAI responses.

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

Clarify the following dispositions to Internal Events F&Os and SR assessments identified in LAR Attachment U that have the potential to impact the FPRA results and do not appear to be fully resolved:

d) The disposition of F&Os TH-01 and TH-02 concludes that difference in the time to core damage is not significant when using either 2000 F or 4000 F "because the exothermic nature of the zircaloy-water reaction rapidly increases the fuel temperature." Relatively small changes in "the time available for human recoveries or other non-recovery events such as loss of offsite power recoveries" could change the likelihood of some events.

Provide further clarification on the difference in the time available and what is meant by not significant and negligible impact. The response should address not significant and negligible in the context of both the RG 1.174 risk guidelines for transition and the post-transition change evaluation criteria, which is two orders-of-magnitude less than the RG 1.174 risk guidelines.

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Duke Energy Response:

Based on a review of the T/H runs performed in support of success criteria and HRA timing analysis for verification, there is no impact on the results if core damage is defined as 2000°F as the F&Os describe. There was no change in the success criteria. Also, the new plant-specific HRA timing supported the HEPs used in the FPRA with respect to the core damage definition of 2000 0 F.

PRA RAI 04 Section 2.4.3.3 of NFPA-805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA-805 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC.

RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff review of the information in the LAR has identified additional information that is required to fully characterize the risk estimates.

LAR Section 4.5.2.2 provides a high-level description of how the impact of transition to NFPA-805 impacts DID and safety margin was reviewed, including using the criteria from Section 5.3.5 of NEI 04-02 and from RG 1.205. However, no explanation is provided of how specifically the criteria in these documents were utilized and/or applied in these assessments.

a) Provide further explanation of the method(s) or criteria used to determine when a substantial imbalance between DID echelons existed in the Fire Risk Evaluations (FREs), and identify the types of plant improvements made in response to this assessment.

Duke Energy Response:

Defense in Depth:

The methodology for assessing DID in the FREs is described in Section 5.4.2 of Catawba Calculation CNC-1435.00-00-0067, "NFPA 805 Transition, Fire Risk Evaluations (FREs)". See Attachment 1 to this RAI response.

b) Also, provide further discussion of the approach in applying the NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Revision 2, (ADAMS Accession No. ML081130188) criteria for assessing safety margin in the FREs.

Duke Energy Response:

Safety Margin Considerations:

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Based on NEI 04-02, the requirements related to safety margins for the Fire Risk Evaluation (FRE) is described for each of the specific analysis types used in support of the fire risk assessment.

• Fire Modeling

• Plant System Performance

• FPRA Logic Model

• Success Path Verification

1. Fire Modeling For fire modeling used in support of the FRE (i.e., as part of the Fire PRA), the results were documented as part of the qualitative safety margin review. The following statement regarding fire modeling was documented and confirmed for each fire area fire risk evaluation.

Fire modeling performed in support of the transitionhas been performed within the Fire PRA utilizing codes and standardsdeveloped by industryand NRC staff to provide realisticyet conservativeresults. Specifically, the heat release rates utilized in the transitionanalysis are based upon NUREG/CR-6850, Appendix E, Severity Factors. These heat releaserates are conservative and representvalues used to screen out fixed ignition sources that do not pose a threat to the targets within specific fire compartments and to assign severity factors to unscreened fixed ignition sources. The combined analysis approachis used during transition;therefore, maximum expected fire scenario/limitingfire scenario have not been analyzed separately. The bases for the applicationof these fire modeling codes and standardswere not alteredin supportof this FRE.

2. Plant System Performance This review documented that the Safety Margin inherent in the analyses for the plant design basis events was preserved in the analysis for the fire event and satisfied the requirements of this section. The following statement regarding Plant System Performance was documented and confirmed for each fire area fire risk evaluation.

Performanceparameters were originally establishedto support nuclear performance criteria containedin the plant specific accident analyses. These analyses establishedcomponent and system performance criterianecessary to establish safe and stable plant operationin the event of a fire. These performanceparameterswere not modified as a result of this FRE.

3. FPRA Logic Model This aspect of Safety Margin was implicitly addressed in the FRE. This treatment was deemed sufficient because the FRE risk quantification method does not impact Safety Margin inherent in fire PRA. The fire PRA is characterized as 28

having safety margin based on the following characteristics which satisfy the guidelines established in NEI 04-02 Rev 2:

" Changes to the internal Events PRA in support of the fire PRA logic model development were performed using the same methods and criteria for the internal events PRA model development. In July of 2010, the at-power fire PRA was subjected to a Peer Review using the applicable requirements of the Addenda to American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) standard, ASME/ANS RA-S-2008, "Standard for Level I / Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant", February 2009. The resolution of the resulting Findings and Suggestions is documented in CNC-1535.00 0113, Revision 2.

• To ensure that safety margin, inherent in the PRA model is preserved, treatments were applied via methods such as the following:

o Application of industry recommended fire related event frequencies and probabilities.

o Increased Human Reliability Analysis failure probabilities with conservative multipliers to account for fire initiating event effects (note that further discussion of this approach is addressed in PRA RAIl b).

o Application of the guidance in NUREG/CR-6850.

4. Success Path Verification (corresponds to Miscellaneous in NEI 04-02)

The "Miscellaneous" category addresses any other analyses not addressed in the three elements discussed above. For Catawba, the applicable "miscellaneous" analysis is Success Path Verification, which is the confirmation that changes in CDF and LERF are below the acceptance criteria in order to establish that a success path effectively remains available. NEI 04-02, 5.3.5.3 guidance requires that codes and standards or their alternatives accepted for use by the NRC are met, and that safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.

PRA RAI 10 Section 2.4.3.3 of NFPA-805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA-805 further states that the change in public health risk arising from transition from the current fire protection program to an NFPA-805 based program, and all future plant changes to the program, shall be acceptable to the NRC.

RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes. The NRC staff review of the information in the LAR has identified additional information that is required to fully characterize the risk estimates.

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LAR Section V.2.6 indicates that only Bins 4 and 15 are applicable to the fire ignition frequency sensitivity analysis. For each of the other Bins having an alpha of less than or equal to 1, provide the basis for concluding that each does not impact the VFDR delta risk results.

Duke Energy Response:

FAQ 08-0048 communicated the NRC's acceptance of the updated frequencies provided a sensitivity study was performed to address the differences in risk and delta-risk results. This provision was limited to the bins with an alpha of less than or equal to 1.

Fire events Bins 1, 4, 9, 11, 13, 15, 22, and 31 have alphas less than or equal to 1. FAQ 08-0048 noted that a sensitivity analysis need not be performed for Bin 9. Bins 4 and 15 were identified as the only bins of concern for potential VFDR delta risk impact. Bin 13 was not used. The non-zero delta risk results based on the NFPA 805 VFDRs were reviewed to ensure the conclusions from the fire risk evaluations are not challenged. Bin I fire scenarios are limited to loss of the battery itself and were not associated with any VFDRs. Similarly, Bin 22 fire scenarios were not associated with any VFDRs. Bin 11 was grouped with Bin 24 in a limited number of transient hot work fire scenarios involving VFDRs; the VFDR impact on delta CDF and delta LERF was less than 5E-10/yr and 2E-1 1/yr, respectively. Bin 31 was grouped with Bins 36 and 37 in a limited number of transient fire scenarios, none of which was associated with any VFDRs. The Bin 4 frequency decrease was limited to Fire Area 21 (Control Room) scenarios where the delta risk considering the frequency difference between NUREG/CR-6850 and FAQ 08-0048 remained below the applicable acceptance thresholds. Similarly, the Bin 15 frequency difference resulted in the change in risk remaining below the applicable acceptance thresholds and did not alter the FRE conclusions for any fire area.

PRA RAI 16 Section 2.4.3.3 of NFPA-805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

The MCB is described as having a horseshoe arrangement that is fully enclosed and is effectively a sub-enclosure. The analysis of MCB fires appears to treat the front and back panels of the horseshoe as an integral part of the MCB.

a) FAQ 14-0008 provides guidance on how MCB fires should be treated for MCBs that are sub-enclosures. Describe how your MCB configuration and MCB fire scenario analysis is consistent with the FAQ.

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Duke Energy Response:

FPRA FAQ 14-0008 was developed after submittal of the Catawba NFPA-805 LAR.

The MCB fire scenario development for Catawba is consistent with the FAQ. The rear side of the MCB is classified as an integral part of the MCB because the rear and front sides are connected together as a single enclosure. There are no unique panel segments on the rear side of the MCB; the rear side of the MCB is actually the rear face of the corresponding panel segment on the front side of the MCB. Consequently, the failures on the rear face are assumed coincident with the failures on the front face assuming a distance between failures of 0' despite an actual separation distance of 7'-6" between the faces. There is a continuous overhead ceiling connecting the front and back sides. Therefore, the methodology in FPRA FAQ 14-0008 is considered applicable. The Catawba MCB fire scenario analysis is consistent with the FAQ's guidance for crediting partitions. For the basis, see the discussion in the "MCB Fire Ignition Frequency" discussion in Part B of this RAI.

b) Describe how MCB fire scenarios are postulated and evaluated, including how the fire ignition frequency is determined for each scenario, how NUREG/CR-6850 Appendix L is applied to individual scenarios, how partitions between panels/cabinets are treated if credited, and how propagation between the front and back sides of the MCB is evaluated including identification of and evaluation of damage to target sets.

Duke Energy Response:

The key elements of the MCB fire scenario development are described below:

MCB Fire Ignition Frequency:

The entire frequency for the MCB for a single unit was applied to each of the MCB fire scenarios. Application of the entire ignition frequency captures the potential addition of risk from the rear of the MCB panel in accordance with Alternative 2 of FPRA FAQ 14-0008.

MCB Severity Factor and Non-Suppression Factors:

Severity factor/non-suppression probability was based on NUREG/CR-6850 Appendix L, Figure L-1, "Likelihood of Target Damage Calculated as the Severity Factor Times the Probability of Non-suppression for MCB Fires." Values selected were based on the distribution for qualified cables. Appendix L probabilities were credited based on distances between the instrumentation and controls for major functions (target sets) on the control board. The front to back propagation is addressed in part a) of this response.

MCB Fire CCDP:

The approach involved calculating the cumulative CCDP as the fire spreads from one end of an MCB segment to the other. In some cases, the entire set of targets for an MCB cabinet was assumed for a given scenario. In other cases, the 31

distance between key functions dictated that multiple scenarios for a given MCB cabinet be postulated.

PRA RAI 18 Section 2.4.3.3 of NFPA-805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

A severity factory of 0.20 is applied to some MCA scenarios "to account for the probability that only 1 in 5 fires are expected to challenge the zone boundary." This is an industry average-type factor that does not account for the design-specific considerations and potential for HGL formation at CNS. In addition, a barrier failure probability of 7.4E-03 is also applied to all MCA scenarios, which only accounts for the barrier having the highest probability of failure (e.g., non-rated barrier, door, damper, or wall)."

a) Is the 0.20 factor only applied when there is a rated fire barrier? If not provide further justification of the use of any factor.

Duke Energy Response:

The 0.2 factor relates to the frequency of a potential multi-compartment analysis (MCA) scenario and is not related to the fire barrier rating. As described in Section 11.5.4 of NUREG/CR-6850, a screening process was used to eliminate certain compartments from MCA consideration. Initially, the compartments where a damaging hot gas layer (HGL) cannot be generated were screened from MCA consideration. For the remaining compartments where a damaging HGL is possible, the additional screening criterion based on the product of ignition frequency, non-suppression probability, and barrier failure probability was applied. The 0.2 factor was multiplied by the entire ignition frequency for the applicable compartments to account for the fact that many ignition sources cannot generate a hot gas layer.

In response to this RAI, it was determined that for many of the compartments, the actual frequency for the scenario(s) capable of generating the HRR input into the HGL was less than the assumed 0.2 "initial term". In some cases, the actual ignition source frequency contribution to HGL scenarios was applied in conjunction with an initial term of 1.0. In other cases, for conservatism, the 0.2 initial term was simply replaced with 1.0 and the entire compartment frequency was input to the MCA screening calculation.

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PRA RAI 21 Section 2.4.3.3 of NFPA-805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a Fire PRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the NRC staff for adopting a fire protection program consistent with NFPA-805. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.

The fire ignition frequency for cable fires caused by welding and cutting (CFWC) is apportioned based on the number of raceways in each compartment in lieu of cable loading per NUREG/CR-6850. Provide a justification of your method that includes a discussion of conservatisms and non-conservatisms relative to the accepted methods. Assess the impact of using the accepted method instead of the proposed method on several high and intermediate risk areas.

Alternatively, replace the current approach with an acceptable approach in the integrated analysis performed in response to PRA RAI 3.

Duke Energy Response:

The intent of the cable weighting factor is to apportion the fires due to cutting and welding among the various areas of the plant. NUREG/CR-6850 and Fire PRA FAQ 13-0005 suggest a weighting factor based on cable load or the amount of cable material in a specific area. This approach is appropriate since the areas with more cables have more opportunities for a fire to start due to cutting and welding activities.

Duke Energy does not have an analysis that determines the cable load (mass or volumetric) for Catawba. In lieu of actual cable loading data, the weighting factors calculated for the "Cable Fires Caused by Welding and Cutting" bins were based on DATATRAK raceway data. Rather than calculating the ratio of quantity of cables in each compartment over the total quantity in the generic location, the ratio was based on cable trays. This is not appreciably different than the approach which would have been taken had cable loading data been available given that typical cable loading calculations in support of Appendix R are based on the number of trays and with an assumed cable fill percentage. In other words, fire compartments with the highest number of cable trays are also likely to have the highest number of cables (and the highest combustible loading due to cable mass).

Both approaches, cable loading and cable trays, apportion fires based on the opportunity for cables to be exposed to cutting and welding activities. Each approach represents a simplified method that uses a surrogate parameter (cable volume vs. cable tray count) in an attempt to divide a compartment fire frequency among existing cable trays/raceways, without any regard to physical barriers that may shield cable trays (including other cable trays), or other potentially realistic considerations that may be used to evaluate these fires. Both approaches inherently assume that cables are routed in roughly the same manner and through compartments with approximately the same exposure to cutting and 33

a welding (relative to the hot work influence factor). Thus, either method is believed to be acceptable for the Fire PRA, and both include inherent conservatisms and non-conservatisms:

Conservatisms:

o The cable tray approach is conservative for areas with cable trays with low cable loading. However, the opportunity for the tray to be exposed to cutting and welding is the same, largely irrespective of cable tray loading.

o The cable tray approach is also conservative for areas that have short cable trays.

o The cable loading approach is conservative for areas where the cables are concentrated in fewer cable trays in one location versus distributed more evenly in cable trays located throughout the area.

• Non-Conservatisms o The cable tray approach is non-conservative for areas with a low number of long cable trays.

o The cable loading approach is non-conservative for areas with a small quantity of cables distributed throughout the area. There may be more opportunities to impact cables due to exposure to more cutting and welding activities.

The generic locations of concern are the Auxiliary Building (CAR), Plant-Wide (PW), and Turbine Building (TB) for Cable Fires Caused by Welding and Cutting Bins 5, 11, and 31, respectively. Similar to Containment (COP), the TB generic location is comprised of two fire compartments, one for Unit I and one for Unit 2. Therefore, the weighting factor is somewhat immaterial to fire compartments in these locations since the frequency should be split equally among Units 1 and 2. Similarly, the cable weighting factor is considered appropriate for the fire compartments in the PW generic location given that the Service Building, which has the largest cable weighting factor among the PW fire compartments, is more significant to the fire risk results and has a higher number of cable trays (and cables) than the other PW fire compartments.

The primary generic location of concern relative to influence of the cable weighting factor on FPRA results is CAR which is comprised of 50 fire compartments. The fire compartments within CAR with the highest cable weighting factor are the Cable Rooms (FA 17 and FA 16), the Battery Rooms (FA 10 and FA 9), the Control Room (FA 21), and common fire areas at Elevation 560' (FA 11) and Elevation 577' (FA 18). These fire compartments are considered high risk fire areas and are also among the largest fire compartments within the CAR generic location. Larger fire areas would not pose a concern relative to an inordinate number of cable trays with short cable lengths that could skew the frequency toward rooms with a higher percentage of cable trays but a lower overall percentage of cable mass. The Control Room has no overhead cable trays so its ranking may appear to be overstated from a hot work-induced cable fire perspective; however, since the entire Bin 5 frequency for the Control Room fire area was included in the non-MCB Control Room abandonment scenario, the allocation of the Bin 5 frequency to the Control Room fire area is believed to be acceptable. Other high 34

risk fire areas such as the Switchgear Areas are much smaller and would not warrant a higher weighting factor than the aforementioned fire areas within CAR. Additionally, the fire compartments within CAR with the lowest cable weighting factor are the Fuel Storage Areas (FA 24 and 23), the Fuel Storage Area HVAC Rooms (FA 38 and FA 47), the Electrical Penetration Rooms at El. 594' (FA 20 and 19), and the Auxiliary Building Roof.

These fire areas are among the very lowest contributors to fire risk and contributed no quantifiable VFDR delta risk.

Therefore, the determination of the weighting factors based on raceway or cable tray counts for apportioning the frequency for cable fires caused by welding and cutting is appropriate given the additional considerations of fire compartment risk contribution and fire compartment size.

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V Attachment 1 36

'4 PRA RAI 4a - Attachment 1 Calculation No. CNC-1435.00-00-0067 Revision No: 1 Applicable Units: Catawba Units 1 &2 Page 12 of 43 Table 5 RG 1.174 Acceptance Criteria Region ACDF/rx-yr ALERF/rx-yr Status Comments/Conditions I 1.OE-05 2-1.OE-06 Unacceptable Proposed changes in this region are not acceptable, regardless of baseline CDF and LERF.

11 < 1.OE-05 and < 1.OE-06 and Acceptable Proposed changes in this region are

->1.OE-06 >-1.OE-07 w/ conditions acceptable provided the cumulative total CDF from all CDF initiators is less than 1.OE-04/rx-yr and from all LERF initiators is <1E-5/rx-yr.

Cumulative effect of changes must be tracked and included in subsequent changes.

III < 1.0E-06 < 1.OE-07 Acceptable Proposed changes in this region are w/ conditions acceptable provided the cumulative total CDF from all initiators is less than 1.OE-03/rx-yr and from all LERF initiators is <1E-4/rx-yr. Cumulative effect of changes must be tracked and included in subsequent changes.

In order to ensure the criteria above were met cumulatively for the plant, the acceptance criteria for an individual Fire Area was initially established as delta CDF less than 1E-07/rx-yr and delta LERF less than IE-08/rx-yr. These acceptance criteria are intended to support a total plant delta CDF and delta LERF within the acceptance guidelines of RG 1.174, delta CDF less than 1E-06/rx-yr and delta LERF less than 1E-07/rx-yr, for plant total CDF/LERF (conservatively including internal events contribution to plant risk) of 1E-4/1E-5/rx-yr, respectively.

Also, to ensure that the acceptance criteria were not solely based on low ignition frequency, the CCDP values for each of the post-transition baseline case scenarios were reviewed.

5.4.2 Defense-in-Depth (DID) 5.4.2.1 Guidance A review of the impact of the change on DID was performed, using the guidance below from NEI 04-02. NFPA 805 defines defense-in-depth as:

" Preventing fires from starting

" Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage

" Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

In general, the DID requirement was satisfied if the proposed change does not result in a substantial imbalance among these elements (or echelons). The review of DID was qualitative and addressed each of the elements with respect to the proposed change. Fire protection features

'P"6 Calculation No. CNC-1435.00-00-0067 Revision No: 1 Applicable Units: Catawba Units 1 & 2 Page 13 of 43 and systems relied upon to ensure DID were identified in the assessment (e.g., detection, suppression system).

Consistency with the DID philosophy is maintained if the following acceptance guidelines, or their equivalent, are met:

" A reasonable balance is preserved among 10 CFR 50.48(c) DID elements.

" Over-reliance and increased length of time or risk in performing programmatic activities to compensate for weaknesses in plant design is avoided.

" Pre-fire nuclear safety system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges to the system and uncertainties (e.g., no risk outliers). (This should not be construed to mean that more than one safe shutdown/NSCA train must be maintained free of fire damage.)

" Independence of DID elements is not degraded.

" Defenses against human errors are preserved.

" The intent of the General Design Criteria in Appendix A to 10 CFR Part 50 is maintained.

5.4.2.2 DID Process Each Fire Area was evaluated for the adequacy of DID. In accordance with NFPA 805 Section 2.4.4, Plant Change Evaluation, "...The evaluation process shall consist of an integrated assessment of the acceptability of risk, DID, and safety margins." NFPA 805 Section 4.2.4.2 refers to the acceptance criteria in this section. Therefore fire protection systems and features required to demonstrate an adequate balance of DID are required by NFPA 805 Chapter 4.

Considerations for determining what constitutes DID criteria are broken down into three echelons, which are provided in Table 5-2. These echelons cover a wide range of administrative, active and passive systems and/or features that are qualitatively reviewed against the particular risk characteristics of a fire area where their incorporation would further provide necessary mitigation effects.

Balance of these three DID echelons is inherent to the process. There are fundamental fire protection features in every fire protection program that will be required to meet NFPA 805 Chapter 3. These features include hot work and combustible controls, pre-fire plans, rated barriers, etc. The FRE utilizes this approach to ensure these fundamental features are always required to meet NFPA 805 Chapter 4.

Calculation No. CNC-1435.00-00-0067 Revision No: 1 Applicable Units: Catawba Units 1 & 2 Page 14 of 43 Table 5 Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 1: Prevent fires from starting

" Combustible Control Combustible and hot work controls are fundamental elements

" Hot Work Control of DID and as such are always in place. The issue to be considered during the FREs is whether this element needs to be strengthened to offset a weakness in another echelon thereby providing a reasonable balance. Considerations include:

" Creating a new Transient Free Areas

" Modifying an existing Transient Free Area The fire scenarios involved in the FRE quantitative calculation should be reviewed to determine if additional controls should be added.

Review the remaining elements of DID to ensure an over-reliance is not placed on programmatic activities to compensate for weaknesses on plant design.

Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage

• Detection system Automatic suppression and detection may or may not exist in

- Automatic fire suppression the Fire Area of concern. The issue to be considered during

• Portable fire extinguishers provided the FRE is whether installed suppression and or detection is for the area required for DID or whether suppression/detection needs to be

• Hose stations and hydrants provided strengthened to offset a weakness in another echelon thereby for the area providing a reasonable balance. Considerations include:

• Fire Pre-Fire Plan " If a Fire Area contains both suppression and detection and fire fighting activities would be challenging, both detection and suppression may be required

" If a Fire Area contains both suppression and detection and fire fighting activities would not be challenging, require detection and manual fire fighting (consider enhancing the pre-plans)

" If a Fire Area contains detection and a recovery action is required, the detection system may be required.

" If a Fire Area contains neither suppression nor detection and a recovery action is required, consider adding detection or suppression.

The fire scenarios involved in the FRE quantitative calculation should be reviewed to determine the types of fires and reliance on suppression should be evaluated in the area to best determine options for this element of DID.

Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed

• Walls, floors ceilings and structural If fires occur and they are not rapidly detected and promptly elements are rated or have been extinguished, the third echelon of DID would be relied upon.

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Calculation No. CNC-1435.00-00-0067 Revision No: 1 Applicable Units: Catawba Units 1 & 2 Page 15 of 43 Table 5 Considerations for Defense-in-Depth Determination Method of Providing DID Considerations evaluated as adequate for the The issue to be considered during the FRE is whether existing hazard. separation is adequate or whether additional measures (e.g.,

" Penetrations in the Fire Area barrier supplemental barriers, fire rated cable, or recovery actions) are rated or have been evaluated as are required offset a weakness in another echelon thereby adequate for the hazard, providing a reasonable balance. Considerations include:

" Supplemental barriers (e.g., ERFBS, 0 If the VFDR is never affected in the same fire scenario, internal cable tray covers, combustible liquid Fire Area separation may be adequate and no additional dikes/drains, etc.) reliance on recovery actions is necessary.

" Fire rated cable a If the VFDR is affected in the same fire scenario, internal Fire

" Reactor coolant pump oil collection Area separation may not be adequate and reliance on a system (as applicable) recovery action may be necessary.

" Guidance provided to operations 0 If the consequence associated with the VFDRs is high personnel detailing the required regardless of whether it is inthe same scenario, a recovery success path(s) including recovery action and / or reliance on supplemental barriers should be actions to achieve nuclear safety considered.

performance criteria. 0 There are known modeling differences between a Fire PRA and nuclear safety capability assessment due to different success criteria, end states, etc. Although a VFDR may be associated with a function that is not considered a significant contribution to CDF, the VFDR may be considered important enough to the NSCA to retain as a recovery action.

The fire scenarios involved in the FRE quantitative calculation should be reviewed to determine the fires evaluated and the consequence in the area to best determine options for this element of DID.

5.4.2.3 Defense-in-Depth - Recovery Action Considerations Reliance on Recovery Actions in lieu of protection is considered part of the third echelon of DID. Per NFPA 805, recovery actions are defined as: "Activities to achieve the nuclear safety performance criteria that take place outside of the main control room or outside of the primary control(s) station for the equipment being operated, including the replacement or modification of components."

If the VFDR is characterized as a 'Separation Issue', and the change in risk (delta CDF and delta LERF) is acceptable, a recovery action can be considered as a means to provide an adequate level of DID. Guidance on the need to establish/rely on a recovery action is provided in Table 5-

2. The 'additional risk presented by the use of the recovery action' would be characterized as the calculated change in risk of the 'Separation Issue,' although it is not explicitly calculated using the Fire PRA.

5.4.3 Safety Margin Assessment A review of the impact of the change on safety margin was performed. The guidelines for making that assessment aresummarized below.

0 Codes and standards or their alternatives accepted for use by the NRC are met, and If 0