NL-12-2566, Supplemental Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

Supplemental Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

ML12359A050
Person / Time
Site:	Farley
Issue date:	12/20/2012
From:	Ajluni M Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-12-2566, TAC ME9741, TAC ME9742
Download: ML12359A050 (67)

Text

Sensitive Information. Withhold Mark J. Ajiluni, P.E. Southern Nuclear from public disclosure per 10 CFR Nuclear Licensing Director Operating Company, Inc. 2.390. Decontrolled upon removal 40 Inverness Center Parkway of Enclosure 6.

Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7673 Fax 205.992.7885SOUTHERN COMPANY December 20, 2012 Docket Nos.: 50-348 NL-1 2-2566 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Supplemental Information Regarding License Amendment Request For Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)

Ladies and Gentlemen:

By letter dated September 25, 2012, the Southern Nuclear Operating Company (SNC) submitted a license amendment request (Ref. TAC NOS. ME9741 AND ME9742) for Joseph M. Farley Units 1 and 2. The proposed amendment requests the review and approval for adoption of a new fire protection licensing basis which complies with the requirements in Sections 50.48(a) and 50.48(c) to Title 10 to the Code of Federal Regulations CFR (10 CFR), and the guidance in Regulatory Guide (RG)1.205, Revision 1, Risk-Informed, Performance-Based Fire Protectionfor Existing Light-Water Nuclear PowerPlants.

By Letter dated December 12, 2012, the NRC Staff requested supplemental information in order to complete the acceptance review.

The requested supplemental information is provided in the Enclosures.

SNC considers Enclosure 6 to be sensitive information and requests that it be withheld from public disclosure pursuant to 10 CFR 2.390.

The supplemental information does not impact the 10 CFR 50.92 evaluation of "No Significant Hazards Consideration" previously provided in SNC letter NL 1893.

This letter contains no NRC commitments.

Ifyou have any questions, please contact Mr. Ken McElroy at (205) 992-7369.

Sensitive Information. Withhold from public disclosure per 10 U.S. Nuclear Regulatory Commission CFR 2.390. Decontrolled upon NL-1 2-2566 removal of Enclosure 6.

Page 2 Mr. Ajiuni states he is Nuclear Licensing Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true.

Sworn to and subscribedbefore me this 2 day ofjei., Lý 2012.

Notary Public My commission expires: U- Z 0 (3 Respectfully submitted, M. J. Ajluni Nuclear Licensing Director MJAGAL/lac

Enclosures:

1. NRC Question 1 Response

2. NRC Question 2 Response

3. NRC Question 3 Response

4. NRC Question 4 Response

5. NRC Question 5 Response

6. NRC Question 6 Response

7. NRC Question 7 Response cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Mr. T. A. Lynch, Vice President - Farley Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. B. J. Adams, Vice President - Fleet Operations RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Ms. E. A. Brown, NRR Project Manager - Farley Mr. J. R. Sowa, Senior Resident Inspector - Farley Mr. P. K. Niebaum, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer

Joseph M. Farley Nuclear Plant Enclosure 2 Question 2 Response E2-1

NRC Question 1 As discussed in Attachment A to the submittal dated September 25, 2012 (Agencywide Documents Access and Management System) Accession No. ML12279A235), the Electrical Raceway Fire Barrier System (ERFBS) Table B-I, Element 3.11.5 states

"[c]omplies" as meeting the requirements of Chapter 3. Table B-1 also states that required ERFBS is identified in Table 4-3 (Attachment C, Table C-2), but ERFBS is never identified as a required feature in this table. Clarify if ERFBS is a required fire protection feature at Joseph M. Farley Nuclear Plant, Units I and 2. If so, identify the fire area locations where ERFBS is being credited. Ifnew ERFBS modifications will be installed, then identify those as well.

SNC Response Two types of currently installed electrical raceway fire barriers are credited at FNP. These barriers are in the form of 3-hr rated enclosures in two stairwells and the use of 1-hr fire rated mineral insulated cables in additional areas. Table B-I, Element 3.11.5 lists document references for the qualification test reports for both the 1-hr fire rated mineral insulated cable and the 3-hr rated enclosures. Table 4-3 (Attachment C, Table C-2) does not specifically use the term ERFBS, but does indicate where 1-hr fire rated cables are credited and where 3-hr fire rated enclosures are used.

The following table was prepared to identify areas where these features are used:

Fire Area / Fire Zone Feature 1-013 / 0300-Ul One Hour Rated Cable 1-013 / 0227-Ul One Hour Rated Cable 1-013 / 0227-U2 One Hour Rated Cable 1-013 / 0300-U2 One Hour Rated Cable 1-041 / 0335-Ul One Hour Rated Cable 1-041 / 0343-Ul One Hour Rated Cable 1-041 / 0335-U2 One Hour Rated Cable 1-041 / 0343-U2 One Hour Rated Cable 1-042/ 0319-Ul One Hour Rated Cable 1-042 / 0339-Ul One Hour Rated Cable 1-042 / 0319-U2 One Hour Rated Cable 1-042 / 0339-U2 One Hour Rated Cable 1-S02 / 1-S02 Three Hour Rated Enclosure 2-013 / 2300-Ul One Hour Rated Cable 2-013 / 2300-U2 One Hour Rated Cable 2-041 / 2335-UI One Hour Rated Cable 2-041 / 2335-U2 One Hour Rated Cable 2-042 / 2319-U1 One Hour Rated Cable 2-042 / 2339-U1 One Hour Rated Cable 2-042 / 2319-U2 One Hour Rated Cable 2-042 / 2339-U2 One Hour Rated Cable 2-S02 / 2-S02 Three Hour Rated Enclosure New ERFBS modifications are planned to be installed. These installations are described in Table S-2, item 6, along with other circuit protection features. Each of these planned E1-2

modifications to install 1-hr fire rated materials are for risk reduction purposes and not for deterministic resolution of any Variances From Deterministic Requirements (VFDRs). The proposed modification discussion includes the cable or raceway to be protected along with an identification of the fire area where the ERFBS will be located. These modifications are discussed in Table C-1 in the comments section where required FP systems and features are listed. A required modification is listed in Table C-2, Table 4-3 for the associated fire areas where new ERFBS are being installed. Additional details are also located in the Fire Risk Evaluations for these fire areas.

All new ERFBS installed as described in Attachment S, Table S-2, will be compliant with NFPA 805, Section 3.11.5.

See markup clarification added to Attachment A, NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements.

E1-3

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference Requirements I Guidance Statement Compliance Basis ENGDOC, U419527 Rev. 1 - Qualification of Large Penetration Seal Upgrades for a Three Hour Fire Endurance Rating 3.11.4 Through Penetration 3.11.4* (b) Conduits shall be provided with an internal fire seal that has an Complies Penetration seal assemblies, including internal fire seals, are Fire Stops. (b) equivalent fire-resistive rating to that of the fire barrier through opening fire designed, qualified and installed in accordance with the stop and shall be permitted to be installed on either side of the barrier in a requirements in the referenced procedures. Conduit internal fire location that is as close to the barrier as possible. seal placement is as close to the barrier as possible as directed Exception: Openings inside conduit 4 in. (10.2 cm) or less in diameter by the notes on Drawing A-177541.

shall be sealed at the fire barrier with a fire-rated internal seal unless the conduit extends greater than 5 ft (1.5 m) on each side of the fire barrier. In this case the conduit opening shall be provided with noncombustible material to prevent the passage of smoke and hot gases. The fill depth of the material packed to a depth of 2 in. (5.1 cm) shall constitute an acceptable smoke and hot gas seal in this application.

References Document ID DRAW, A177541 Sh 19B Rev. 9 - Tray and Conduit Detail and Notes ENGDOC, A508661 Rev. 0 - Fire Rated Penetration Seal Qualification Document PROC, FNP-0-EMP-1 370.02 Rev. 16 - Installation and Repair of Penetration or Conduit Seals PROC, SS1102132 Rev. 11 - Specification for Piping, Instrumentation Tubing, Ductwork, Electrical Raceways and Firewall Penetration Seals for Farley Nuclear Plant - Units 1 & 2 3.11.5 Electrical Raceway Fire 3.11.5" Electrical Raceway Fire Barrier Systems (ERFBS). Complies Test reports reflect that the ERFBS required by Chapter 4 meet; Barrier Systems (ERFBS). ERFBS required by Chapter 4 shall be capable of resisting the fire effects the requirements of this element.

of the hazards in the area. ERFBS shall be tested in accordance with and shall meet the acceptance criteria of NRC Generic Letter 86-10, Required ERFBS are noted in Table 4-3 of the LAR.

Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Safe Shutdown Trains Within the Same Fire Required ERFBS are noted in Table 4-3 of Area." The ERFBS needs to adequately address the design requirements and limitations of supports and intervening items and their impact on the the LAR as one hour rated cable and three fire barrier system rating. The fire barrier system's ability to maintain the required nuclear safety circuits free of fire damage for a specific thermal hour rated enclosures. The term ERFBS is exposure, barrier design, raceway size and type, cable size, fill, and type not used in Table 4-3. New ERFBS shall be demonstrated.

modifications are identified in Attachment S, Exception No. 1: When the temperatures inside the fire barrier system Table S-2, Item 6 of the LAR for installation exceed the maximum temperature allowed by the acceptance criteria of Generic Letter 86-10, "Fire Endurance Acceptance Test Criteria for Fire of one hour rated electrical raceway fire Barrier Systems Used to Separate Redundant Safe Shutdown Training barriers for risk reduction purposes. These Within the Same Fire Area," Supplement 1, functionality of the cable at these elevated temperatures shall be demonstrated. Qualification are not used for deterministic compliance, demonstration of these cables shall be performed in accordance with the but all installations will comply with 3.11.5.

electrical testing requirements of Generic Letter 86-10, Supplement 1, Attachment 1, "Attachment Methods for Demonstrating Functionality of Cables Protected by Raceway Fire Barrier Systems During and After Fire Endurance Test Exposure."

Exception No. 2: ERFBS systems employed prior to the issuance of Fire Safety Analysis Data Manager (4.1) Farley 111-4 Run: 08/30/2012 16:55 Page: 48 of 49

Attachment A NEI 04-02 Table B-1 Transition of Fundamental FP Program & Design Elements Compliance NFPA 805 Ch. 3 Reference Requirements I Guidance Statement Compliance Basis Generic Letter 86-10, Supplement 1, are acceptable providing that the system successfully met the limiting end point temperature requirements as specified by the AHJ at the time of acceptance.

References Document ID DRAW, U419491 Rev. 1 - APPENDIX R, ONE-HOUR FIRE RESISTIVE, MINERAL INSULATED CONTROL CABLE FIRE TEST REPORT DRAW, U732257 Rev. 1 - QUALIFICATION OF THREE-HOUR RATED STAIRWELL ENCLOSURES Fire Safety Analysis Data Manager (4.1) Farley 111-5 Run: 08/30/2012 16:55 Page: 49 of 49

Joseph M. Farley Nuclear Plant Enclosure 2 Question 2 Response E2-1

NRC Question 2 Recovery actions are identified as the resolution of certain variances from the deterministic requirements (VFDRs) by fire area in Attachment C Table B-3 to the submittal. However many of the recovery actions required for VFDR resolutions do not appear to be included in Attachment G Table G-1.

Clarify whether Attachment G identifies all recovery actions credited in Attachment C Table B-3 VFDR resolutions. If not, provide the additional recovery actions.

Additionally, recovery actions in Attachment G do not identify the associated VFDRs. In some cases, the recovery actions can be traced back to the associated VFDRs by searching component numbers described in the action with those in the VFDR, but in many cases the correlation between recovery action and VFDR resolution cannot be made. Provide a revised Table G-1 that associates the recovery actions with the appropriate VFDR being resolved.

SNC Response Attachment G identifies all recovery actions credited in Attachment C Table B-3 VFDR resolutions. However, there is not necessarily a one to one correlation between Table B-3 and Attachment G because a single recovery action in Table G-1 may apply to more than one VFDR in Table B-3, or a single VFDR may have more than one recovery action associated with it. To help with the correlating between the two tables, an updated Table G-1 is provided which includes a column that identifies the corresponding VFDR from Table B-3.

E2-2

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 044 N1B21TI0410 RCS COLD LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-410 Ul 044 NIB21TI0413 RCS HOT LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-413 U1 044 NIB31LIO459Z PRESSURIZER LEVEL INDICATOR Monitor parameter at HSP Primary Control Station LI-459Z Ul 044 NIB31PI0444Z HOT SHUTDOWN PANEL Monitor parameter at HSP Primary Control Station PRESSURIZER PRESSURE INDICATOR PI-444Z Ul 044 N1NllLI0477A STEAM GENERATOR 1A WIDE Monitor parameter at HSP Primary Control Station RANGE LEVEL INDICATOR LI-477A Ul 044 N1N11LI0487A STEAM GENERATOR 1B WIDE Monitor parameter at HSP Primary Control Station RANGE LEVEL INDICATOR LI-487A Ul 044 NlNlLI0497A STEAM GENERATOR 1C WIDE Monitor parameter at HSP Primary Control Station RANGE LEVEL INDICATOR LI-497A Ul 044 NlN11PI3371A MAIN STEAM ATMOSPHERIC Monitor parameter at HSP Primary Control Station RELIEF 1A PRESSURE INDICATOR PI-3371A Ul 044 NIN11PI3371B MAIN STEAM ATMOSPHERIC Monitor parameter at HSP Primary Control Station RELIEF 1B PRESSURE INDICATOR PI-3371B Ul 044 NIN11PI3371C MAIN STEAM ATMOSPHERIC Monitor parameter at HSP Primary Control Station RELIEF IC PRESSURE INDICATOR PI-3371C Ul 044 NlPllLI0515 CONDENSATE STORAGE TANK Monitor parameter at HSP Primary Control Station LEVEL INDICATOR LI-515 E2-3

Table G-1 Recovery Actions and Activities Occurring at the Prrimary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 044 Q1B13HV0001 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP Ul 044 QIB13HV0002 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP Ul 044 Q1B13HV0003 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U1 044 Q1B13HV0004 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U1 044 Q1B31L0001B PRESSURIZER HEATER GROUP Transfer control and operate Primary Control Station 1B DISTRIBUTION PANEL component at HSP Ul 044 Q1B31V0027A PRESSURIZER PORV ISOLATION Transfer control and operate Primary Control Station component at HSP U1 044 QIB31V0027B PRESSURIZER PORV ISOLATION Transfer control and operate Primary Control Station component at HSP Ul 044 Q1B31V0053 PRESSURIZER PORV Transfer control and operate Primary Control Station component at HSP Ul 044 Q1B31V0061 PRESSURIZER PORV Transfer control and operate Primary Control Station component at HSP Ul 044 QIC55NI0048A GAMMA METRICS SOURCE Monitor parameter at HSP Primary Control Station RANGE MONITOR NI-48A Ul 044 Q1E21P0002B 1B CHARGING/HHSI PUMP Transfer control and operate Primary Control Station component at HSP Ul 044 Q1E21P0002C IC CHARGING/HHSI PUMP Transfer control and operate Primary Control Station component at HSP Ul 044 Q1E21V0253A LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 45 GPM component at HSP U! 044 QIE21VO253B LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 60 GPM component at HSP Ul 044 Q1E21V0253C LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 60 GPM component at HSP E2-4

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 044 QIE21V0332 SEAL WATER INJECTION Transfer control and operate Primary Control Station (QIE21HCV0186) component at HSP Ul 044 Q1E21V0336A RWST TO CHARGING PUMP Transfer control and operate Primary Control Station component at HSP Ul 044 Q1E21V0336B RWST TO CHARGING PUMP Transfer control and operate Primary Control Station component at HSP Ul 044 QlN11PV3371A MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP Ul 044 QlN11PV3371B MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP Ul 044 QlN11PV3371C MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP Ul 044 QlN11VO001A 1A SG MSIV Transfer control and operate Primary Control Station component at HSP Ul 044 QINllVO001B 1B SG MSIV Transfer control and operate Primary Control Station component at HSP Ul 044 QIN11VO001C 1C SG MSIV Transfer control and operate Primary Control Station component at HSP Ul 044 QIN12V0001A TDAFP STEAM SUPPLY Transfer control and operate Primary Control Station ISOLATION VALVE HV3235A component at HSP Ul 044 QIN12VO001B TDAFP STEAM SUPPLY Transfer control and operate Primary Control Station ISOLATION VALVE HV3235B component at HSP Ul 044 Q1N23HV3228A TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1A component at HSP Ul 044 Q1N23HV3228B TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1B component at HSP Ul 044 Q1N23HV3228C TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1C component at HSP Ul 044 Q1P17P0001A 1A COMPONENT COOLING Transfer control and operate Primary Control Station WATER PUMP component at HSP E2-5

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 044 Q1P17POOO1B 1B COMPONENT COOLING Transfer control and operate Primary Control Station WATER PUMP component at HSP Ul 044 QIP17VO030 CCW TO SEC HXS Transfer control and operate Primary Control Station component at HSP U1 044 Q1CllEOOO4A Reactor Trip Switchgear 1 Locally verify reactor trip. If the Recovery Action U1-044-PCS-036 Q1C11EOO04B Reactor Trip Switchgear 2 reactor is not tripped then Ul-044-PCS-037 manually open the reactor trip breakers at Reactor Trip Switchgear 1 & 2.

The reactor is tripped from the Control Room prior to abandonment. A confirmatory action is taken locally to verify the Control Room trip was successful.

The reactor trip breakers are opened locally if the trip actuation from the Control Room was not successful.

E2-6

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 044 QIB41POOO1A Reactor Coolant Pump 1A Locally verify at 4.16 kV Buses 1A, Recovery Action Ul-044-PCS-031 Q1B41POOO1B Reactor Coolant Pump 1B IB, & 1C (NIR15AO001, Ul-044-PCS-033 QIB41POO01C Reactor Coolant Pump 1C N1R15AO002, N1R15AO003) that Ul-044-PCS-035 RCPs IA, 1B, & 1C are tripped.

Locally trip any RCP that is still running. Remove the circuit breaker control power by opening the control power breaker for all three (3) RCPs.

The RCPs are tripped from the Control Room prior to abandonment. A confirmatory action is taken locally at the RCP switchgear to confirm that the RCP circuit breakers are open and the pumps are off. The RCP circuit breakers are opened locally if the Control Room trip was not successful. To ensure the RCPs do not spuriously restart due to hot shorts on the pump circuit breaker control circuits, the control power breaker for the RCP circuit breakers are tripped.

E2-7

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U1 044 QIP16V0518 Train B SW to Diesel Building Remove power from service water Recovery Action U1-044-PCS-145 Q1P16V0522 Unit 2 SW Supply to EDG 1B valves associated with cooling for U1-044-PCS-146 Q1P16V0530 Unit 2 SW Return from EDG EDG 1B by opening the supply Ul-044-PCS-147 Q1P16V0523 Unit 1 SW Supply to EDG 1B circuit breakers at MCC 1T. Power Ul-044-PCS-148 Q1P16V0531 Unit 1 SW Return from EDG 1B to these valves is removed prior to U1-044-PCS-149 Q1P16V0536 Train B SW Return from Diesel starting the EDG to prevent Ul-044-PCS-150 Q1P16V0538 Building spurious operation of the valves U1-044-PCS-151 Q1P16V0545 Train B SW Discharge to Pond once EDG 1B is placed in service.

Train B SW Discharge to River If SW to EDG 1B is not available from Unit 1, manually open the U2 SW supply and return valves to EDG lB. These valves are manually opened because they are normally closed.

These actions ensure that SW is provided to the EDG from at least one of the units' SW system.

U1 044 QIR15AO007 4.16 kV Switchgear Bus 1G Ifoff-site power is lost, locally Recovery Action U1-044-PCS-040 verify tripped or trip at 4.16 kV Bus U1-044-PCS-046 1G supply breakers from startup U1-044-PCS-115 transformers, non-essential loads U1-044-PCS-127 and large motor loads. Ul-044-PCS-128 Non-essential loads and large motor loads supplied from Bus 1G are confirmed off with their supply breaker open in preparation for manual loading of the EDG.

E2-8

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station Ul 044 Q1R43A0502 Emergency Diesel Generator 1B If off-site power is lost and EDG 1B Recovery Action U1-044-PCS-126 does not automatically start, manually start EDG 1B locally and close at 4.16 kV Bus 1G the EDG supply breaker.

Train B essential AC power is restored to 4.16 kV Bus 1G to ensure self-sustaining conditions for the AC power distribution system. The EDG breaker is closed with essential station service transformers connected to the bus; therefore, power to downstream non-motor loads is immediately restored once the EDG breaker is closed onto the bus.

Ul 056A QlN11PV3371A Main steam atmos relief, Operator action to align Recovery Action Ul-056A-BC-010 Service air compressor 1C emergency air system U1 056A Q1R21EO009F 7.5 KVA Inverter IF Operator action to provide Recovery Action U1-056A-HVAC-002 alternate cooling for battery charger room due to loss of Train A SW U1 056A Q1R21E0009F 7.5 KVA Inverter IF Operator action to provide Recovery Action U1-056A-HVAC-002 alternate cooling for battery charger room due to loss of fan U1 056B QIR42B1000B 125V DC Bus 1B (Train B Battery Operator action to provide Recovery Action Ul-056B-HVAC-001 Charger Room) alternate cooling for battery charger due to loss of fan Ul 056B QOR42B0001B 125V DC Bus 11 (Train B Battery Operator action to provide Recovery Action U1-056B-HVAC-001 Charger Room) alternate cooling for battery charger due to loss of SW E2-9

Table G-I Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 057 Q1R42B0001B 125V DC BUS 1B Operator action to provide Recovery Action U1-057-HVAC-002 alternate cooling to BC room due to loss of Train B SW Ul 058 Q1R42B0001B 125V DC BUS 1B Operator action to provide Recovery Action Ul-058-HVAC-001 alternate cooling for battery charger room due to loss of Train B SW and fan failure U1 060 Q1R42B0001A 125V DC BUS 1A Operator action to provide Recovery Action U1-060-HVAC-004 alternate cooling to battery charger room due to loss of Train A SW Ul 072 QIR42B0001A 125V DC BUS IA Operator action to provide Recovery Action U1-072-HVAC-005 alternate cooling for battery charger due to the loss of Train A SW Ul 072 Q1R42E0001B 125V DC BUS lB Operator action to provide Recovery Action U1-072-HVAC-006 alternate cooling for battery charger due to the loss of Train A SW Ul 075 QIR42BOO01B 125V DC BUS 1B Operator action to provide Recovery Action U1-075-HVAC-003 alternate cooling for battery charger room due to loss of Train B SW U1 076 Q2R42B0001A 125V DC BUS 1A Operator action to provide Recovery Action U1-076-HVAC-003 alternate cooling for battery charger room due to loss of Train A SW U1 1-001 OP-RECOV-RCBC-A BATTERY CHARGER 1A Operator action to provide Recovery Action UI-1-001-HVAC-004 alternate cooling for battery charger room due to loss of fan U1 1-001 OP-RECOV-RCBC-B BATTERY CHARGER 1B Operator action to provide Recovery Action UI-1-001-HVAC-005 alternate cooling for battery charger room due to loss of fan E2-10

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station Ul 1-001 OP-RECOV-BCSW- BATTERY CHARGER lB Operator action to provide Recovery Action UI-1-001-HVAC-005 B alternate cooling for battery charger room due to loss of train B SW U1 1-001 OP-RECOV-BCSW- BATTERY CHARGER 1A Operator action to provide Recovery Action Ul-1-001-HVAC-004 A alternate cooling for battery charger room due to loss of train A SW Ul 1-004 OP-RECOV-RCBC-A 125V DC BUS IA Operator action to provide Recovery Action UI-1-004-HVAC-005 alternate cooling for battery charger room due to loss of fan Ul 1-004 OP-RECOV-RCBC-B 125V DC BUS 1B Operator action to provide Recovery Action U1-1-004-HVAC-006 alternate cooling for battery charger room due to loss of fan U1 1-009 OP-RECOV-RCBC-B 125V DC BUS lB Operator action to provide Recovery Action UI-1-009-HVAC-001 alternate cooling for battery charger room due to loss of fan U1 1-009 OP-RECOV-BCSW- 125V DC BUS 18 Operator action to provide Recovery Action U-1-o009-HVAC-0O01 B alternate cooling for battery charger due to the loss of Train B SW Ul 1-012 QlN11PV3371C MAIN STEAM ATMOS RELIEF Operator action to align Recovery Action UL1-1-012-SEP-008 emergency air to PORV Ul 1-013 Q1B31V0053 PRESSURIZER PORV Operator action to align Recovery Action Ul-1-013-IA-001 emergency air system U1 1-019 Q1B31VO053 PRESSURIZER PORV Operator action to open IA lines Recovery Action U1-1-019-SEP-002 Ul 1-021 Q1R42B0001B 125V DC BUS 1B Operator action to provide Recovery Action Ul-1-021-HVAC-001 alternate cooling for battery charger room due to loss of fan Ul 1-021 Q1R42B0001B 125V DC BUS 1B Operator action to provide Recovery Action U1-1-021-HVAC-001 alternate cooling for battery charger due to loss of Train B SW E2-11

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station Ul 1-030 Q1B31V0053 PRESSURIZER PORV Operator action to locally open Recovery Action U1-1-030-IA-001 Instrument Air lines Ul 1-031 Q1B31VO061 PRESSURIZER PORV Operator action to align nitrogen Recovery Action U1-1-031-IA-001 to PORVs Ul 1-031 QlN11PV3371A Main Steam Atmos Relief Operator action to align Recovery Action U1-1-031-SEP-014 emergency air system U1 1-040 N1B21TI0410 RCS COLD LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-410 Ul 1-040 N1B21T10413 RCS HOT LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-413 U1 1-040 NIB31LI0459Z PRESSURIZER LEVEL INDICATOR Monitor parameter at HSP Primary Control Station LI-459Z U1 1-040 NIB31P10444Z HOT SHUTDOWN PANEL Monitor parameter at HSP Primary Control Station PRESSURIZER PRESSURE INDICATOR PI-444Z Ul 1-040 NlN11LI0477A STEAM GENERATOR 1A WIDE Monitor parameter at HSP Primary Control Station RANGE LEVEL INDICATOR LI-477A U1 1-040 NlN11LIO487A STEAM GENERATOR 1B WIDE Monitor parameter at HSP Primary Control Station RANGE LEVEL INDICATOR LI-487A U1 1-040 NlN11LI0497A STEAM GENERATOR 1C WIDE Monitor parameter at HSP Primary Control Station RANGE LEVEL INDICATOR LI-497A Ul1 1-040 NlN11P13371A MAIN STEAM ATMOSPHERIC Monitor parameter at HSP Primary Control Station RELIEF 1A PRESSURE INDICATOR PI-3371A U1 1-040 NIN11PI3371B MAIN STEAM ATMOSPHERIC Monitor parameter at HSP Primary Control Station RELIEF 1B PRESSURE INDICATOR PI-3371B E2-12

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station U1 1-040 NlNllPI3371C MAIN STEAM ATMOSPHERIC Monitor parameter at HSP Primary Control Station RELIEF IC PRESSURE INDICATOR PI-3371C Ul 1-040 NIPllLI0S15 CONDENSATE STORAGE TANK Monitor parameter at HSP Primary Control Station LEVEL INDICATOR LI-515 U1 1-040 QIB13HV0001 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U1 1-040 QIB13HV0002 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1B13HV0003 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP Ul 1-040 QIB13HV0004 RX VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1B31L0001B PRESSURIZER HEATER GROUP Transfer control and operate Primary Control Station 1B DISTRIBUTION PANEL component at HSP Ul 1-040 Q1B31V0027A PRESSURIZER PORV ISOLATION Transfer control and operate Primary Control Station component at HSP U1 1-040 QIB31VO027B PRESSURIZER PORV ISOLATION Transfer control and operate Primary Control Station component at HSP Ul 1-040 Q1B31V0053 PRESSURIZER PORV Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1B31VO061 PRESSURIZER PORV Transfer control and operate Primary Control Station component at HSP Ul 1-040 QIC55NI0048A GAMMA METRICS SOURCE Monitor parameter at HSP Primary Control Station RANGE MONITOR NI-48A U1 1-040 Q1E21P0002B 1B CHARGING/HHSI PUMP Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1E21P0002C 1C CHARGING/HHSI PUMP Transfer control and operate Primary Control Station component at HSP E2-13

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U1 1-040 Q1E21V0253A LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 45 GPM component at HSP U1 1-040 Q1E21V0253B LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 60 GPM component at HSP U1 1-040 Q1E21VO2S3C LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 60 GPM component at HSP U1 1-040 Q1E21V0332 SEAL WATER INJECTION Transfer control and operate Primary Control Station (QIE21HCV0186) component at HSP Ul 1-040 Q1E21V0336A RWST TO CHARGING PUMP Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1E21V0336B RWST TO CHARGING PUMP Transfer control and operate Primary Control Station component at HSP U1 1-040 QlNllPV3371A MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP Ul 1-040 QIN11PV3371B MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP U1 1-040 QlNllPV3371C MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1N11VOOO1A 1A SG MSIV Transfer control and operate Primary Control Station component at HSP Ul 1-040 QlN11V0001B 1B SG MSIV Transfer control and operate Primary Control Station component at HSP U1 1-040 Q1N11VOOO1C 1C SG MSIV Transfer control and operate Primary Control Station component at HSP Ul 1-040 Q1N12VOOO1A TDAFP STEAM SUPPLY Transfer control and operate Primary Control Station ISOLATION VALVE HV3235A component at HSP Ul 1-040 Q1N12VOOO1B TDAFP STEAM SUPPLY Transfer control and operate Primary Control Station ISOLATION VALVE HV3235B component at HSP Ul 1-040 QIN23H-V3228A TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1A component at HSP E2-14

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U1 1-040 QIN23HV3228B TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1B component at HSP Ul 1-040 QIN23HV3228C TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1C component at HSP Ul 1-040 Q1P17POOO1A 1A COMPONENT COOLING Transfer control and operate Primary Control Station WATER PUMP component at HSP Ul 1-040 Q1P17POOO1B 1B COMPONENT COOLING Transfer control and operate Primary Control Station WATER PUMP component at HSP U1 1-040 Q1P17VOO3O CCW TO SEC HXS Transfer control and operate Primary Control Station component at HSP Ul 1-040 QlCllEO004A Reactor Trip Switchgear 1 Locally verify reactor trip. If the Recovery Action U1-1-040-PCS-033 QlCllEO004B Reactor Trip Switchgear 2 reactor is not tripped then U1-1-040-PCS-035 manually open the reactor trip breakers at Reactor Trip Switchgear 1 & 2.

The reactor is tripped from the Control Room prior to abandonment. A confirmatory action is taken locally to verify the Control Room trip was successful.

The reactor trip breakers are opened locally if the trip actuation from the Control Room was not successful.

E2-15

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U 1 1-040 QIB41POO01A Reactor Coolant Pump IA Locally verify at 4.16 kV Buses IA, Recovery Action U1-1-040-PCS-027 Q1B41P0001B Reactor Coolant Pump 1B 1B, & 1C (N1R15A0001, Ul-1-040-PCS-029 QIB41POO01C Reactor Coolant Pump 1C N1R15A0002, N1R15A0003) that Ul-1-040-PCS-031 RCPs 1A, 1B, & 1C are tripped.

Locally trip any RCP that is still running. Remove the circuit breaker control power by opening the control power breaker for all three (3) RCPs.

The RCPs are tripped from the Control Room prior to abandonment. A confirmatory action is taken locally at the RCP switchgear to confirm that the RCP circuit breakers are open and the pumps are off. The RCP circuit breakers are opened locally if the Control Room trip was not successful. To ensure the RCPs do not spuriously restart due to hot shorts on the pump circuit breaker control circuits, the control power breaker for the RCP circuit breakers are tripped.

U1 1-040 QIP16V0519 Train A SW to Diesel Building Remove power from service water Recovery Action UI-1-040-PCS-182 Q1P16V0526 Unit I SW Supply to EDG 1-2A valves associated with cooling for U1-1-040-PCS-184 Q1P16V0534 Unit 1 SW Return from EDG 1- EDG 1-2A by opening the supply U1-1-040-PCS-186 Q1P16V0537 2A circuit breakers at MCC 1S and IN. U1-1-040-PCS-188 QIP16V0539 Train A SW Return from Diesel U1-1-040-PCS-192 Q1P16V0546 Building Power to these valves is removed Ul-1040-PCS-337 Train A SW Discharge to Pond prior to starting the EDG to Ul-1-040-PCS-339 Train A SW Discharge to River prevent spurious operation of the Ul-1-040-PCS-343 valves once EDG 1-2A is placed in service.

E2-16

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U1 1-040 Q1R15A0006 4.16 kV Switchgear Bus IF Ifoff-site power is lost, locally Recovery Action Ul-1-040-PCS-037 verify tripped or trip at 4.16 kV Bus Ul-1-040-PCS-051 IF supply breakers from startup Ul-1-040-PCS-261 transformers, non-essential loads Ul-1-040-PCS-263 and large motor loads.

Non-essential loads and large motor loads supplied from Bus IF are confirmed off with their supply breaker open in preparation for manual loading of the EDG.

Ul 1-040 QSR43A0501 Emergency Diesel Generator 1- Ifoff-site power is lost and EDG 1- Recovery Action U1-1-040-PCS-291 2A 2A does not automatically start, manually start EDG 1-2A locally and close at 4.16 kV Bus IF the EDG supply breaker.

Train A essential AC power is restored to 4.16 kV Bus iF to ensure self-sustaining conditions for the AC power distribution system. The EDG breaker is closed with essential station service transformers connected to the bus; therefore, power to downstream non-motor loads is immediately restored once the EDG breaker is closed onto the bus.

Ul 1-041 QIN11PV3371A Main Steam Atmos Relief Operator action to align Recovery Action Ul-1-041-SEP-028 QIN11PV3371B emergency air system Ul-1-041-SEP-029 Ul 1-041 Q1B31VO061 PRESSURIZER PORV Operator action to align nitrogen Recovery Action Ul-1-041-IA-001 to PORVs E2-17

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR I I rnnfrnl qt:kfilnn Prim rn Ul 1-061 QIR42BOO01A 125V DC BUS 1A Operator action to provide Recovery Action U1-061-HVAC-001 alternate cooling to battery charger due to loss of train A SW U1 1-075 QINIIPV3371A Main Steam Atmos Relief Valves Operator action to align Recovery Action UI-1-075-SEP-020 QIN11PV3371B emergency air system U1-1-075-SEP-021 QINllPV3371C U1-1-075-SEP-022 Ul 1-076 Q1B31V0053 Pressurizer PORV Operator action to locally open IA Recovery Action U1-1-076-IA-001 lines U1 1-DU- Q1R42B0001B 125V DC Bus lB Operator action to provide Recovery Action U1-1-DU-DGSWIS-B-HVAC-DGSWIS-B alternate cooling for battery 002 charger room Ul 1-SVB1-A QIR42BOO01A 125V DC BUS IA Operator action to provide Recovery Action U1-1-SVB1-A-HVAC-003 alternate cooling to BC room due to loss of train A SW Ul 1-SVB2-A Q1R42B0001A 125V DC BUS 1A Operator action to provide Recovery Action U1-1-SVB2-A-HVAC-003 alternate cooling to BC room due to loss of Train A SW Ul 1-SVB2-B Q1R42B0001B 125V DC BUS 1B Operator action to provide Recovery Action U1-1-SVB2-B-HVAC-002 alternate cooling to BC room due to loss of train B SW Ul 1-SVB4-A Q1R42B0001A 125V DC BUS 1A Operator action to provide Recovery Action Ul-1-SVB4-A-HVAC-003 alternate cooling to BC room due to loss of Train A SW Ul 1-SVB4-B Q1R42B0001B 125V DC BUS 1B Operator action to provide Recovery Action U1-1-SVB4-B-HVAC-002 alternate cooling to battery charger room due to loss of train B SW Ul YARD- Q1R42B0001B 125V DC Bus 1B Operator action to provide Recovery Action U1-YARD-SWIS-HVAC-003 SWIS alternate cooling to battery charger room due to loss of train B SW E2-18

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U2 044 N2B21TI0410 RCS COLD LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-410 U2 044 N2B21TI0413 RCS HOT LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-413 U2 044 N2B31LI0459Z PRESSURIZER LEVEL INDICATOR Monitor parameter at HSP Primary Control Station LI-459Z U2 044 N2B31PI0444Z HOT SHUTDOWN PANEL Monitor parameter at HSP Primary Control Station PRESSURIZER PRESSURE INDICATOR PI-444Z U2 044 N2N11PI3371A STEAM GENERATOR 2A ATMOS Monitor parameter at HSP Primary Control Station RELIEF PRESSURE INDICATOR U2 044 N2N11PI3371B STEAM GENERATOR 2B ATMOS Monitor parameter at HSP Primary Control Station RELIEF PRESSURE INDICATOR U2 044 N2N11PI3371C STEAM GENERATOR 2C ATMOS Monitor parameter at HSP Primary Control Station RELIEF PRESSURE INDICATOR U2 044 N2P11LI0515 CONDENSATE STORAGE TANK Monitor parameter at HSP Primary Control Station LEVEL INDICATOR LI-515 U2 044 Q2B13HV0001 REACTOR VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U2 044 Q2B13HV0002 REACTOR VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U2 044 Q2B13HV0003 REACTOR VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U2 044 Q2B13HV0004 REACTOR VESSEL HEAD VENT Transfer control and operate Primary Control Station component at HSP U2 044 Q2B31L0001B PRESSURIZER HEATER GROUP Transfer control and operate Primary Control Station 2B DISTRIBUTION PANEL component at HSP U2 044 Q2B31V0027A PORV BLOCK VALVE Transfer control and operate Primary Control Station component at HSP E2-19

Table G-1 Recovery Actions and Activities Occurring at the Pirimary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station U2 044 Q2B31V0027B PORV BLOCK VALVE Transfer control and operate Primary Control Station component at HSP U2 044 Q28331V0053 PRESSURIZER POWER Transfer control and operate Primary Control Station OPERATED RELIEF component at HSP U2 044 Q2B33IV0061 PRESSURIZER POWER Transfer control and operate Primary Control Station OPERATED RELIEF VALVE component at HSP U2 044 Q2C55NI0048A GAMMA METRICS SOURCE Monitor parameter at HSP Primary Control Station RANGE MONITOR NI-48A U2 044 Q2E2 1POOO2B 2B CHARGING PUMP Transfer control and operate Primary Control Station component at HSP U2 044 Q2E21POOO2C 2C CHARGING PUMP Transfer control and operate Primary Control Station component at HSP U2 044 Q2E21V0253A LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 45 GPM component at HSP U2 044 Q2E21V0253B LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 60 GPM component at HSP U2 044 Q2E21V0253C LETDOWN ORIFICE ISOLATION Transfer control and operate Primary Control Station 60 GPM component at HSP U2 044 Q2E21V0332 SEAL WATER INJECTION Transfer control and operate Primary Control Station (Q2E21HCV0186) component at HSP U2 044 Q2E21V0336A RWST TO CHARGING PUMP Transfer control and operate Primary Control Station component at HSP U2 044 Q2E21V0336B RWST TO CHARGING PUMP Transfer control and operate Primary Control Station component at HSP U2 044 Q2N11L-10477A STEAM GENERATOR 2A (WIDE Transfer control and operate Primary Control Station RANGE) component at HSP U2 044 Q2N11LIO487A STEAM GENERATOR 2B (WIDE Transfer control and operate Primary Control Station RANGE) component at HSP U2 044 Q2N11LI0497A STEAM GENERATOR 2C (WIDE Transfer control and operate Primary Control Station RANGE) component at HSP E2-20

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station U2 044 Q2N11PV3371A MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP U2 044 Q2N11PV3371B MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP U2 044 Q2N11PV3371C MAIN STEAM ATMOS RELIEF Transfer control and operate Primary Control Station component at HSP U2 044 Q2N11V0001A 2A SG MSIV Transfer control and operate Primary Control Station component at HSP U2 044 Q2N11V0001B 2B SG MSIV Transfer control and operate Primary Control Station component at HSP U2 044 Q2N11V0001C 2C SG MSIV Transfer control and operate Primary Control Station component at HSP U2 044 Q2N12V0001A TDAFP STEAM SUPPLY Transfer control and operate Primary Control Station ISOLATION VALVE HV3235A component at HSP U2 044 Q2N12V0001B TDAFP STEAM SUPPLY Transfer control and operate Primary Control Station ISOLATION VALVE HV3235B component at HSP U2 044 Q2N23HV3228A TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1A component at HSP U2 044 Q2N23HV3228B TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1B component at HSP U2 044 Q2N23HV3228C TDAFW SUPPLY TO STEAM Transfer control and operate Primary Control Station GENERATOR 1C component at HSP U2 044 Q2P17P0001A 2A COMPONENT COOLING Transfer control and operate Primary Control Station WATER PUMP component at HSP U2 044 Q2P17P0001B 2B COMPONENT COOLING Transfer control and operate Primary Control Station WATER PUMP component at HSP U2 044 Q2P17V0030 CCW TO SEC HXS Transfer control and operate Primary Control Station component at HSP E2-21

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station U2 044 Q2C11EOOO4A Reactor Trip Switchgear 1 Locally verify reactor trip. Ifthe Recovery Action U2-044-PCS-156 Q2C11EOOO4B Reactor Trip Switchgear 2 reactor is not tripped then U2-044-PCS-157 manually open the reactor trip breakers at Reactor Trip Switchgear 1 & 2.

The reactor is tripped from the Control Room prior to abandonment. A confirmatory action is taken locally to verify the Control Room trip was successful.

The reactor trip breakers are opened locally if the trip actuation from the Control Room was not successful.

E2-22

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U2 044 Q2B41POOO1A Reactor Coolant Pump 2A Locally verify at 4.16 kV Buses 2A, Recovery Action U2-044-PCS-079 Q2B41POOO1B Reactor Coolant Pump 2B 2B, & 2C (N2R15AO001, U2-044-PCS-080 Q2B41POO01C Reactor Coolant Pump 2C N2R15A0002, N2R15A0003) that U2-044-PCS-153 RCPs 2A, 2B, & 2C are tripped. U2-044-PCS-154 Locally trip any RCP that is still U2-044-PCS-155 running. Remove the circuit breaker control power by opening the control power breaker for all three (3) RCPs.

The RCPs are tripped from the Control Room prior to abandonment. A confirmatory action is taken locally at the RCP switchgear to confirm that the RCP circuit breakers are open and the pumps are off. The RCP circuit breakers are opened locally if the Control Room trip was not successful. To ensure the RCPs do not spuriously restart due to hot shorts on the pump circuit breaker control circuits, the control power breaker for the RCP circuit breakers are tripped.

E2-23

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station U2 044 Q2P16V0518 Train B SW to Diesel Building Remove power from service water Recovery Action U2-044-PCS-007 Q1P16V0592 Unit I SW Supply to EDG 2B valves associated with cooling for U2-044-PCS-051 Q1P16V0593 Unit 1 SW Return from EDG 2B EDG 2B by opening the supply U2-044-PCS-052 Q2P16V0592 Unit 2 SW Supply to EDG 2B circuit breakers at MCC 1P, 2DD, U2-044-PCS-053 Q2P16V0593 Unit 2 SW Return from EDG 2B 2V and 2T. Power to these valves U2-044-PCS-054 Q2P16V0536 Train B SW Return from Diesel is removed prior to starting the U2-044-PCS-055 Q2P16V0538 Building EDG to prevent spurious operation U2-044-PCS-145 Q2P16V0545 Train B SW Discharge to Pond of the valves once EDG 2B is U2-044-PCS-146 Train BSW Discharge to River placed in service.

If SW to EDG 2B is not available from Unit 2, manually open the U1 SW supply and return valves to EDG 2B. These valves are manually opened because they are normally closed.

These actions ensure that SW is provided to the EDG from at least one of the units' SW system.

U2 044 Q2R15A0007 4.16 kV Switchgear Bus 2G If off-site power is lost, locally Recovery Action U2-044-PCS-058 verify tripped or trip at 4.16 kV Bus Us-044-PCS-067 2G supply breakers from startup U2-044-PCS-121 transformers, non-essential loads U2-044-PCS-151 and large motor loads.

Non-essential loads and large motor loads supplied from Bus 2G are confirmed off with their supply breaker open in preparation for manual loading of the EDG.

E2-24

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action I VFDR Primary Control Station U2 044 Q2R43A0505 Emergency Diesel Generator 2B If off-site power is lost and EDG 2B Recovery Action U2-044-PCS-066 does not automatically start, manually start EDG 2B locally and close at 4.16 kV Bus 2G the EDG supply breaker.

Train B essential AC power is restored to 4.16 kV Bus 2G to ensure self-sustaining conditions for the AC power distribution system. The EDG breaker is closed with essential station service transformers connected to the bus; therefore, power to downstream non-motor loads is immediately restored once the EDG breaker is closed onto the bus.

U2 056A O2R21E0009F 7.5 KVA INVERTER 2F Operator action to provide Recovery Action U2-056A-HVAC-001 alternate cooling to battery charger due to loss of train A SW U2 056A Q2B31V0061 PRESSURIZER POWER Operator action to align Recovery Action U2-056A-SEP-003 OPERATED RELIEF VALVE emergency air system U2 056A Q2N11PV3371A MAIN STEAM ATMOS RELIEF Operator action to align Recovery Action U2-056A-BC-012 emergency air system U2 056B QO2R42B0001B 125V DC BUS 2B Operator action to provide Recovery Action U2-056B-HVAC-001 alternate cooling to battery charging room due to loss of train B SW U2 056B Q2R42B0001B 125V DC BUS 2B Operator action to provide Recovery Action U2-056B-HVAC-001 alternate cooling to battery charging room due to fan failure.

E2-25

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U2 058 Q2R42B0001B 125V DC Bus 2B Operator action to provide Recovery Action U2-058-HVAC-001 alternate cooling for battery charger room due to fan failure U2 058 Q2R42B0001B 125V DC Bus 2B Operator action to provide Recovery Action U2-058-HVAC-001 alternate cooling for battery charger room due to loss of Train B SW U2 059 Q2R42B0001B 125V DC BUS 2B Operator action to provide Recovery Action U2-059-HVAC-001 alternate cooling to BC room due to loss of Train B SW U2 059 Q2R42B0001B 125V DC BUS 2B Operator action to provide Recovery Action U2-059-HVAC-001 alternate cooling to BC room due to fan failure U2 060 Q2R42B0001A 125V DC Bus 2A Operator action to provide Recovery Action U2-060-HVAC-001 alternate cooling for battery charger room due to loss of Train A SW U2 061 Q2R42B0001A 125V DC BUS 2A Operator action to provide Recovery Action U2-061-HVAC-001 alternate cooling for battery charger room due to loss of Train A SW U2 061 Q2R42B0001A 125V DC BUS 2A Operator action to provide Recovery Action U2-061-HVAC-001 alternate cooling for battery charger room due to loss of fan U2 061 Q2N11PV3371A MAIN STEAM ATMOS RELIEF, Operator action to align Recovery Action U2-061-BC-013 Service air compressor 2A/2C emergency air system U2-061-SEP-008 U2-061-SEP-009 U2 075 Q2R42B0001B 125V DC BUS 2B Operator action to provide Recovery Action U2-075-HVAC-001 alternate cooling for battery charger due to loss of Train B SW E2-26

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U2 076 Q2R42BOO01A 125V DC BUS 2A Operator action to provide Recovery Action U2-076-HVAC-001 alternate cooling for battery charger room due to loss of Train A SW U2 1-008 Q2R42B0001A 125V DC BUS 2A Operator action to provide Recovery Action U2-1-008-HVAC-001 alternate cooling for battery charger room due to loss of Train A SW U2 1-021 Q2R42BOO01B 125V DC BUS 2B Operator action to provide Recovery Action U2-1-021-HVAC-001 alternate cooling for battery charger room due to loss of Train B SW U2 1-031 Q2R42BOO01A 125V DC BUS 2A Operator action to provide Recovery Action U2-1-031-HVAC-001 alternate cooling for battery charger room due to loss of Train A SW U2 1-041 Q2R42BOO01A 125V DC BUS 2A Operator action to provide Recovery Action U2-1-041-HVAC-001 alternate cooling for battery charger U2 1-042 Q2R42BOO01A 125V DC Bus 2A Operator action to provide Recovery Action U2-1-042-HVAC-001 alternate cooling to BC room due to loss of Train A SW U2 1-075 Q2R42BOO01A 125V DC Bus 2A Operator action to provide Recovery Action U2-1-075-HVAC-001 alternate cooling for battery charger room due to loss of Train A SW U2 1-DU- Q2R42BOOO1A 125V DC BUS 2A Operator action to provide Recovery Action U2-1-DU-DGSWIS-A-HVAC-DGSWIS-A alternate cooling for battery 001 charger room due to loss of Train A SW U2 2-001 Q2N23POO01A 2A MDAFW PUMP Operator action to provide Recovery Action U2-2-001-HVAC-004 alternated cooling to AFW pump room due to loss of fan E2-27

Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station(s)

Fire Area Component Component Description Actions Recovery Action / VFDR Primary Control Station U2 2-004 Q2R42B0001B 125V DC BUS 2B Operator action to provide Recovery Action U2-2-004-HVAC-002 alternate cooling for battery charger room due to loss of Train B SW U2 2-005 Q2N11PV3371A MAIN STEAM ATMOS RELIEF Operator action to align Recovery Action U2-2-00S-SEP-010 emergency air to ARVs U2 2-008 Q2B31V0061 PRESSURIZER POWER Operator action to align nitrogen Recovery Action U2-2-008-IA-001 OPERATED RELIEF VALVE to PORVs U2 2-009 Q2R42BOO01B 125V DC BUS 2B Operator action to provide Recovery Action U2-2-009-HVAC-001 alternate cooling for battery charger room due to loss of fan U2 2-009 Q2R42BOO01B 125V DC BUS 2B Operator action to provide Recovery Action U2-2-009-HVAC-001 alternate cooling for battery charger room due loss of Train B SW U2 2-018 Q2N11PV3371C MAIN STEAM ATMOS RELIEF Operator aligns emergency air to Recovery Action U2-2-018-SEP-016 ARV U2 2-019 Q2B31V0053 PRESSURIZER POWER Operator action to open IA valves Recovery Action U2-2-019-SEP-003 OPERATED RELIEF U2 2-030 Q2B31VO053 Pressurizer Power Operated Operator action to locally open IA Recovery Action U2-2-030-IA-001 Relief Valve valves U2 2-031 Q2B31V0061 PRESSURIZER POWER Operator action to align nitrogen Recovery Action U2-2-031-IA-001 OPERATED RELIEF VALVE to PORVs U2 2-040 N2B21TI0410 RCS COLD LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-410 U2 2-040 N2B21TI0413 RCS HOT LEG TEMPERATURE Monitor parameter at HSP Primary Control Station INDICATOR TI-413 U2 2-040 N2B3110459Z PRESSURIZER LEVEL INDICATOR Monitor parameter at HSP Primary Control Station LI-459Z E2-28

Joseph M. Farley Nuclear Plant Enclosure 3 Question 3 Response E3-1

NRC Question 3 Section 4.2.1.2 of the submittal states that "safe and stable ... assumes the plant to be taken subcritical and maintained in anyone of the modes of hot standby, hot shutdown, cold shutdown, or refueling conditions." However, the performance goals listed in Attachment C Table B-3 appear to show achievement of only hot shutdown goals, not cold shutdown or refueling. Clarify the safe and stable condition for the plant. If cold shutdown, explain why the Nuclear Safety Capability Analyses and fire risk evaluations do not include this equipment.

SNC Response The Nuclear Safety Criteria Analysis (NSCA) for Farley Nuclear Plant (FNP) has been developed in accordance with NFPA 805 requirements and clarifications contained in FAQ 08-0054, Demonstrating Compliance with Chapter 4 of National Fire ProtectionAssociation 805.

The concept of Safe and Stable is captured in Clauses 1.3.1 and 1.6.56 of NFPA 805:

1.3.1 Nuclear Safety Goal. The nuclear safety goal is to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition.

1.6.56 Safe and Stable Conditions. For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling.

As specified by NFPA 805, the fundamental concept of "Safe and Stable" is applicable to all operating modes; hence, Southern Nuclear interprets the question being asked as pertaining to what elements of the NSCA are contained in the "At Power" fire-origination portion of the NSCA vs. the non-power operations (NPO) portion of the analysis. As discussed in FAQ 08-0054, the specific transition point dictates what equipment and performance goals are contained in the Attachment C Table B-3. The performance goals listed in Section C, Table B-3 are established to align with the performance goals associated with a fire originating while the plant is in either Modes 1 or 2. It assumes the reactor is shut down and then maintained in a safe and stable condition in Mode 3 for an unspecified duration (Note: the "At Power" model has been augmented with additional pressure control features - pressurizer heaters, auxiliary spray, and pressurizer PORVs - to ensure positive pressure control independent of time). The "At Power" model includes all equipment necessary to maintain Safe and Stable conditions down to the cut-in temperature and pressure for RHR operation (Mode 4). Consistent with NFPA 805, the NPO analysis includes all equipment associated with Modes 3 and below. The NPO analysis assumes the fire originates while the plant is not in Modes 1 or 2. The wording in LAR Section 4.2.1.2 will be revised to clarify the "At Power" and NPO analyses, and thereby remove any potential ambiguity.

E3-2

Joseph M. Farley Nuclear Plant Enclosure 4 Question 4 Response E4-1

NRC Question 4 Attachment S Table S-2 Item #11 to the submittal identifies a proposed modification to "install a new code compliant fire detection system and provide a code compliance evaluation to support the new system." Provide a more explicit description of this modification. Address whether there will only be detectors, or will it also include new local and main panels. Discuss whether this is intended to be installed throughout the plant. Provide the version of the National Fire Protection Association (NFPA)-72, "National Fire Alarm and Signaling Code," will be used for this new system. It appears that Attachment C, Table C-2 (Table 4-3) identifies detection used in each fire area/zone where it is credited for NFPA 805, "Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants." Address whether this table is based on existing detection or the new detection system, or both.

SNC Response The detection system at FNP is credited as a required fire protection system for defense-in-depth and it supports existing engineering equivalency evaluations. The current system is obsolete and is being replaced by a project that is being performed in parallel with the transition to NFPA 805.

The project scope includes the replacement of the fire detection and alarm system (detection devices, local and main alarm panels, and a graphical user interface computer in the main control room). The scope of buildings/areas under detection system replacement includes the following:

- Unit 1 Auxiliary Building

• Unit 1 Containment Building

• Unit 1 Turbine Building (Battery Room only)

• Diesel Generator Building

• River Water Intake Structure a Service Water Intake Structure 0 Water Treatment Building M Unit 2 Auxiliary Building 0 Unit 2 Containment Building M Unit 2 Turbine Building (Battery Room only)

The stated objective of the project modification is that the final installed fire detection and alarm system will be compliant with NFPA 72 (2007 edition). The stated intent is that where possible a one for one replacement of the existing fire detection devices shall be the preferred approach, unless such an approach will not result in compliance with NFPA 72 (2007) or there is a technical reason why an alternate technology should be employed.

The goal of the fire detection and alarm systems modification is to maintain conformance with the applicable sections of NFPA 72 (2007 edition) in a performance based approach. As changes to the fire detection system are proposed, this performance based approach will be used to reconcile the existing and proposed NFPA 805 licensing basis with the proposed design to ensure that proper coverage of the fire detection system is maintained to support both the E4-2

existing FNP and proposed NFPA 805 Fire Protection Licensing Basis as well as conformance to NFPA 72 (2007 edition).

The control equipment (type, quantity and install locations) and power supply requirements are specific to the new system, and will be provided in a manner that meets operational requirements, optimizes system design and meets the code requirements.

Attachment C Table C-2 (table 4-3) is based on both the existing and the new detection system.

E4-3

Joseph M. Farley Nuclear Plant Enclosure 5 Question 5 Response E5-1

NRC Question 5 Address whether Table V-2 includes all supporting requirements identified as either capability category (CC)-l or not met by the peer review. In doing so, clearly identify those SRs being carried forward as CC-I (or not met), and provide justification as to why doing so is acceptable for transition.

SNC Response Table V-1 has been modified to include the column F&O #, and to change the "SR" column to "Related SR" (Currently 'SR'). The F&O # will identify the specific Finding that was written, while the Related SRs column will identify all applicable SRs that are associated with it.

Included with each SR will be the CC that the SR was assessed at, these may be Not Met, Met CC 1, Met CC II, Met CC Ill, Met 1/11, Met Il/111, Met or Not Applicable. Most Findings have been addressed by updating either the analysis, the documentation, or both. In some cases there are SRs that were Met CC I which were determined to be acceptable for this application. These specific SRs are identified below and have also been identified within Table V-I. The justification for carrying forward the CC I SRs is provided within the disposition of the Finding as found in Table V-I.

• FSS-C1

• FSS-C2

• FSS-E3 The modified Table V-1 is attached. Changes are indicated by double underline.

Please note that typos were corrected.

E5-2

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations Related SR Topic Status Finding/Observation Disposition CS-B 1(CC I) The Farley breaker Closed The PRA components are not explicitly discussed in The Farley circuit analysis coordination a coordination calculation. Calculation SE- calculation, SE-C051326701-002, documentation was C051326701-002 is titled NSCA components; has been updated to address all identified to be however, informal review has determined that PRA coordination concerns. This incomplete based on components are addressed. The information to update identified two panels that the Farley Fire PRA determine the status of coordination for PRA were found to not be coordinated; components being components consists of informal queries and all other panels were credited, spreadsheets. Supporting Requirement CS-Bl-01 dispositioned as being Category II requires all buses credited in the Fire acceptable. The two panels are PRA to be analyzed for proper over current N1R1 9L00504 and coordination and protection. N2R19L00504. Calculation PRA-BC-F-1 1-003 (Cable Selection Revise calculation SE-C051326701-002 to formally and Detailed Circuit Analysis) has validate that PRA buses are addressed for proper been updated to address this coordination and incorporate results into the Fire coordination issue. Based on PRA model ad needed. these conclusions these two panels have been failed in every scenario for the Farley Fire PRA.

Associated Circuits Analysis Common Power Supply And Common Enclosure calculation, SE-C051326701-002 has been updated to reflect this update.

The Farley Component Selection Report, PRA-BC-F-1L-002, has also been updated to reflect the inclusion of these panels to the UNL list, see Page F-24 and F-180, Appendix F.

S 0 CS-B1iCC If The Farley breaker Closed E-068 identifies cases where the cable lengths of An analysis was completed that coordination electrical loads were credited to demonstrate reviewed the panels that credited calculations use cable selective coordination for the Cable Spreading cable length as part of the length as part of the room. This assumption is only valid for the justification for coordination. The justification for proper Appendix R fire where the equipment and cables entire function of these panels Page 1 01 Rev 01 Page1I E5-3

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O#

_SR Topic Status Finding/Observation Disposition coordination. This is are assumed damaged for the entire fire area. was then failed for any fire that not a justifiable Supporting Requirement CS-B1-01 Category II impacted the cable within the disposition for use in requires all buses credited in the Fire PRA to be identified length. Once the length the Fire PRA. analyzed for proper over current coordination and requirement was met the function protection. of that cable was the only function failed. See calculations, Analyze impacted PRA buses for proper Associated Circuits Analysis coordination and incorporate results into the Fire Common Power Supply And PRA model. Common Enclosure calculation, SE-C051326701-002, and PRA-BC-F-1 1-003 (Cable Selection and Detailed Circuit Analysis) for more information. A modification is also scheduled to improve coordination for six additional 125VDC load distribution panels per unit. For further information on the modification of these panels see Plant Modifications Committed in Table S-2 of Attachment S.

FQ-A3-Q1 FQ-A3 M(Met Appendix L of Closed A non-suppression probability of 3.04E-5 is used for The Farley MCR analysis has NUREG-CRJ6850 had the Main Control Room (NSP-0401

• basic events), been updated to accurately apply been incorrectly A review of the Scenario Development report, the the non-suppression factors as applied to the Main Summary Report, and the MCR Report did not appropriate to the Main Control Control Board locate the justification of this probability. Based on Board scenarios. The Farley Fire scenarios in the Farley discussion with the Farley team, the values were Scenario Report discusses the Fire PRA. The ignition derived from NUREG/CR-6850, Attachment L. A scenario development process frequencies have since review of that Attachment did not support a NSP for the Main Control Board and been updated to below 1E-4 under the best of circumstances. A the use of Appendix L in section accurately apply NSP of 2E-2 (similar to other NSP events) would 13.1.2 of PRA-BC-E-11-014.

Appendix L. make MCR fire the highest contributor to plant risk.

Re-evaluate the NSP used for the Control Room Rev 01 Page 2 E5-4

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations F&O_ B SR Topic Status Finding/Observation Disposition and document the evaluation clearly in one of the reports.

FQ-Cl01 FQ-C1 Me Possible combination Closed A review of cutsets for different sequences found Every COMBO event is evaluated events were missing multiple HRA combinations that are not being and incorporated in the fire PRA.

from the cutset results replaced by a COMBO* event and do not appear to An updated dependency analysis for the Fire PRA be evaluated for dependence. One such was completed after the peer model, combination is 1OP-MS032-IH-F and OAR B 1--- review findings were addressed

-H-F which has a combined failure probability of in the model. The latest results approximately 7E-5. A review of the HRA Calculator of the dependency analysis can package supplied shows that no eval uation was be found in the Human Reliability performed for this combination of events. Other Analysis for Fire Events, PRA-HRA combinations could also be missing, BC-F-11-016.

particularly with new operator actions added for the fire scenarios. HRA dependence could significantly increase cutsets since the rule file makes HRAs independent unless the events are replaced by an evaluated combination.

Review the FPRA cutsets without recovery (all events set to screening values) to ensure that all important combinations are evaluated.

Q101 -D (Not Met The CCFP for Farley Closed In Section 3 of the Farley Nuclear Plant Summary The Farley Fire PRA has FQ-F1 (Not Met) Fire PRA was much Report, Farley reports a CDF of 9.65E-05/year and continued to evolve and be FBM-B14 (Me greater than what the a LERF of 1.92E-5/year. This yields a Conditional refined throughout the analysis.

PRM-Cl (Men FPIE number was. Containment Failure Probability (CCFP) of 1.99E- Currently the CCFP is at a much UNC-A1 (Not Metf After continued 01. For the FPIE PRA, the reported CDF was of the more reasonable value based on refinement the Fire order of 3.5E-05/year and the reported LERF was of the final CDF and LERF results.

PRA CCFP has the order of 2E-07/year. This translates to a CCFP The results and insights related decreased to a more of about 4E-03. This is a significant difference, to CDF and LERF can be found reasonable value as especially when considering that the leading in the Farley Summary Report compared with the contributor to LERF for the FPIE PRA, SGTR, is not section 3 of PRA-BC-E-1 1-017.

FPIE. applicable for fire. This yields inconsistent results.

Rev 01 Page 3 E5-5

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O_ Belated SR Topic Status Finding/Observation Disposition While the current results may be correct, Farley needs to look at the contributors to LERF to explain the basis for the high CCFP with respect to the FPIE PRA CCFP. Farley should look at sequences where the fire not only causes core damage but also directly affects containment integrity. Two likely candidates are sequences that lead to a new ISLOCA scenario and sequences that lead to containment isolation scenarios.

FQEl-01 FQ-E1=(, The Farley Fire PRA Closed The summary report lists and describes significant The Farley Summary report documentation did not contributors to core damage and LERF. The back includes additional details accurately address the references require consideration of analysis issues describing the types of reviews types of reviews that which are not described in the report as having been that were completed on the were performed during done. For example, the back references require a Farley Fire PRA. The type of the scenario cutset review the results of the PRA for modeling review and the detailed cutset review sessions. consistency, a review of results to determine that reviews are described in section the flag event settings, mutually exclusive event C. 1 of Appendix C in the rules, and recovery rules yield logical results, a Summary Report, PRA-BC-F-1 1-review of contributors for reasonableness and a 017.

review of the importance results for reasonableness.

Appendix F notes that these were accomplished and typically refers back to Appendix C. Appendix C does not describe these reviews as being accomplished, nor does it describe the results of the reviews. In addition, back Reference D5 requires a review of non-significant cutsets for reasonableness.

Appendix F states that dominant cutsets were reviewed and those that were reduced in frequency to non-significance as a result of the review constitute the review of non-significant cutsets. This does not satisfy the requirement to review non-significant cutsets. Non-significant cutsets generated in the solution of the model need to be reviewed to confirm that their frequency is not underestimated due to modeling errors.

Rev 01 Page 4 E5-6

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations E&O

__SR Topic Status Finding/Observation Disposition Expand the discussion of model solution and review in the summary report to indicate that required review items have been accomplished.

FO-Fl01 FQ-FlI(,,*4~ Level of detail Closed The documentation of the FPRA results does not The Farley Summary report has describing the risk adequately describe the top risk contributors such been updated to reflect the significant scenarios that it is clear why these scenarios, basic events, insights by reviewing the top was identified as not and human actions are dominant. Based on other contributors for CDF and LERF.

being sufficient in findings (FQ-Di-01 and FQ-El-01), it is not clear This describes the fire induced detail that the Farley team understands the bases for impacts as well as the random these top scenarios. Results presentation is failures. The resolution of this important for PRA acceptability. Understanding of finding is found in Appendix C of the PRA results is necessary for performing any RI PRA-BC-F-1 1-017.

application to support the plant Provide more detailed discussions of the fire impacts and results to represent a strong understanding of the fire scenarios.

FSS-A2-01 FSS-A2=e4U Target set definition in Closed FNP is missing the basis for not including targets A review of the full room burnout FSS-A4 (Met Fire Zones (FZs) that outside the fire compartment for full room burnout scenarios was completed that do not have fire rated scenarios. For full room burnout scenarios, all looked for open boundaries to the boundaries on all sides targets in the fire compartment are included, adjoining FZs and the possible as it relates to However, there is no documented basis for not interactions that could take place.

scenarios that are including targets outside the fire compartment for full For some particular fire areas a classified as full room room burnout scenarios. If the compartment has an scenario was postulated that burnouts. opening to an adjacent compartment, it was not would fail all targets within the fire verified that targets in the adjacent compartment area. However, in most cases it would be outside of the ZOI of all the ignition was determined that there was sources in the compartment analyzed for full room no ignition source near the open burnout. boundary that would impact targets in an adjoining FZ. The See F&O PP-B3-01 (F) for a possible resolution. Farley Fire Scenario Report includes discussion of the scenario development process Rev 01 Page 5 E5-7

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations E&O_ Related SR Topic Status Finding/Observation Disposition for these specific cases in section 3.1.1 of PRA-BC-F-1 1-014.

FSS-B2-01 FSS-B1 (Met) The Main Control Closed An office workstation fire scenario is discussed in The Farley Main Control Room FSS-B2 C Room Abandonment the documentation, but is not fully justified. The Abandonment Calculation calculation identifies workstation fire scenario is potentially the most includes the discussion of the the potential for a significant fire scenario considered. workstation fire in section B.8 as workstation fire but a sensitivity to the analysis with does not describe the Provide better documentation of how the the results shown in Table B-8.

fire type in significant workstation fire was modeled and the results of this NUREG/CR-6850 does not detail fire scenario. provide any basis for this type of fire from an ignition frequency standpoint. Therefore it is not included as one of the potential ignition sources in the base calculation. A review of the sensitivity analysis involving the workstation shows that the analysis is not sensitive to that type of fire given the design of the Main Control Room envelope.

FSS-C1-01 FSS-C1IJfl.) The Farley Fire PRA Closed Two-point fire intensity model that encompass low The development of fire FSS-G1 (Me!) does not employ the likelihood, but potentially risk contributing, fire scenarios for the Farley Fire PRA use of a two point fire events were not used in all cases. Fire scenarios did not identify any instances modeling treatment in were done with ignition sources characterized with where further analysis resolution the development of the one fire intensity, would be gained by the treatment fire scenarios. as inferred by the requirements To reach Capability Category II, use a two-point for CC II and CC Ill. The intensity model for all ignition sources. implications of retaining the CC I treatment in lieu of refining as Utility Comment: The development of fire described for CC II or CC IIIis scenarios for the Farley Fire PRA did not identify potentially a higher calculated any instances where further analysis resolution CDF contribution. The CC I would be gained by the treatment as inferred by the treatment inherently will not result requirements for CC II and CC Ill. The implications in under-estimation of fire risk.

Rev 01 Page 6 E5-8

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V.1 Fire PRA Peer Review - Facts and Observations E&O" RelatedSR Topic Status Finding/Observation Disposition of retaining the CC I treatment in lieu of refining as As such, the current treatment is described for CC II or CC IIIis potentially a higher conservative. Provided this calculated CDF contribution. The CC I treatment treatment does not result in inherently will not result in under-estimation of fire masking of risk increases in risk. As such, the current treatment is conservative, future applications, further Provided this treatment does not result in masking refinements are not considered of risk increases in future applications, further necessary.

refinements are not considered necessary.

Response: The SR stipulates that a two-point model is required for CC-Il. As you stated in your comment, Farley feels that the one-point model is conservative and justified. This would be viewed as the proposed resolution, but the F&O stands.

F C FSS-C2(CCI The Farley Fire PRA Closed Ignition source intensity were characterized such The only readily available FSS-G1 (Me did not characterize that fire is initiated at full peak intensity and ignition reference for a time dependent the ignition source sources that are significant contributors to fire risk growth rate that could be intensity for a time were not characterized using a realistic time- considered in the analysis is 12 dependent growth rate dependant fire growth profile. Generic methods minutes as recommended in in the scenario from the Hughes Associates Generic Fire Modeling NUREG/CR-6850. The treatment development. Treatments were used to characterize ignition would involve a t grow*t rate. If___, .......-1Formatted: Superscript I source intensity. These generic methods did not a particular source/target incorporate fire growth curves, interaction has a spacing where the target is at the critical Characterize ignition sources that are significant damage spacing threshold, such contributors to fire risk using a realistic time- a treatment may provide some dependant fire growth profile, benefit as successful suppression with that time period would Utility Comment: The only readily available prevent target damage.

reference for a time dependent growth rate that However, if the target is located could be considered in the analysis is 12 minutes as well within the calculated damage recommended in NUREG/CR-6850. The treatment distance, the corresponding time would involve a t2 growth rate. If a particular to reaching the damage threshold source/target interaction has a spacing where the is very short and effectively target is at the critical damage spacing threshold, precludes any meaningful credit Rev 01 Page 7 E5-9

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations Bn.JatRJe,.SR Topic Status Finding/Observation Disposition such a treatment may provide some benefit as for suppression. In the case of successful suppression with that time period would the Farley Fire PRA, the majority prevent target damage. However, if the target is of the target spacing for the located well within the calculated damage distance, dominant risk contributors is such the corresponding time to reaching the damage that no meaningful credit for threshold is very short and effectively precludes any suppression is available. In other meaningful credit for suppression. In the case of the dominant risk contributors, the Farley Fire PRA, the majority of the target spacing scenario involves high energy for the dominant risk contributors is such that no arcing fault (HEAF) events were meaningful credit for suppression is available. In no growth time is applicable. The other dominant risk contributors, the scenario implications of retaining the CC I involves high energy arcing fault (HEAF) events treatment in lieu of refining as were no growth time is applicable. The implications described for CC Il/111 is of retaining the CC I treatment in lieu of refining as potentially a slightly higher described for CC Il/111 is potentially a slightly higher calculated CDF contribution. The calculated CDF contribution. The CC I treatment CC I treatment inherently will not inherently will not result in under-estimation of fire result in under-estimation of fire risk. As such, the current treatment is conservative, risk. As such, the current Provided this treatment does not result in masking treatment is conservative.

of risk increases in future applications, further Provided this treatment does not refinements are not considered necessary. result in masking of risk increases in future applications, Response: The Farley modeling was found to be further refinements are not consistent with CC-I but did not meet the considered necessary.

requirements of CC-Il. The comment provides the basis for stating that the existing treatment is adequate. It does not provide evidence that a time-dependent heat release rate model was used.

FSS-Di-Ol FSS-D1 Lvlet The treatment of Closed The fire modeling tools selected for use are The Farley Scenario FSS-D2 (Met) secondary appropriate for evaluating the zone of influence development notebook was FSS-D3 (CC combustibles was not associated with individual fixed and transient ignition updated to include additional clearly defined in the sources, but do not provide for estimating fire details on how the treatment of scenario development growth and damage behavior for fire scenarios secondary combustibles is dealt documentation. involving ignition and fire spread on secondary with during scenario combustibles. With the generic fire modeling development. Further Rev 01 Page 8 E5-10

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O# Related SR Topic Status Finding/Observation Disposition treatment selected for this fire PRA, there does not information regarding this finding appear to be a way to model fire growth on can be found in section 4.0 of secondary combustibles. Consequently, the extent PRA-BC-F-11-014.

of fire development cannot be modeled.

Where secondary combustibles are located within the zone of influence, develop methods for estimating fire growth on secondary combustibles and the damage caused by this additional fire development.

FSS-D7-01 FSS-D7joMt The Fire PRA credits Closed SR FSS-D7 requires credited fire suppression Supporting documentation has the in cabinet C02 systems to be installed and maintained in been included in the Farley Fire system installed at accordance with applicable codes and standards, Scenario report to further discuss Farley. There was no and the credited systems must be in fully the in cabinet C02 suppression documentation operational state during plant operation. These system and the associated test provided to support the requirements are not met, but fire suppression and inspection procedures that availability of this systems are still being credited. As noted in the are credited in the Fire PRA. It system. Conclusions section of Document # 0005-0012-002- has also been identified that the 002-04 (Hughes Associates), "The other main system does require concern with the systems installed at FNP is the modifications, such as periodic maintenance and subsequent corrective mechanical equipment and action. Firstly, the plant procedures for inspection, detection upgrades, to be made testing and maintenance (ITM) do not address a few to make the system operable as key activities required by NFPA 12. Secondly, the designed. This is found in prioritization of work orders sometimes results in section 8.1.1 of PRA-BC-F-1 1-extended impairments (e.g., observed CR / work 014.

request tags over two years old), which negatively affects the fire protection program objective to maintain working systems." Credit is being taken for fire suppression systems that do not meet the requirements of FSS-D7 for taking this credit.

Verify that credited fire suppression systems are Rev 01 Page 9 E5-11

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O_# lWtdSR Topic Status FindinglObservation Disposition installed and maintained in accordance with applicable codes and standards and demonstrate that credited systems are in a fully operable state during plant operation.

FSS-D7-02 FSS-D7 J The non-suppression Closed In Tables 13-1 through 13-12 the equation eA(- The application of the MCR probability that was lambda*t) was used to calculate the non- abandonment non suppression originally used to suppression probability for MCR abandonment probability has been re-evaluated calculate the MCR scenarios. The control room lambda value from using the floor value of 1.00E-03 abandonment Table P-2 was selected. The time, t, was obtained for all bins that are determined to frequency was un- through the CFAST runs and plugged into the reach the abandonment conservative based on equation. In scenarios in which the time to threshold. The results of this direction provided in abandonment was greater than 25 minutes a review are identified in section 13 Appendix P of nominal NSP of 0 was selected. A NSP of 0 should of PRA-BC-F-1 1-014.

NUREG-CR/6850. not be assumed for these cases. Instead, it is suggested to run the CFAST cases longer than 25 minutes such that the analysis can credit a larger time with no abandonment conditions reached (i.e. if the case is ran to 60 minutes with no abandonment conditions reached, t can be credited up to 60 minutes) and still use the e(Ia1bdalt) equation to calculate NSP.

Additionally, the MCR equipment rooms are normally unoccupied and NSP should be associated with the electrical equipment room vs. the control room. If the control room lambda is used, a basis should be developed why the control room lambda is more appropriate than the electrical cabinet lambda. If the control room lambda basis has been justified, then a sensitivity analysis should be performed using the lambda of electrical fires. This calculation can be non-conservative.

FSS-D8-01 FSS-D8 The Farley Fire PRA Closed Note 8 associated with SR FSS-D8 suggests The Fire PRA was first developed FSS-D1 1 (Not Met) does not look at the consideration of the time available to suppress a fire without credit for suppression or Rev 01 Page 10 E5-12

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&# RelgatdedSR Topic Status Finding/Observation Disposition time available for a prior to target damage and specific features of detection, the target set for a suppression system to physical analysis units and fire scenarios under given scenario was based on the successfully suppress analysis that might impact suppression system ignition source type. Further in a fire before target activation and coverage. Such consideration is not the analysis credit for the existing damage. documented. Credit is taken for automatic fire detection and suppression, and suppression in some scenarios without in some cases plant consideration of the factors required under this SR. modifications, systems were credited. For these cases where Perform an analysis that considers the time the credit was taken the target available to suppress a fire prior to target damage set was not changed based on and the specific features of the PAUs and fire the time to suppression or scenarios under analysis to determine what impact distance to target. Instead a they have on suppression system activation and conservative approach was taken coverage. to leave the original target set included in the Fire PRA along with the failure rate of the suppression system, therefore not requiring a review of damage time vs. suppression time. This is found in section 8.0 of PRA-BC-F- 11-014.

F FSS-E3,(CC I The Farley Closed Supporting requirement E3 asks to provide a mean The documentation has been documentation did not value of, and statistical representation of, the updated to include discussions address the uncertainty intervals for the parameters used for fire related to the uncertainty for fire uncertainty related to modeling the fire scenarios. Farley performed fire modeling. See Table D-1 of the the use of fire size and heat release rate selection in accordance Farley Fire PRA Summary report, modeling for the fire with NUREG/CR-6850 and/or applicable FAQs. PRA-BC-F-11-017.

scenarios. However, the methods for developing the statistical representation of the uncertainty intervals and mean The associated SR was values currently do not exist. However, this is not dispositioned as CC I which is reported in the documentation. judged to be sufficient given the two concerns noted.

In the documentation, explain that it is understood that methods for developing the statistical Rev 01 Page 11 E5-13

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations F&O_# Related SR Topic Status Finding/Observation Disposition representation of the uncertainty intervals and mean values currently do not exist.

Utility Comment: This specific F&O was issued against a technical element and the indicated resolution involves a documentation clarification.

This documentation clarification will be implemented.

FSS-Fl-01 FSS-FIoM The exposed structural Closed Section 2.11 of the FNP Summary Report A review and analysis was FS-F2 jj steel evaluation was (FNPSummaryReport final.pdf) claims that, "The completed of the structures at FSS-F3 (N4A* not originally Structural Steel Evaluation performed to evaluate Farley for both units to determine performed as part of the potential for fire to impact structural steel the amount of exposed structure the Farley Fire PRA. capacity which could impact fire compartment steel that is susceptible to fire boundaries is documented in the FNP Fire PRA damage and ultimately leading to Report PRA-BC-F-1 1-014, Rev. 0, Fire Scenarios a building collapse. The analysis Report [5]." This documentation was not found in the concluded that there is a potential referenced report. for this scenario to occur in the Include in the Fire Scenarios report the structural Turbine Building. This scenario steel evaluation identified in final Summary Report has been added and is and update self-assessment. accounted for in the total plant risk and delta risk calculations.

This is found in section 10.5 of PRA-BC-F-1 1-014.

FSS-G6-01 FSS-G6,PJI The Farley Fire PRA Closed The Multi-compartment analysis identifies several The Farley Fire PRA MCA FQ-A3aM MCA analysis was areas where further evaluation is required. This analysis has been completed with incomplete at the time evaluation has not been completed to either screen all scenarios being evaluated.

of review with many the zone or develop a fire scenario based on multi- The HGLJMCA report has been open items. compartment fire. A screening of the multi- updated to show the final results compartment scenarios were done, those that were for the analysis. This is found in screened out were not included in the quantification. Attachment B of PRA-BC-F The multi-compartment scenarios flagged for further 015.

evaluation are in Table 3-1 of the Multi-Compartment Analysis. Further evaluation is still being worked on, so these scenarios have not been Rev 01 Page 12 E5-14

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations

" RelatedSR Topic Status Finding/Observation Disposition included in quantification. Given the current CDF, the MCA could increase risk above 1E-4/yr.

Complete the MCA to either quantify the PAUs where a fire could spread to an adjacent PAUs or screen the PAUs for MCA FSS-1A I1/11) Non-Fire PRA targets Closed For fire scenarios considered during the peer review The scenario development FSS-D10 (CC II/II1 were removed from walkdown, the nature and characteristics of the database has been re-populated FSS-HI=,,* the database. Leading damage target set were different in three different with all target set information, FSS-H10 Met__ to inconsistencies sets provided for review, including the computer targets specifically modeled in between the scenario printout of the fire scenario summary and two sets of the Fire PRA and those that are development sheets walkdown notes. One consistent set of not. The scenario printout sheets and what was documentation should be maintained in a retrievable found in Appendix A of the Fire identified in the field, format. scenario development report contain all targets identified Include all relevant target sets in the computer- during the walk down phase based documentation and handle by disposition regardless of the relationship to those targets that are not risk significant for a Fire PRA components. This is particular scenario, found in Appendix A of PRA-BC-F-1 1-014.

Ei5-01 FSS-H5,*j* The Farley Closed The generic fire modeling tool referenced in the Fire The documentation has been documentation did not Scenario Report, Reference 6 (Hughes generic updated to include discussions address the treatment) is used for generic treatment of ignition related to the uncertainty for fire uncertainty related to sources as an approach to bound many scenarios, modeling in response to SR FSS-the use of fire but its use does not provide uncertainty treatment E3. See Table D-1 of the Farley modeling for the fire on a fire scenario basis. Fire PRA Summary report, PRA-scenarios. BC-F-1 1-017.

Provide uncertainty evaluations at least generically for those scenarios that use the generic treatment tools and on a case by case basis for the sources that use additional detailed fire modeling to further describe the scenarios used.

Utility Comment: This specific F&O is inconsistent Rev Of Page 13 E5-15

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V.1 Fire PRA Peer Review - Facts and Observations FBe,.dSR Topic Status Finding/Observation Disposition with F&O FSS-E3-01. The indicated resolution for FSS-E3-01 states in part that the analysis documentation should be enhanced to note that methods for developing the statistical representation of the uncertainty intervals and mean values currently do not exist. However, F&O FSS-H5-01 then asks to undertake evaluations to address uncertainty. These this latter F&O should be revised so that it is consistent with FSS-E3-01.

Response: The F&Os address the specific SR requirements. The response to F&O FSS-E3-01 may be used to justify the treatment of uncertainty for FSS but the F&O documents compliance with the standard and as such remains.

LGN-A7D2U FSS-A1 (Met) Newly installed Closed During the walkdown - ignition sources (specifically The Farley Fire PRA has been in IGN-A7,WIW potential Ignition electrical cabinets) were found in the plant that is development for some time. The sources were identified not listed on the list of ignition sources for the ignition source walk down and in the field that were particular PAU. Specific examples include scenario development were some not included as part of N1RIL0001 in the cable spreading room and of the first tasks that were the original scenario N1R15AO02X and N1R5AO03X in the switchgear completed as part of this development. room. A walkdown and/or review of plant analysis. A qualitative review of modification is necessary to ensure the plant FPRA the panels identified during the reflects the as built as operated configuration. peer review walkdown showed no significant change in the plant This issue may be due to new plant equipment that CDF. This is based on the fire was added after the initial ignition frequency zones these panels were located walkdown - nevertheless the fire PRA should be in and the level of scenario reconciled to include these new ignition sources. development already included in these fire zones. The panels are located in a part of the room that already contains detailed scenarios and the introduction of the new sources are not Rev 01 Page 14 E5-16

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations F&O#_ RelatedSR Topic Status Finding/Observation Disposition expected to change the target set of adjoin scenarios. Section 3.5 of the Summary Report, PRA-BC-F-11-017, provides steps that will be taken to account for changes in the plant design that have occurred since the initial Fire PRA development.

IGN-A7-04 IGN-A7=(=J The yard transformers Closed Bins 27-29 have not been filled. Large Yard The Farley Fire PRA task 6 has had been incorrectly Transformers have been incorrectly binned in Bin 23 been updated to accurately binned during the Task ("indoor transformers"). It is clearly stated in represent the transformers 6 development and NUREG/CR-6850 that large yard transformers are located in the yard to their should be moved to not part of this count. As a result each large outdoor applicable bins and have been their appropriate bins. transformers (MT, UAT, SuT) should be binned in removed from bin 23. The both Bin 27 (Yard Transformer - Catastrophic) and frequency per component has Bin 28 (Yard Transformer- Non Catastrophic). been updated accordingly and Additionally, Bin 29, Transformer Yard - Others, used for the applicable scenarios.

should also be filled. See Appendix C of Plant Partitioning and Fire Ignition Since Bin 23 may have been misinterpreted, it is Frequency for Farley Fire PRA, suggested that indoor transformers typically PRA-BC-F-1 1-009.

associated with essential lighting, etc. be looked at for applicability in the FPRA if not already evaluated.

Indoor transformers over 45kVA should be included in the count for this bin.

IGN-A7=(Mj IGN-A705 The Farley scenario Closed During the walkdown of the Bravo 4160 Switchgear The Farley Fire PRA has been development did not room, it was observed that the Foxtrot Switchgear updated to accurately correct the accurately account for was split between 2 PAUs. The switchgear had a scenario development to account the frequency split count of 15 vertical sections. PAU 335 had a count for the ignition source split between the two fire of 8 switchgear vertical sections and PAU 343 had a between the two fire zones of the zones as it was count of 8 vertical sections. This is a clear example SWGR room. The ignition source identified in the field. of inadequate PAU boundary. count of the SWGRs has not been changed to reflect the Recommend that the PAU such that the Foxtrot accurate number of cubicles.

Rev 01 Page 15 E5-17

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O#_ Related SR Topic Status Finding/Observation Disposition switchgear is contained in one PAU and the count of This change would result in a the entire switchgear should be 15 vertical sections. non-significant impact to the total In cases where ignition sources have been split plant ignition frequency based on between PAUs the count should be verified correct. the total count for Bin 15. The ignition frequency for the scenarios related to the SWGRs are accurately represented.

These updates can be found in the Farley Scenario report, Appendix A, PRA-BC-F-1 1-014.

IGN-A9-01 IGN-A9(Me-tf The transient factors in Closed PAU 2321 (Sample Panel Room) has a transient fire The transient ignition frequency the ignition frequency frequency of zero. Similar to the first page of allocation has been re-visited for development had Appendix B, a storage factor of "low" or 1 should be the Farley Fire PRA based on identified fire zones chosen such that 2321 has a non-zero transient fire this finding. The appropriate that had a 0 factor frequency. Right now 2321 has a non-zero ignition changes have been made to which led to a frequency due to a small number of cable in the accurately reflect the transient frequency of 0. area filling Bins 11 and 12. ignitions sources located within each fire zone. These updates A non-zero transient factor should be filled in. were made in Fartey Plant Partitioning and Ignition Source Task 1 and 6 report, PRA-BC-F-11-003, and carried into the ignition source calculation for the scenario development, PRA-BC-F-1 1-014.

RP-B3-01 FSS-A2 (Met_ The Farley Fire PRA Closed SNOC has not provided sufficient evidence that Fire The Farley scenario development PP-B2 (Not Met) did not contain Zone PAUs were evaluated for fire resistance report has been updated to PP-B3INtM sufficient information capabilities of barriers, nor was there sufficient provide more details on the PP-C3 (Not Met) on scenario evidence that credited spatial separations were scenario development based on development with analyzed. Specific examples are cited in PRA-BC- the ignition source and target respect to the crediting F-11-001, Section 2.2, for PAUs that use "natural identification process. This can of fire barriers, divisions." The document cites that the lack of fire be found in the Farley Scenario barriers between these PAUs will be evaluated Development report, PRA-BC-F-during the MCA. However, the MCA analysis11-014, section 3.1.1. The Rev 01 Page 16 E5-18

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O_# RelatedSR Topic Status Finding/Observation Disposition appears to only discuss hot layer issues, and does impact on the Hot Gas Layer and not consider whether a fire propagates outside of Multi-Compartment Analysis has the PAU or ifthere is a zone of influence and target also been revisited to assure that damage outside of the PAU. Another example of the boundaries of the rooms have where spatial separation is credited is Tool Room been adequately represented in 0441. the calculation of the volumes.

Full room burnout scenarios are developed and quantified, but without sufficient evidence that fire barriers or spatial separation issues have been evaluated. It appears that specific PAUs are screened from having multi-compartment impacts without consideration of fire propagation or ZOI impact across spatial divisions.

SNOC has presented a plan to resolve the Fire Zone PAU vs. Fire Area PAU issue. Implementation of this plan is sufficient to address the issues identified in PP-B2 and PP-B3. In the plan, Fire Areas will be treated as PAUs. Particularly, SNOC staff have acknowledged that for "full burn" and "base case" fire scenarios, they will review and document the capabilities of barriers and the appropriateness of credited spatial separations, and will not inappropriately credit barriers or spatial separations for fire scenarios. The plan includes the following:

1. Those APs that have one or more boundaries that are not physical features or are not rated fire barriers will be identified and a requirement will be added to clarify that this must be recognized in the development of fire scenarios. There will be confirmation that the results of the above Rev 01 Page 17 E5-19

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations E&# Related SR Topic Status Finding/Observation Disposition have been observed and documented.

2. Enhance the documentation to acknowledge the crediting of non-rated physical boundaries and provide a basis recognizing that the justification will rely on physical observations during plant walkdowns or through equivalent means as well as general construction methods (masonry block wall, concrete walls, etc.).

3. Address the nature and consequence of anticipated fire events for all APs for which explicit fire scenarios are not developed (base cases) and confirm that the results are appropriate given the boundaries for the AP.

4. Confirm that bounding room burn-out cases are not used for any APs that are not fully bounded by physical fire barriers, and that there is a justification for crediting those physical barriers.

5. Confirm that the resulting analysis does not change (reduce) the level of resolution associated with the existing fire scenarios developed to support the requirements of SRs associated with FSS.

Modify the hot gas layer and multi-compartment analysis (MCA) so that any unnecessary conservatism caused by using a smaller volume artificially caused by an assumed AP boundary are removed.

Rev 01 Page 18 E5-20

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations Related SR Topic Status Finding/Observation Disposition PP-B1 (Met1 The Farley Fire PRA Closed Plant personnel have given verbal assurance that The Farley Task 1 and 6 report PP-B7 (Met) did not contain plant walkdowns have been performed to confirm identifies the ignition sources PP-C3 ot sufficient information the plant partitioning boundaries. It is reasonable to identified in each fire zone. The on scenario presume that the fire protection engineer would results of the walkdowns are development with perform this walkdown task. In addition, walkdowns input into a database that respect to the were performed to support the Fire PRA ignition contains the necessary identification of fire frequency task. Furthermore, some notes were information related to Task 1 and barriers, found as further evidence that some walkdowns 6. This database is considered to were performed. However, documentation of the be the controlled copy of the plant partitioning walkdown is not readily available results of these tasks. These for peer review. SR PP-C3 requires documentation results are found in Appendix D of key or unique features of the partitioning of report PRA-BC-F-1 1-009.

elements for each physical analysis unit. SR PP-B7 Section 3.1.1 of the Farley requires a confirmatory walkdown of partitioning Scenario Report, PRA-BC-F-1 1-elements. 014, describes the process of identifying applicable scenarios Include Plant Partitioning walkdown sheets as part based on the ignition source, of PRA secondary documentation, and refer to the surrounding targets and fire walkdown sheets in PRA-BC-F-11-001, Farley Fire barriers.

PRA Tasks 1 & 6, Plant Partitioning and Fire Ignition Frequency. In particular, fire barriers and spatial separations that are credited in fire scenarios should be validated. When where no prior documentation can be found, new walkdowns may be required.

PP-C3-02 I The Farley Fire PRA Closed Fire Zones are identified as Fire PRA plant analysis The Farley Fire PRA PP-B7 (Met) did not contain units in PRA-BC-F-11-00. Fire PRA staff have documentation has been updated PP-C3

• sufficient information expressed that the Fire Areas, not Fire Zones, to be consistent in the naming on scenario should be assessed as the PAUs. However, the convention throughout the development with Fire Zone PAU form the basis for initial PAU ignition analysis concerning the use of respect to the frequency, whole room burns, and initial screening PAU and fire zone. The 'rooms' documentation of fire in later PRA analysis Fire Zones as PAUs are used at Farley are considered fire barriers, consistently and extensively in the FPRA zones, while the fire areas are documentation. There is a disconnect between the considered PAUs. The Task I PAUs defined in PRA-BC-F-1 1-00 and SNOC staffs and 6 report, Plant Partitioning Rev 01 Page 19 E5-21

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O_# Relate-dSR Topic Status Finding/Observation Disposition statements of what constitutes a PAU. This and Ignition Frequency PRA-BC-adversely affected the review of the Plant F-11-009, Cable selection and Partitioning technical element. SNOC desires to call Detailed Circuit Analysis PRA-the entities that are currently described as Fire Zone BC-F-11-003, and the Farley PAUs as Administrative Partition, and to treat Fire Scenario report, PRA-BC-F-1 1-Areas as PAUs. 014 contain this clarification.

F&O PP-B3-01 identifies an acceptable plan to address the technical issues around the definition of PAUs, that Fire Areas, not Fire Zones, form the basis for PAUs. Fire Zones and similar entities will be identified as "Administrative Partitions" (AP).

Since the term "Physical Analysis Unit" or PAU is extensively in Fire PRA documentation to describe Fire Zone PAUs, all Fire PRA documents should be reviewed and revised to call these compartments Administrative Partition. Furthermore, the term "Administrative Partition" (AP) should be defined in the PP documentation and the APs descriptions (formally, Fire Zone PAUs), should be retained.

PRM-B2-01 PRM-B2 (Not Metl The Farley internal Closed Internal Events PRA peer review exceptions and Table 1 of Fire PRA logic events finding had only deficiencies have only partially been dispositioned. Development, PRA-BC-F-1 1-004 been partially Table 1 of the Fire Model document (PRA-BC-F-1 1- has been updated to address all addressed in respect 004 V0a) lists some of the internal events findings, internal events PRA findings and to the impact on the but not all. All findings included in the internal their impacts on the fire PRA.

Fire PRA. events peer review must be included and disposed in the PRM notebook. Disposition of findings could not be verified. Discussion with Southern Company personnel indicated that some of the findings had not been addressed.

Expand Table 1 of the Fire Model document to include all findings. Describe the impact of the Rev 01 Page 20 E5-22

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-I Fire PRA Peer Review - Facts and Observations E&O

__SR Topic Status Finding/Observation Disposition finding on the fire PRA. For those that impact model elements applicable to the fire analysis, describe the resolution in sufficient detail to allow a reviewer to conclude the finding has been dispositioned.

PRM-Cl-01 PRM-CI=*,"f The RCP shutdown Closed The new RCP shutdown seals are included in the Fire PRA has been developed seals were not fault tree model but are not described in Appendix based on internal events PRA adequately discussed B. Appendix B should be revised to describe these model having model of RCP in the documentation new seals and their impact on RCP seal failure flow shutdown seal. Section 2.0, for the Fire PRA model rate. The Fire PRA modeling pertaining to RCP seal Appendix B of Fire PRA logic development, failure is not adequately described in PRA-BC-F-1 1- Development, PRA-BC-F-1 1-004 004. has been updated to add RCP shutdown seal modeling.

Revise Appendix B to describe the new shutdown seals and their impact.

UNC-Al-01 UNC-AI= *=NLet The Farley fire PRA Closed Farley presents the CDF results in Section 3.0 of the Appendix D of the Farley provided Train A and B Summary Report. The way the results are Summary report, PRA-BC-F-1 1-CDF results but did not presented are as an annualize CDF for Train A 017, has been updated with a define total plant CDF. operating and an annualize CDF for Train B revised parametric uncertainty The parametric operating and both are called total plant CDF. analysis for both CDF and LERF uncertainty analysis There is no discussion as to what these two CDF for Train A and B individually.

should be more values meant or a value for the "true" plant CDF. In The quality of the analysis was specific in scope and Appendix D of the Summary Report, Farley presents improved by applying the Monte use a greater sampling the results of their parametric uncertainty analysis Carlo method with 50,000 size. for CDF. Although not documented, this appears to samples. The resulting curves be for CDF related to Train A Operating only. The are well behaved and the parametric uncertainty analysis was performed calculated means show minimal using the Latin Hypercube method with only 1000 difference when compared to the samples. The resulting curve was not well behaved point estimates.

and the calculated mean is well below the point estimate in Section 3. Discussion of how the total plant CDF/LERF is calculated is also As a start, Farley needs to define what the two provided in the Summary Report.

results, annualize CDF for Train A operating and an This describes how the Train A annualize CDF for Train B operating, mean and a and Train B results are averaged Rev 01 Page 21 E5-23

Southern Nuclear Operating Company Attachment V - Fire PRA Quality Table V-1 Fire PRA Peer Review - Facts and Observations F&O__# Rlate.LSR Topic Status Finding/Observation Disposition single total Plant CDF needs to be presented. This together to obtain the total plant will probably be the average of the original two CDF/LERF.

values. For the uncertainty analysis, Farley needs to document what is covered by the analysis, Train A results, Train B results or both. Farley did run an uncertainty case using 10,000 samples and the results seemed to be better behaved. Farley is running an uncertainty case with 50,000 samples which is consistent with their FPIE PRA process.

The results of this analysis should be presented in Appendix D in the Summary Report instead of the current analysis.

UNC-A1-Q2 UNC-AI1bot-M The Farley Closed Farley did quantify the fire-related LERF for Unit 1 The Farley Summary report has documentation did not but failed to meet the requirements from LE-F2 and been updated to reflect the adequately address LE-F3 from Section 2 which require that "REVIEW insights by reviewing the top the review of LERF contributors for reasonableness (e.g., to assure contributors for LERF. This scenarios in the excessive conservatisms have not skewed the describes the fire induced analysis to show that results, level of plant-specificity is appropriate for impacts as well as the random the appropriate significant contributors, etc.)" and "IDENTIFY and failures. The resolution of this reviews had been characterize the LERF sources of model uncertainty finding is found in Appendix C of completed. and related assumptions in a manner consistent PRA-BC-F-11-017.

with the applicable requirements of Tables 2-2.7-2(d) and 2-2.7-2(e)." As discussed in F&O FQ-D1-01, the calculated LERF and CCFP indicate that there some potential issues with the LERF calculation.

See F&O FQ-D1-01 (F) and perform the reasonableness reviews after requantifying 1 - The identified SR is deemed to be acceotable at CC I and will be carried forward as such. The Oustification for this acceptability is provided within the disposition for each identified Findino.

Rev 01 Page 22 E5-24

Southern Nuclear Operatinq Connvany Attachment V - Fire PRA Quality 1

Table V-2 Fire PRA- Category I Summary SR Topic Status N/A

- All Fire PRAs SRs characterized as Capability Category I were identified as Findings in the Peer Review. Refer to Table V-1 for identification and resolution of findings.,. -.. [ Deleted: - - -Section Break (Next Page). - - )

Rev 01 Page 23 E5-25

Joseph M. Farley Nuclear Plant Enclosure 7 Question 7 Response E7-1

NRC Question 7 Section 5.5 and Attachment S of the submittal proposes a completion schedule for plant modifications of November 6, 2017. Provide justification for this schedule.

SNC Response Earlier in the NFPA-805 project and fire PRA development, the modification scope was unclear. Budgetary projections for modifications were placed in the out years to allow sufficient time for conceptual design development. The date of November 6, 2017 was chosen to provide time to ensure capital and O&M budgets were aligned to fund the respective NFPA 805 modification projects.

SNC intends to implement NFPA-805 at Farley in the first half of 2015. However, the date of November 6, 2017 was also chosen as the scheduled date for the completion of modifications because it is expected to take two refueling outages on each unit to complete the implementation of the modifications. Farley performs refueling outages on a staggered train basis where only a single train is taken out of service for major work.

Major work is then performed on the opposite train on the following refueling outage. In addition, more detailed conceptual designs have shown some NFPA 805 modifications are more extensive than originally envisioned and may require two refueling outages for completion.

The site Plant Health Committee (PHC) approved the proposed modifications on November 1, 2012 and proposed a schedule for implementation. The next step in the process is for the Plant Review Committee (PRC) to determine if the budget can support the proposed schedule for modification implementation and to approve the budget funding for these modifications. The PRC meeting to review these projects is scheduled for December, 2012. The modifications will be added to the design work list for 2013 and the actual design process will begin in early 2013 for the modifications.

SNC understands that contingency actions will be required for any modifications not completed at the time NFPA 805 is implemented at Farley in 2015 until all modifications are completed.

E7-2