Text

Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generational Plants
	ML12117A052
Person / Time
Site:	Duane Arnold
Issue date:	04/23/2012
From:	Wells P; NextEra Energy Duane Arnold
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NG-12-0177, TAC ME6818
	Download: ML12117A052 (167)
	v • d • e

NExTera" ENERGY 7 DARNOL April 23, 2012 NG-12-0177 10 CFR 50.90 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Duane Arnold Energy Center Docket No. 50-331 Renewed Op. License No. DPR-49 Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Liaht Water Reactor Generatina Plants

References:

1) License Amendment Request (TSCR-128): Transition to 10 CFR 50.48(c) - NFPA 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants (2001 Edition), NG-11-0267, dated August 5, 2011

2) Clarification of Information Contained in License Amendment Request (TSCR-128): Transition to 10 CFR 50.48(c) - NFPA 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants (2001 Edition), NG-1 1-384, dated October 14, 2011 In the Reference 1 letter, as clarified by Reference 2, NextEra Energy Duane Arnold, LLC (hereafter NextEra Energy Duane Arnold) submitted a License Amendment Request for the Duane Arnold Energy Center (DAEC) pursuant to 10 CFR 50.90. Subsequently, the NRC Staff requested, via electronic mail, additional information regarding that application.

As a result of discussions with the Staff held on February 22, 2012, NextEra Energy Duane Arnold committed to providing responses to a portion of the requested information by April 23, 2012. Attachment 1 to this letter contains the requested information. Per those same discussions, NextEra Energy Duane Arnold committed to providing responses to the remaining requested information by May 23, 2012.

'400(o NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324

Document Control Desk NG-12-0177 Page 2 of 3 This additional information does not impact the 10 CFR 50.92 evaluation of "No Significant Hazards Consideration" previously provided in the referenced application.

This does not make changes to any existing commitments and makes the following new commitments.

RAI Response Number Description Fire Protection Engineering 1 Reference to the exemption from 10 CFR 50, Appendix R requirements for full coverage by automatic suppression systems in FPE-S06-004 will be removed during implementation.

Fire Protection Engineering 2 Update the appropriate fire protection program document(s) to provide a requirement that if the plant elects to implement the methodologies in EPRI Report TR1006756, that the methodologies will be implemented in their entirety as they pertain to the fire protection systems or features being evaluated.

Fire Modeling 4 The Hot Gas Layer analysis in Appendix C of the Fire Scenario Report will be updated.

This change will be incorporated in the updated FPRA model in response to RAI PRA-01.

Fire Modeling 6 Element 6 of the FPRA Fire Scenario Report will be updated to include the RAI response details. Element 7 of the FPRA Fire Scenario Report will be updated to clarify that the refinements do not change the conservative assumptions used in the input parameters of the fire models.

Safe Shutdown Analysis 5 The UFSAR will be revised to reflect transition to NFPA 805 in accordance with FAQ 12-0062 and 10 CFR 50.71(e).

Probabilistic Risk Assessment 10 During transition to NFPA 805, the monitoring program will be enhanced to satisfy new requirements.

Probabilistic Risk Assessment 11 Revise Section 5.1.8.4 and Appendix A of the Fire Scenario Report to reflect the discussion in response to RAI PRA-1 1.

Document Control Desk NG-12-0177 Page 3 of 3 RAI Response Number Description Probabilistic Risk Assessment 13 Revise Section 2 and Table 2.2-2 of the Plant Partitioning and Fire Ignition Frequency report to reflect the expanded justification provided in response to RAI PRA-13.

Probabilistic Risk Assessment 16 Table 3.3-1 of the Fire Model Development Report, the HRA, and the basic event probability will be updated to reflect the lack of a cue and that the operator action is not credited in the FPRA. In addition, the basic event probability will be applied in the updated FPRA model in response to RAI PRA 01.

Probabilistic Risk Assessment 33 Revise Section 6 of the Fire Model Development Report to include the details provided in response to RAI PRA-33.

Probabilistic Risk Assessment 57 Revise the Fire Model Development Report to include the table provided in response to RAI PRA-57.

If you have any questions or require additional information, please contact Steve Catron at 319-851-7234.

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

eu C On April 23, 2012 Peter Wells Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC

Attachment:

Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Light Water Reactor Generating Plants cc: NRC Regional Administrator NRC Resident Inspector NRC Project Manager M. Rasmusson (State of Iowa)

Attachment to Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For Fire Protection For Liqht Water Reactor Generatinq Plants 162 pages follow

RAI - Fire Protection Engineering 1 DAEC RAI FP 1 Fire Area PH1 is described as meeting the deterministic requirements of National Fire Protection Association (NFPA) Standard 805, Section 4.2.3.3(b), which requires 20-feet of separation with detection and suppression throughout the area. Engineering Evaluation (EE) FPE-S06-004, Table 6.2, Paragraph 1-6, references an exemption from 10 CFR 50, Appendix R requirements for full coverage by automatic suppression systems, but there is no discussion of this previous exemption request for partial system coverage in license amendment request (LAR) Section 4.2.3 or Attachment K. Clarify the applicability of the exemption request referenced in FPE-S06-004.

RESPONSE

Section 4.2.3 and Attachment K of the enclosure to the License Amendment Request (ML11221A280) (LAR) are correct; DAEC did not intend to carry forward the exemption referenced in FPE-S06-004, Table 6.2, Paragraph 1-6. The basis for acceptability of partial area suppression meeting the requirements of NFPA 805, Section 4.2.3.3(b) is found in the Engineering Evaluation enclosed with supporting document FPLDA013-PR-016 Attachment 2. The reference to the exemption from 10 CFR 50, Appendix R requirements for full coverage by automatic suppression systems in FPE-S06-004, Table 6.2, Paragraph 1-6 will be removed during implementation.

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RAI - Fire Protection Engineering 2 DAEC RAI FP 2 The compliance statement for LAR Table B-i, Element 3.2.3(1) is "complies with clarification." The apparent clarification is to allow modification of surveillance frequencies in accordance with the methodology in Electric Power Research Institute (EPRI) Technical Report (TR) 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features." The proposed future use of the EPRI methodology is not a clarification to current compliance as these methods are not currently incorporated in the surveillance program and there is no associated implementation item in Attachment S of the LAR. Discuss the planned application and incorporation of the EPRI methodologies at Duane Arnold Energy Center (DAEC).

RESPONSE

Attachment A of the enclosure to the License Amendment Request (ML11221A280)

Table B-1, Element 3.2.3(1) compliance is amended as follows:

Compliance Statement:

Complies with Clarification Compliance Basis:

Procedures are established for inspection, testing, and maintenance of fire protection systems as identified in the DAEC Fire Plan.

Clarification: Surveillance frequencies are outlined in the DAEC Fire Plan and may be modified in accordance with the methodology in EPRI Report TR1 006756, Fire Protection Equipment Surveillance Optimization and Maintenance Guide. See Implementation Item in Table S-2 of Attachment S.

References:

DAEC Fire Plan - Volume 1 Rev. 58 [Sections 8.0 and 12.0] - Program Open Item:

Update the appropriate fire protection program document(s) to provide a requirement that if the plant elects to implement the methodologies in EPRI Report TR1 006756, that the methodologies will be implemented in their entirety as they pertain to the fire protection systems or features being evaluated.

In addition, an Implementation Item will be added in Table S-2 referencing this LAR Table B-1 Element with the description of the implementation item being the same as the above open item.

Page 1 of I Rev A. Page 1 of 1

RAI - Fire Protection Engineering 3 DAEC RAI FP 3 LAR, Attachment L, Approval Request 1, references previous evaluations of epoxy floor coatings that were performed in response to an unresolved item from an NRC inspection report. Resolution of the unresolved item included evaluation of the potential for fire propagation across barriers and for those areas crediting spatial separation of redundant safe shutdown equipment. The Approval Request states the combustible contribution of the epoxy does not present a challenge to the plant fire barriers, but does not specifically address the potential for propagation between redundant success paths for achieving the nuclear safety performance criteria. Clarify how epoxy coatings will not propagate fire between spatially separated nuclear safety capability systems or components.

RESPONSE

Epoxy floor coatings will not propagate fire between spatially separated nuclear safety capability systems or components and are not considered an intervening combustible.

NextEra Energy Duane Arnold submitted Approval Request 1 of Attachment L of the enclosure to the License Amendment Request (ML11221A280) (LAR) due to isolated thicknesses in excess of that specified in the definition of limited combustible material.

Isolated thicknesses of epoxy floor coatings in excess of the NFPA 805 definition of limited combustible material would not result in increased flame spread across a floor such that fire would propagate between spatially separated nuclear safety systems or components.

Industry and NRC guidance on epoxy floor coatings has determined that they may be considered non-combustible material as stated in Reference 4: "depending on the (1) the thickness with which the coating is applied, and (2) an independent laboratory testing of the flame spread rating for specific epoxy floor coating." Based on this, epoxy floor coatings are non-combustible material when applied in the appropriate thickness and when the coatings have an acceptable flame spread rating. As such, the coatings are not considered intervening combustibles and would not support flame propagation.

The following supporting information further explains the history of the issue.

NextEra Energy Duane Arnold provided the following statement in Reference 1 (page 2) to substantiate the lack of fire propagation. NextEra Energy Duane Arnold concluded that the epoxy floor coatings in these configurations were acceptable.

The potential for fire spread via floor coatings has been evaluated for the following fire zone configurations:

" those on the same building elevation for which no physical fire barrier separation exists, but which credit spatial separation and no intervening combustibles with providing the necessary fire area separation, and

those on the same building elevation which credit a single door with providing the necessary fire area separation In Reference 2 (page 19), the NRC agreed and closed the unresolved issue.

Rev A. Page I of 2

RAI - Fire Protection Engineering 3 Based on this, epoxy floor coatings at DAEC are non-combustible material. Therefore, they are not intervening combustibles as described in Information Notice 2007-26 and Generic Letter 86-10. NextEra Energy Duane Arnold determined that the isolated excess thicknesses were not in the areas evaluated for separation concerns.

In addition, the following definition has been added to the Fire Hazards Analysis in FHA-400. Floor coating procedures are in place to manage the application of coatings.

Non-combustible material

a. Materialthat will not ignite, burn, support combustion or release flammable vapors when subjected to fire or heat.

b. Materialhaving a structuralbase of noncombustible material, as defined in a.

above, with a surfacing not over 1/16-inch thick that has a flame spreadrating not higher than 50 when measured using ASTM E 84 test "Surface Burning Characteristicsof Building Materials".

References

1. NG-03-0527, Response to NRC Unresolved Item 50-331/03-02(DRS): Epoxy Floor Coatings, dated July 25, 2003

2. NRC Letter, Burgess to VanMiddlesworth dated July 12, 2005

3. NEI Letter, Alex Marion to Sunil Weerakokody dated June 28, 2004

4. NRC Information Notice 2007-26: Combustibility of Epoxy Floor Coatings at Commercial Nuclear Power Plants Page 2 of 2 Rev A.A. Page 2 of 2

RAI - Fire Protection Engineering 4 DAEC RAI FP 4 The safety margin discussions in Approval Requests 1, 2, and 3 in Attachment L of the LAR contain the following statements, respectively:

These precautions and limitations on the use of these materials have been defined by the limitations of the analytical methods used in the development of the Fire Probabilistic Risk Assessment (FPRA). Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.

The use of these materials has been defined by the limitations of the analytical methods used in the development of the FPRA. Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.

The use of these systems has been defined by the limitations of the analytical methods used in the development of the FPRA. Therefore, the inherent safety margin and conservatisms in these methods remain unchanged.

Clarify these statements and describe how safety margins are met for the individual approvals that are requested.

RESPONSE

The safety margin discussions in Approval Requests 1, 2, and 3 in Attachment of the enclosure to the License Amendment Request (ML11221A280) (LAR) are clarified as follows:

Approval Request 1: Epoxy Floor Coating Safety Margin and Defense-in-Depth:

The use of epoxy floor coating does not affect safety margin as it, in general, meets the definition of a limited combustible material with isolated thickness excesses. The floor coating materials were evaluated to have a negligible effect on combustibility.

Application of epoxy floor coatings is controlled via a DAEC procedure. The areas with epoxy floor coatings have been analyzed in their current configuration. The precautions and limitations on the use of these materials do not impact the analysis of the fire event.

Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.

Approval Request 2: Plastic Conduit for Embedded Installations Safety Margin and Defense-in-Depth:

The plastic conduit material is embedded in non-combustible configurations. The material is protected when embedded from mechanical damage and from damage resulting from either an exposure fire or from a fire within the conduit impacting other targets. The areas with plastic conduit have been analyzed in their current Rev A. Page I of 2

RAI - Fire Protection Engineering 4 configuration. The precautions and limitations on the use of these materials do not impact the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.

Approval Request 3: Lack of Separate Water Supply Connections for Fixed Fire Suppression Systems and Fire Hose Stations Safety Margin and Defense-in-Depth:

The configuration of the Control Building Standby Filter Unit deluge systems to the Turbine Building standpipe water supply system and the diesel fire pump and day tank room suppression system to the Pumphouse standpipe water supply system does not affect safety margin. The nuclear safety analysis does not credit these fixed suppression systems and if there were a fire in the area, backup suppression is available via alternative sources. The use of these systems has been defined by the limitations of the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.

Page 2 of 2 A.

Rev A. Page 2 of 2

RAI - Fire Protection Engineering 5 DAEC RAI FP 5 NFPA 805, Paragraph 3.3.8 states that bulk storage of flammable and combustible liquids is not permitted in structures containing systems or components important to nuclear safety. The compliance basis statements in LAR Table B-I, Element 3.3.8 do not address this prohibition on bulk storage in certain areas of the plant, which is independent of the NFPA 30 code citation in Table B-1. Describe how DAEC complies with the first sentence of 3.3.8 of NFPA 805 for storage of bulk flammable and combustible liquids.

RESPONSE

Table B-i, Element 3.3.8 of Attachment A of the enclosure to the License Amendment Request (ML11221A280) (LAR) compliance should be replaced in its entirety with the following:

Compliance Statement:

Complies Compliance Basis:

There is no bulk storage of flammable and/or combustible liquids inside structures containing systems, equipment, or components important to nuclear safety.

DAEC considers bulk storage to be flammable and/or combustible liquid storage in tanks, drums, etc. that is at a staged location and not connected to a system.

Flammable and/or combustible liquid storage vessels that are installed as part of a designed system (e.g., day tanks for diesel generators or fire pumps, turbine lube oil tanks) do not constitute bulk storage and are not considered to be under the requirements of Section 3.3.8 of NFPA 805. This is consistent with industry interpretations and clarifications with the NRC during the pilot plant process that led to the withdrawal of FAQ 06-0023 in 2007 (ML072700552).

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RAI - Fire Protection Engineering 6 DAEC RAI FP 6 Implementation Item No. 14 (Attachment S, Table S-2,) states: "Implement the results of the Radioactive Release Analysis," but is not specific to the actions to be completed.

Provide additional discussion of the actions to be implemented to address the analysis results for radioactive release.

RESPONSE

Attachment S of the enclosure to the License Amendment Request (ML11221A280)

(LAR) Table S-2, Implementation Item 14 states "Implement the results of the Radioactive Release Analysis" and references LAR Section 4.4 and Attachment E.

LAR Section 4.4.2, Results of the Evaluation Process states "The radioactive release review determined the fire protection program will be compliant with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 upon completion of the implementation items identified in Attachment E. See Attachment S for an Implementation Item.

The main strategy for complying with the radioactiverelease requirements is ensuring that all buildings or areas containingradioactivehazards or the potential for an uncontrolled release during a fire have adequate strategiesto minimize the uncontrolled release of radioactivematerialduring fire fighting activities. This includes the revision or creation of documentation such as pre-fire plans, fire brigade training materials, standardoperatingprocedures, and administrative controls."

LAR Attachment E identifies actions to minimize the risk of radioactive release. These actions are identified in the Administrative Controls and Fire Brigade Training sections of the compartment analysis. These actions include enhancements to the area fire plans and modifications to fire brigade training if necessary.

Therefore, Implementation Item 14 will implement the recommendations found in the DAEC report, "DAEC NFPA 805 Radioactive Release Review", which outlines in detail actions necessary for revision or creation of the documentation described in LAR Section 4.4.2 and Attachment E.

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RAI - Fire Protection Engineering 7 DAEC RAI FP 7 LAR Attachment L, Approval Request 3, requests approval for fire water supply system designs that do not provide separate water supply connections for fixed fire suppression systems and fire hose stations provided for manual backup. The approval request states that an alternative source for manual suppression can be provided from yard fire hydrants. Provide additional discussion of the actions necessary to provide backup suppression to the Standby Filter Unit location, including the approximate distance and any elevation change from the nearest fire hydrant(s) and other alternatives available (e.g., local hose stations) in the proximity. Also, with regard to the Pump House system design, clarify if Sprinkler System 21 that protects the 747-ft elevation should be included in the scope of the approval request. Lastly, the approval request identifies the fixed systems served by a common source but does not identify the associated hose stations. Describe both the fixed and manual systems associated with this request.

RESPONSE

Control Room HVAC Area (Standby Filter Unit location):

The two manually actuated carbon (charcoal) bed deluge suppression systems for the Control Building Standby Filter Units (Deluge Systems 21 and 22) are fed from the Turbine Building standpipe system. The Standby Filter Unit carbon bed systems are located in the Control Room HVAC Room (Fire Zone 12B) on the 800' elevation of the Control Building. This room is provided with a sprinkler system (System #12), full smoke detection, and thermal detection for the charcoal beds. Back-up fire suppression to this room includes Sprinkler System #12 (fed from a different header), portable fire extinguishers in the Control Building HVAC Room and the Control Room, manual hose stations in adjacent areas, and hose streams from yard fire hydrants.

There are two hose stations on elevation 786' located within approximately 100' of the access to the Control Building HVAC Room. Each hose station has at least 75' of hose along with 800' of additional hose available on the Fire Brigade Hose Trailer.

1. Hose Station #37 in the Administration Building just outside the north access to the Control Room

2. Hose Station #38 in the Control Room Note that these hose stations are on the same header from Turbine Building as the deluge suppression systems. Drawing BECH-M133 Sheet 3 (Revision 12) quadrant F-5 shows the details of this hose station branch line. Other hose stations on this branch line include those in the Battery Room Corridor (#24), those in the Control Building/Admin Building (#'s 35, 36, 37, 38, & 76), and those in the Technical Support Center (#70).

The nearest alternative independent manual suppression would be from the fire hydrants on the yard main fire loop. Fire Hydrant #6 is located northwest of the north access doors to the Administration Building on elevation 757'. The Administration Rev A. Page 1 of 5

RAI - Fire Protection Engineering 7 Building hallway provides access to the Control Building elevator and stairwells up to elevation 786' and then to the Control Building HVAC Room. The distance via hose is approximately 350' which is well within the available hose on the hose trailer. The yard loop and hydrants are designed in accordance with NFPA 24 as noted in LAR Attachment A, Section 3.5.10 and will provide adequate service to the building locations in question.

Pumphouse:

The original Request #3 omitted Sprinkler System #21. This system should have been included in the request. The Pumphouse fire suppression systems are fed directly from the fire water system piping in the Pumphouse. The fire suppression systems include Sprinkler System #21 protecting the ESW and RHRSW pump areas and Sprinkler System #7 protecting the diesel fire pump room and the day tank room. In addition, there are two hose stations (#46 and #47) inside the Pumphouse from the same water supply as the sprinkler systems. Backup suppression to this area includes portable fire extinguishers and hose streams from yard fire hydrants. Loss of the single header would isolate both of the sprinkler systems and both of the hose stations. Yard fire hydrants would be used as the alternative independent manual suppression source.

Reference Drawings BECH-M133 sheet 2 (quadrant C-2) and BECH-M133 sheet 5 (quadrant B-3). Note that drawing BECH-M133 sheet 2 shows the systems connected to the incoming service water system. This line is normally closed (valve V46-0010, Drawing BECH-M146). Therefore, these systems are aligned to the fire water system.

Approval Request #3 is therefore reworded as follows:

Approval Request 3 NFPA 805 Section 3.5.11 NFPA 805 Section 3.5.11 states:

"Meansshall be provided to isolate portions of the yard fire main loop for maintenance or repairwithout simultaneously shutting off the supply to both fixed fire suppression systems and fire hose stationsprovided for manual backup. Sprinkler systems and manual hose station standpipes shall be connected to the plant fire protection water main so that a single active failure or a crack to the water supply piping to these systems can be isolated so as not to impair both the primary and backup fire suppression systems".

The two manually actuated charcoal bed deluge suppression systems for the Control Building Standby Filter Units (Deluge Systems 21 and 22) are fed from the Turbine Building standpipe system. The Standby Filter Unit systems are located in the Control Room HVAC Room (Fire Zone 12B) on the 800-foot elevation of the Control Building.

This room is provided with an automatic sprinkler system (System 12), full smoke detection in the room, and thermal detection for the charcoal beds. The detection would result in early warning for fire brigade response. The two hose stations that normally Rev A. Page 2 of 5

RAI - Fire Protection Engineering 7 service the Control Room HVAC Room are #37 and #38 which are located outside of the room but within approximately 100 feet. These hose stations are located on the same piping as the suppression systems such that isolating the line prior to the suppression systems would impair both the deluge systems and the hose stations. The fire brigade is trained and equipped with fire hose to connect to a nearby yard fire hydrant and provide fire fighting water through the nearest available access stairwell.

Therefore backup fire suppression is readily available. The backup fire suppression water supply would be a hose connection to the main fire water system via yard fire hydrants. The nearest fire hydrant is #6, which would result in approximately a 350-foot hose run from the hydrant to the Control Room HVAC Room. There is an approximate 45 foot elevation difference from the yard hydrant to the Control Room HVAC Room.

The use of the yard loop would provide sufficient water pressure and flow to reach the Control Room HVAC Room.

In addition, the Pumphouse standpipe and fire suppression (sprinkler) systems are fed directly from the fire water system piping in the Pumphouse. The fire suppression systems include Sprinkler System #21 protecting the ESW and RHRSW pump areas and Sprinkler System #7 protecting the diesel fire pump room and the day tank room.

Therefore the primary fire suppression systems (Sprinkler Systems #21 and 7) and the backup fire suppression system (standpipe Hose Stations #46 and 47) could be affected by isolating the water supply or by a single active failure. Backup fire suppression for these areas is manual suppression by the fire brigade using an alternative water supply.

The alternative water supply would be a hose connection to the main fire water system via yard fire hydrants. There are yard fire hydrants in close proximity to the Pumphouse which would be used in the event of a fire concurrent with an impairment/break of the water supply piping to the sprinkler/standpipe system. Therefore backup fire suppression is readily available.

The basis for the approval request of performance-based method is:

" Backup suppression is readily available via alternative sources.

" The fire brigade is trained and has access to hose lines connected to the unaffected yard fire water system to provide backup fire suppression in the event of loss of suppression system and manual hose station water.

" The Control Room HVAC Room is protected by room smoke detection, charcoal filter bed thermal detection, and an area sprinkler system supplied via an independent water system.

Nuclear Safety and Radiological Release Performance Criteria:

The configuration of the Control Building Standby Filter Unit deluge systems to the Turbine Building standpipe water supply system and the diesel fire pump and day tank room suppression system to the Pumphouse standpipe water supply system does not affect nuclear safety. There are alternative measures available to ensure suppression of a fire if one were to occur. Therefore there is no impact on the nuclear safety performance criteria.

The configuration of the Control Building Standby Filter Unit deluge systems to the Turbine Building standpipe water supply system and the diesel fire pump and day tank Rev A. Page 3 of 5

RAI - Fire Protection Engineering 7 room suppression system to the Pumphouse standpipe water supply system has no impact on the radiological release performance criteria. The radiological release review was performed based on the manual fire suppression activities in areas containing or potentially containing radioactive materials and is not dependent on the suppression system water supplies. The suppression system water supplies do not change the radiological release evaluation, which concluded that potentially contaminated water is contained and smoke monitored. The configuration of water supply systems does not add additional radiological materials to the area or challenge systems boundaries that contain these systems.

Safety Margin and Defense-in-Depth:

The configuration of the Control Building Standby Filter Unit deluge systems to the Turbine Building standpipe water supply system and the diesel fire pump and day tank room suppression system to the Pumphouse standpipe water supply system does not affect safety margin. The nuclear safety analysis does not credit these fixed suppression systems and if there were a fire in the area, backup suppression is available via alternative sources. The use of these systems has been defined by the limitations of the analysis of the fire event. Therefore, the inherent safety margin and conservatisms in these analysis methods remain unchanged.

The three echelons of defense-in-depth are 1) to prevent fires from starting (combustible/hot work controls), 2) rapidly detect, control and extinguish fires that do occur thereby limiting damage (fire detection systems, automatic fire suppression, manual fire suppression, pre-fire plans), and 3) provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed (fire barriers, success path remains free of fire damage, recovery actions). Echelon 2 is maintained by the availability of automatic detection and suppression (sprinkler system) in the Control Room HVAC Room and the availability of alternative fire brigade water sources for manual fire fighting activities for the Control Room HVAC Room and the diesel fire pump and day tank rooms in the Pumphouse.

The water supply configuration does not affect echelons 1 or 3. The water supply configuration does not compromise automatic fire detection functions or post-fire safe shutdown capability. Alternative hose station connections are available as the primary means of suppression in the event of the loss if the primary water supply.

==

Conclusion:==

NRC approval is requested for the configuration of the Control Building Standby Filter Unit deluge systems to the Turbine Building standpipe water supply system and the diesel fire pump and day tank room suppression system to the Pumphouse standpipe water supply system.

The engineering analysis performed determined that the performance-based methods for evaluating an equivalent level of fire protection for the requirements of NFPA 805 Chapter 3!

A. Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release; B. Maintains safety margins; and Rev A. Page 4 of 5

RAI - Fire Protection Engineering 7 C. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).

Page 5 of 5 Rev A.

A. Page 5 of 5

RAI - Fire Protection Engineering 8 DAEC RAI FP 8 LAR Attachment A, Table B-i, Element 3.5.16 contains requirements regarding shared use of fire water supply and storage for nuclear safety functions (i.e., Exceptions 1 and 2). The LAR states that DAEC utilizes both of the exceptions.

a. With regard to Exception 1 on use of fire water supply system to provide a means of make up to the reactor vessel or spent fuel pool, describe the availability of both fire pumps for all fire areas where a fire pump is credited to support the nuclear safety function.

b. With regard to Exception 2, describe how sufficient water volume is provided to supply both the fire water suppression demand as well as the demand of other plant systems when a fire pump is used to supply these systems. If river or well pumps are credited for makeup to the wet pit, describe how these pumps are protected from fire damage.

RESPONSE

Responses as follows:

a. Further review has determined that the alternate injection via the fire protection water system is not a credited nuclear safety function. Neither the Nuclear Safety Capability Analysis (NSCA) nor the Fire PRA model credit alternate injection. Therefore DAEC does not require compliance via Exception No. 1 to Table B-1 Element 3.5.16.

b. The fire water storage (wet pit) is the primary source of fire water and also the common suction to the circulating service water system and the general service water system. The wet pit capacity of 400,000 gallons alone is sufficient to supply the largest sprinkler system demand plus hose stream. The fire pumps are not used to supply any other plant systems for nuclear safety. The wet pit is supplied from the four river water supply pumps taking suction from the Cedar River. The river water pumps are rated for 6,000 gpm each. The river water pumps are separated and the power supplies are divisionalized. Two river water pumps are separated from the other two pumps via a three-hour fire barrier. The NSCA has determined that at least two river water pumps are available in every fire area. In addition, the Technical Specifications do not allow the plant to operate with less than two river water pumps operable. The largest fire protection system demand is 3,115 gpm including a manual hose stream allowances.

Therefore, the availability of two 6,000 gpm river water pumps is more than sufficient to resupply the wet pit as necessary to maintain maximum fire flow water demands.

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RAI - Fire Protection Engineering 8 Table B-I, Element 3.5.16 compliance should be replaced in its entirety with the following:

Compliance Statement:

Complies with Clarification Compliance Basis:

DAEC does not have a dedicated fire protection water supply system. DAEC complies with clarification via Exception 2. The clarification is that the DAEC fire protection water storage for the fire pump suction is via a common supply and is not dedicated. The fire protection water supply system is not used to provide backup to systems credited for nuclear safety.

Exception No. 2: Water supply to the fire pumps is obtained from a 400,000 gallon wet pit in the pumphouse. The wet pit also provides a common suction to the circulating service water system and the general service water system. The wet pit is supplied by gravity drain from the cooling tower basins in the circulating water system. Based on water level, makeup is provided to the wet pit from up to four 6,000 gpm River Water pumps. Two River Water pumps are analyzed as being available in every fire area.

The 400,000 gallon storage capacity of the wet pit, with make-up from at least two River Water pumps, exceeds the capacity required for 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months fire flow of the largest sprinkler demand (sprinkler system #4 at 2,115 gpm) plus 1,000 gpm for hose streams. In the event of an emergency, the circulating service water and general service water systems can be adjusted to ensure sufficient fire protection water is available.

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RAI - Fire Protection 9 Fire Protection RAI 9 LAR Attachment A, Table B-1, Element 3.6.4, states compliance with previous approval.

The compliance basis for this element does not address the provision of this element to provide manual fire suppression in areas containing systems and components needed to perform nuclear safety functions following a safe-shutdown earthquake (SSE).

Describe how DAEC will provide manual fire suppression protection of nuclear safety functions following a SSE.

RESPONSE

Manual fire suppression is addressed in Duane Arnold's Fire Plan Volume I "Program",

Volume II "Fire Brigade Organization", Volume Ill, "Catastrophic Event Plan", the site Emergency Plan, and the site Severe Accident Management Plan. The purpose of these plans is to facilitate a coordinated response to any emergency event including; devastating fires, natural disasters and large scale security events. These plans provide coordinated direction for manual fire suppression using the Fire Plan, the Site Fire Brigade as supplemented through the operation of the Emergency Plan and Severe Accident Management Plan in regard to:

" Plant operation under emergency conditions

Incident Command for coordinating fire fighting Site and Responding Community Fire Brigades

" Fire Fighting Equipment available includes o Normal plant fire protection equipment functioning after an SSE o Fire Brigade and Fire Brigade Equipment available following an SSE o B.5.b diesel driven fire pump and associated equipment if available after an SSE o Equipment from Responding Community Units

" Fire Water supplies include o Normal fire suppression system if available o Fire hydrants, Circ Water Pit and the Cooling Tower basin if available o Suction from the Cedar River The Fire Plans are implemented via Abnormal Operating Procedure 913, "Fire" and the Area Fire Plans which provide the brigade with information in regard to the affect of fire on the plant, ingress, egress, hazards and fire protection equipment.

Page 1 of I A.

Rev A. Page 1 of 1

RAI - Fire Protection Engineering 10 DAEC RAI FP 10 LAR Attachment E, describes the radioactive release transition evaluation and describes using plant filter, off-gas, and drain collection systems to manage airborne and liquid effluents. Clarify the following:

a. For the Low Level Radwaste Processing Facility, the "Smoke and By Products of Combustion - Airborne Effluent Evaluation," states that any signal from a smoke detector will align dampers to purge the facility through particulate filters. Clarify how these filters are capable of functioning with exposure to, or loading associated with, products of combustion.

b. For the Offgas Retention Building, the "Smoke and By Products of Combustion - Airborne Effluent Evaluation," states that smoke can be monitored but does not provide means to mitigate smoke release. Discuss the actions if monitoring detects radioactive release.

c. For the Offgas Retention Building, the "Fire Suppressant Run Off - Liquid Effluent Evaluation," describes collection of runoff in the Floor Drain Collector Tank. What is the volume of this tank and describe the evaluation of this volume relative to expected fire flow?

d. For the Reactor Building, the "Smoke and By Products of Combustion -

Airborne Effluent Evaluation," states that a high radiation alarm on the exhaust ventilation will result in alignment of the exhaust to the standby gas treatment system. Discuss how this system is designed to handle products of combustion.

RESPONSE

The following is the DAEC response to the sub-items identified above:

a. The filters are not explicitly evaluated for function related to exposure to, or loading associated with, products of combustion. Should a filter become clogged/blocked due to smoke particulate, the flow path would be eliminated and air would not be processed. The ventilation system would be stopped and any airborne effluents would therefore be contained by the physical building structure and administrative processes for monitoring and controlling the release of smoke effluent would take over (similar scenario as an area without fixed smoke filtrations/removal equipment). The administrative processes will be documented in fire brigade training and standard operating procedures.

b. Attachment E of the enclosure to the License Amendment Request (ML11221A280) (LAR) Offgas Retention Building "Smoke and By Products of Rev A. Page 1 of 3

RAI - Fire Protection Engineering 10 Combustion - Airborne Effluent Evaluation" (page E-1 0 of the LAR) states that "The exhaust ventilation for the Offgas Retention Building discharges to the torus area of the Reactor Building. From this point smoke or effluent would be exhausted through the Reactor Building exhaust system. This system monitors radioactive release limits to ensure technical specification requirements are not exceeded." The system being referred to for monitoring is the Reactor Building Exhaust system. The Reactor Building "Smoke and By Products of Combustion

- Airborne Effluent Evaluation" (page E-1 3 of the LAR) states upon high radiation alarm, the normal exhaust ventilation will stop and the system is aligned to stand-by gas treatment.

c. The floor drain collector tank has a capacity of 10,000 gallons. There is no specific evaluation for expected fire flow. There are no suppression systems in this area. Based on this, fire flow would be from manual fire fighting activities.

This flow has the potential to vary greatly based on the type and size of the fire.

The fire brigade has/will be trained to be cognizant of radioactive release and to not let effluent flow freely to uncontrolled areas if at all possible.

The Offgas Retention Building collects fire suppression water in the equipment drain sump. The Offgas equipment drain sump then transfers to the floor drain collector tank (UFSAR Figures 11.2-2<1> and 11.2-4/Drawings BECH-M1 37<1>

and BECH-MI 39). The floor drain collector tank is located in the Radwaste Building. The floor drain collector tank is interconnected to various other tanks in the facility in the event of excess capacity. In addition, per UFSAR Section 11.2.2.9; "A major leak in the radwaste system, such as a tank rupture, would result in a dose to an individual at the plant perimeter not exceeding the annual limits of 10 CFR 20." If the mechanism to transfer to the floor drain collector tank is lost, the Offgas Retention Building is bounded by the Reactor Building and the Low-Level Radwaste Processing and Storage Facility with no direct path to the exterior.

d. The Reactor Building exhaust isolates on a high radiation alarm and aligns to the stand-by gas treatment system. The stand-by gas treatment system will filter and monitor the discharge effluent to the plant vent stack. Should the filters become clogged/blocked due to smoke particulate the flow path would be thereby be eliminated and air would not be processed. The ventilation system would be stopped and any airborne effluents would therefore be contained by the physical building structure and administrative processes for monitoring and controlling the release of smoke effluent would take over (similar scenario as an area without Rev A. Page 2 of 3

RAI - Fire Protection Engineering 10 fixed smoke filtrations/removal equipment). The administrative processes will be documented in fire brigade training and standard operating procedures.

The DAEC LAR includes Implementation Item # 14 which will implement the recommendations of the DAEC Report "DAEC NFPA 805 Radioactive Release Review." These recommendations include enhancement/development of the pre-fire plans, the training program, and a standard operating procedure to stress actions and methods to mitigate/prevent release of contaminated materials.

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RAI - Fire Modeling 3 DAEC RAI FMod RAI 3 NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models." Section 4.5.1.2 of the Transition Report states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). Details pertaining to the verification and validation of the fire models that were used are provided in Attachment J. Regarding the verification and validation of fire models:

a. Describe how the empirical equations/correlations in the Generic Fire Modeling Treatments document and supplements were verified, (i.e., how was it ensured that the empirical equations/correlations were coded correctly).

b. Describe the verification and validation of the empirical equations and correlations identified in the supplements to the Generic Fire Modeling Treatments document and provide assurance that these equations/correlations were applied within their appropriate range of applicability.

RAI RESPONSE:

Part a At the time the "Generic Fire Modeling Treatments" report was prepared, a calculation development and review process was used that consisted of a calculation preparer, a calculation review, and a calculation approver. The general responsibilities for each of these elements are as follows:

" The calculation prepare develops and prepares the calculation using appropriate methods.

The calculation reviewer provides a detailed review of the report and supporting calculations, including spreadsheets and fire model input files. The reviewer provides comments to the preparer for resolution. The reviewer has the option of using a Design Review or an Alternate Calculation review method.

Calculation approver provides a reasonableness review of the report and approves the document for release.

The calculation preparation occurred over a two year period ending in late 2007. The review stage was conducted in late 2007. The calculation was approved January 23, 2008 and signed by the developer (Sean Hunt), the reviewer (John Cutonilli), and the approver (Dr. Craig Beyler). The approved document, the signature page, and an affidavit were transmitted to the Document Control Desk at the Nuclear Regulatory Commission in Washington, D. C. on January 23, 2008. A copy of the transmittal letter, the affidavit, and the signature page are provided as part of this RAI response.

In the case of the empirical equations/correlations that form part of the basis of the "Generic Fire Modeling Treatments," a considerable amount of verification was performed during the preparation stage by the preparer. The empirical Rev D. Page 1 of 8

RAI - Fire Modeling 3 equations/correlations were solved using Excel spreadsheets using either direct cell solutions (algebraic manipulation) or Visual Basic macros. All direct cell solutions were validated by the preparer through the use of alternate calculation. For simple equations, this entailed matching spreadsheet solution to the solution obtained using a hand calculator. For more complex solutions, the alternate calculation verification entailed either subdividing the problem into many sub-components and matching the solution using a hand calculator or matching the solution to a verified solution (i.e., the NUREG 1805 [2004] Solid Flame Heat Flux models). The verification of the Visual Basic macros also depended on the type of macro. In situations where the macro is used to perform multiple direct computations, the macro results were verified against the verified spreadsheet solutions that were verified through alternate calculation. In cases where the macro is used to find a root, the root was verified to be a zero by direct substitution into an alternate form of the solved equation.

The empirical equations/correlations were further verified by the reviewer using a Design Review method as indicated in the signature sheet. An independent reviewer was provided access to the draft report and all supporting calculation materials in late 2007. The reviewer conducted a detailed review of the implementation of the equations within the spreadsheets and the reporting of the equation result in the draft report.

Comments and insights were provided to the preparer over the review period and were addressed to the satisfaction of the reviewer. Upon the completion of the review, a revised draft was prepared for review by the approver about December, 2007. The approver provided a higher level reasonableness check of the methods, approach, and the results. Comments and insights that were provided by the approver were addressed to the satisfaction of the reviewer and Rev. 0 of the report was prepared and approved on January 23, 2008.

ATTACHMENT PROVIDED Part b With the exception of the assumed cable propagation rate described in the report titled "Supplemental Generic Fire Modeling Treatments: Hot Gas Layer Tables", all empirical equations and correlations used in the supplements applied at DAEC are identical to those contained in the "Generic Fire Modeling Treatments" report. As such, validation and verification of these correlations will be described first.

The empirical equations and correlations are drawn from a variety of sources and are all documented in various chapters of the Society of Fire Protection Engineers Handbook of Fire Protection Engineeringor peer reviewed journals (i.e., Fire Safety Journal). The empirical models primarily fall into three groups:

Flame height;

" Plume temperatures; and

Heat fluxes (at a target location).

Table J-2 in Attachment J of the enclosure to the License Amendment Request (ML11221A280) (LAR) identifies all empirical models that are used either directly or Rev D. Page 2 of 8

RAI - Fire Modeling 3 indirectly in the "Generic Fire Modeling Treatments" report. The table also identifies the original correlation source documentation and the correlation range in terms of non-dimensional parameters. The table also provides where applicable supplemental validation work that may have been performed on the correlations and provides limits applied in the "Generic Fire Modeling Treatments" report as applicable.

Except for the cable tray Zone of Influence (ZOI) calculation, the flame height calculation is used only as a means of placing a limit on the applicability of the ZOI tables which are based on the plume temperature and thermal radiation heat flux. The flame height calculation for axisymmetric source fires is robust and has considerable pedigree. The original documentation and basis of the flame height correlation cited in Attachment J of the LAR is Heskestad [1981]. Although there are earlier forms of the flame height equation, Heskestad provides a link between the flame height and plume centerline temperature calculation and identifies the range over which the plume equations are applicable. Because the flame height and plume centerline temperature equations are linked, the plume centerline range cited by Heskestad applies to the flame height calculation as well. The plume centerline temperature equations, and thus the flame height correlation, is applicable over the following range [Heskestad, 1981; Heskestad, 1984]:

v

/ (1) where C* is the heat capacity of ambient air (kJ/kg-K [Btu/Ib-°R]), T. is the ambient temperature (K ['R ]), .9 is the acceleration of gravity (m/s 2 [ft/s 2]), p. is the ambient air density (kg/m 3 [lb/ft]), Q is the fire heat release rate (kW [Btu/s), r is the stoichiometric fuel to air mass ratio, D is the fire diameter (m [ft]), and AE4 is the heat of combustion of the fuel (kJ/kg [Btu/Ib]). Application of Equation (1) depends on the fuel as well as a non-dimensional form of the fire heat release rate (fire Froude Number). In practice, the heat of combustion to air fuel ratio for most fuels will fall between 2,900 - 3,200 kJ/kg (1,250 - 1,380 Btu/Ib), and for typical ambient conditions the IM ratio for which the plume equations have validation basis is between 7 - 700 kW215/m (4 - 9 Btu 215/ft)

[Heskestad, 1984]. For fire sizes on the order of 25 kW (24 Btu/s) or greater, this means that the plume centerline equation is valid for heat release rates of 100 kW/m2 (8.81 Btu/s-ft2 ) to well over 3,000 kW/m 2 (264 Btu/s-ft2). For weaker fires (heat release rates less than 100 kW/m2 [8.81 Btu/s-ft2 ], the tendency of the model is clearly to over-predict the temperature and flame height; thus for applications outside the range but below the lower limit the result will be conservative. The concern is therefore entirely on the upper range of the empirical model. The tables in the "Generic Fire Modeling Treatments" are specifically developed with transient, lubricant spill fires, and electrical panel fires with a heat release rate per unit area within the validation range. When the heat release rate per unit area falls outside the applicable range, the table entry is not provided and it is noted that the source heat release rate per unit area is greater than the applicable Rev D. Page 3 of 8

RAI - Fire Modeling 3 range for the correlations. This applies to the flame height and the plume temperature for axisymmetric source fires.

Note that Equation (1) is somewhat different from the equation provided in Table J-2 on page J-6 of the LAR. The equation on page J-6 of the LAR under the column "Original Correlation Range" should be replaced with Equation (1) of this RAI. This change does not affect the results in the "Generic Fire Modeling Treatments" report or any of its supplements because the equations in Table J-2 of the LAR are not part of the models themselves but rather a range over which the model constants have been correlated.

The flame height and plume centerline temperature for line type fires (fires having a large aspect ratio) are applied only to cable tray fires. The correlation used has pedigree and has existed in its general form since at least Yokoi [1960]. Most recently, Yuan et al. [1996] provided a basis for the empirical constant using experimental data with source fires having a width of 0.015 m - 0.05 m (0.05 - 0.15 ft) and a length of 0.2 - 0.5 m (0.7 - 1.5 ft) [Yuan et al., 1996]. When normalized, the applicable height to heat Z

release rate per unit length range (T') for the correlations based on the experiments of Yuan et al. [1996] is between 0.002 and 0.6. This range includes the flame height as well as the elevation at which the temperature is between 204 - 329°C (400 - 625 0 F),

the temperature at which cable targets are considered to be damaged under steady state exposure conditions. Yuan et al. [1996] also provide a tabular comparison of the empirical constant against seven preceding line fire test series, which include a broader range of physical fire sizes and dimensions. The Yuan et al. [1996] constant is greater than the other seven and thus the temperatures and flame heights are more conservatively predicted using the Yuan et al. [1996] data. The application of the Yuan et al. [1996] correlation in the "Generic Fire Modeling Treatments" falls within the normalized applicability range reported by Yuan et al. [1996].

Note that the physical description of the source fire is misreported in Table J-2 of the LAR as 0.15 - 0.5 m (0.5 - 1.5 ft). The "Original correlation range" entry listed in Table J-2 for the "Line fire flame height" (page J-9) and the "Line fire plume centerline temperature" (page 10) should be replaced be 0.015 - 0.05 m (0.05 - 0.15 ft). This change does not affect the results in the "Generic Fire Modeling Treatments" report or any of its supplements because the equations in Table J-2 of the LAR are not part of the models themselves but rather a range over which the model constants have been correlated.

Four flame heat flux models are used in the "Generic Fire Modeling Treatments" as described in Appendix J of the LAR: the Point Source Model, the (simple) Method of Shokri and Beyler, the Method of Mudan and Croce, and the Shokri and Beyler Method.

The former two are simple algebraic models using the heat release rate, separation distance, and the fire diameter. The latter two are considered detailed radiant models that account for the emissivity of the fire and the shape of the flame. Due to limitations in the target placement, the (Simple) Method of Shokri and Beyler are shown to be Rev D. Page 4 of 8

RAI - Fire Modeling 3 inapplicable for calculating the ZOI dimensions. Similarly, for the fuels considered, it is shown that the Method of Mudan and Croce produce a net heat flux that exceeds the fire size. The ZOls are therefore determined using the Point Source Model and the Method of Shokri and Beyler. The method that produces the largest ZOI dimension is used for each fuel and fire size bin.

The Point Source Model and the Method of Shokri and Beyler have been shown in the NUREG 1824 verification and validation study to provide reasonably accurate redictions when the target separation to fire diameter (IF) ratio is between 2.2 and 5.7 [NUREG 1824, Volume 1, 2007]. Furthermore, the fire size ranges considered in the "Generic Fire Modeling Treatments are between about 25 - 12,000 kW (24 - 11,400 Btu/s) and the heat release rates per unit area range between about 100 - 3,000 kW/m 2 (8.1 - 264 Btu/s-ft2 ) for all fuels and fire size bins. Using this information, the following table may be assembled for the applicable target heat flux range, based on the NUREG 1824, Volume 1 [2007] verification and validation range:

NUREG 1824, Volume 1 [2007] "Generic Fire Modeling Treatments" Applicable Heat Flux Range Heat Release Point Source ModeaHetReelerHea Shokri and Fire Size Rate Per Unit Fire Diameter Model Heat Beyler Heat M2Flux Range Flux Range (kW [Btu/s]) Area (kW/m2 (m [ft]) (kW/m 2 [Btu/s- (kW/m 2 [Btu/s-

[Btu/s-ft2 ]) ft2]) ft2 ])

25(24) 100 (8.8) 0.56 (1.9) 0.07 - - 0.45 (0.006 0.04) 0.36 - 3.8 (0.03-0.4) 2- 13.6 2.84- 10 0.1 (0.3) (0.2 - 1.2) (0.3 - 0.9) 25 (24) 3,000 (264) 0.07 - 0.45 0.55 - 5 12,000 (11,400) 100 (8.8) 12.4 (41) (0.006-0.04) (0.05-0.4) 2-13.6 0.45-04.6 12,000 (11,400) 3,000 (264) 2.3 (7.4) (0.2-1.2) (0.04-0.4)

The threshold heat fluxes that define the steady state ZOI dimensions range from 5.7 -

11.4 kW/2 (0.5 - 1 Btu/s-ft2). Transient ZOI dimensions, addressed in the "Supplemental Generic Fire Modeling Treatments: Transient Ignition Source Strength" may approach 16 - 18 kW (1.4 - 1.6 Btu/s-ft2 ). Clearly, the steady state ZOI dimensions based on critical heat fluxes of 5.7 - 11.4 kW/2 (0.5 - 1 Btu/s-ft2 ) overlay with the range of valid predicted heat fluxes identified in NUREG 1824, Volume 1 [2007]. Fuels that identify the most conservative value over a range of heat release rates per unit area (transient and electrical panels) will thus include at least one point within the validation range (i.e., 5.7 kW/m 2 [0.5 Btu/s-ft2]). Since the algorithm searches for the most adverse value, the result will be not less conservative than the value obtained within the model validation and verification range.

Rev D. Page 5 of 8

RAI - Fire Modeling 3 There are combinations of fuels and source strength ranges that do not produce heat fluxes that fall within the validation range, however. This is especially true for the higher target heat flux values (11.4 kW/m 2 [1 Btu/s-ft2 ] and higher) combined with the lower transient fuel package heat release per unit area range (200 - 1,000 kW/m 2 [17.6 - 88.1 Btu/s-ft2 ]). This is addressed through an extended Verification and Validation range of the heat flux models provide by the SFPE [SFPE, 1999]. As noted in Appendix J of the LAR, the SFPE assessed the predictive capabilities of the Point Source Model and the Method of Shokri and Beyler against available pool fire data. The pool diameters ranged from 1 - 80 m (3.3 - 262 ft). The conclusion was that the Point Source Model was conservative, but not necessarily bounding, when the predicted heat flux is less than 5 kW/m 2 (0.44 Btu/s-ft2 ) and the empirical constant (radiant fraction) is 0.21. The method is bounding when a safety factor of two is applied to the predicted heat flux. The application in the "Generic Fire Modeling Treatments" uses an empirical constant (radiant fraction) of 0.35, indicating the application is essentially bounding. Similarly, it was concluded that that Method of Shokri and Beyler is conservative when the predicted heat flux is greater than 5 kW/m 2 (0.44 Btu/s) and the method is bounding when a safety factor of two is applied to the predicted heat flux. The implementation in the "Generic Fire Modeling Treatments" is conservative, though not bounding. Although the SFPE considered fire diameters greater than about 1 m (3.3 ft), smaller diameter pool fires are not optically thick and have a lower emissive power [Babrauskas, 2008].

Thus, the use of the methods for smaller fires is conservative though outside the SFPE validation range.

The use of the heat flux models thus generally directly falls within the NUREG 1824, Volume 1 [2007] verification and validation parameter range; however there are cases where this is not so. However, for larger diameter fires, the SFPE provides comprehensive validation against full scale test data of the methods applied. The application in the "Generic Fire Modeling Treatments" report and the applicable supplements necessarily fall within the validation range or are more conservative because the solution algorithm identifies the most adverse solution among the methods.

Smaller fires may fall outside the validation range of both studies, but such fires have a lower emissive power and are conservatively treated using the methods designed for high emissive power source fires.

A number of other empirical models that appear in the generic fire modeling treatments are applied within the stated range of the models or the data for which the models were developed. For example, the cable heat release rate per unit area model is based on cables that have a small scale heat release rate that ranges between 100 - 1,000 kW/m 2 (8.8 - 88.1 Btu/s-ft2). The solution tables are provided for this range. The unconfined spill fire model (heat release rate reduction factor) is based on observations of pool fires having a diameter between 1 - 10 m (3.3 - 33). The diameter range for which ZOI data is provided is 0.7 - 5 m (2.2 - 17 ft). The lower range value is less of a concern due the reduction in the optical thickness of the fire when the diameter falls below 1 m (3.3 ft). The upper range is maintained in the ZOI solutions. The offset distance for flame extensions outside a burning panel have an upper observational limit of about 1,000 kW (950 Btu/s), though it is applied in a normalized form (extension to panel height ratio). The ratio is applied as determined from the test data.

Rev D. Page 6 of 8

RAI - Fire Modeling 3 The assumed fire propagation rate is the only empirical correlation that is used within the supplemental documentation to the "Generic Fire Modeling Treatments" report applied at DAEC that is not described in the original "Generic Fire Modeling Treatments" report. The fire propagation rate is assumed constant, with a value for thermoplastic cables (0.0009 m/s [0.003 ft/s]) and a value for thermoset cables (0.0003 m/s [0.0009 ft/s]). The spread rate values originated from a simple application of a flame propagation correlation in Appendix R of NRUEG/CR 6850 [2005]. Although the application of the flame propagation correlation is material specific, the values listed in the NUREG/CR 6850 [2005] Appendix R application were found to be generally applicable in test series documented in NUREG/CR 7010, Volume 1 [2010]. The cables tested in the NUREG/CR 7010 [2010] test series included thermoset and thermoplastic cables.

Specific cable materials included neoprene, polyvinyl chloride, cross-linked polyethylene, ethylene propylene rubber, chlorsulfonated polyethylene. In a strict sense, the verification and validation of NUREG/CR 7010 [2010] is applicable to cables and configurations tested. However, the dominant parameter in flame propagation correlation is the ignition temperature of the burning material. This temperature can vary among cables, but conservative treatments assume a minimum value for a particular cable class and so define this as a generic ignition temperature. These temperatures are 2040C (400°F) for thermoplastic cables and 3290C (625°F) for thermoset cables.

Specific exceptions do exist as identified in Appendix R of NUREG/CR 6850 and NUREG/CR 7102 [2011]. The spread rates used within the supplemental documentation to the "Generic Fire Modeling Treatments" do not assert a particular cable type; however, the verification and validation basis for the assumed spread rates are applicable to cables that may be treated generic cables per NUREG/CR 6850

[2005].

ATTACHMENTS:

Referenced Pages of the Generic Fire Modeling Treatments Page 7 of 8 Rev D. D. Page 7 of 8

RAI - Fire Modeling 3

REFERENCES:

Babrauskas, V. (2008), "Heat Release Rates," Section 3-1, The SFPE Handbook of Fire Protection Engineering,4 th Edition, P.J. DiNenno, Editor-in-Chief, National Fire Protection Association, Quincy, MA, 2008.

Heskestad, G., "Peak Gas Velocities and Flame Heights of Buoyancy-Controlled Turbulent Diffusion Flames," Eighteenth Symposium on Combustion, The Combustion Institute, Pittsburg, PA, pp. 951-960, 1981.

Heskestad. G. (1984), "Engineering Relations for Fire Plumes," Fire Safety Journal, 7:25-32, 1984.

NUREG 1804 (2004), "Fire Dynamics Tools (FDTS), lqbal, N. and Salley, M. H.,

NUREG-1805, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., October, 2004.

NUREG 1824, Volume 1 (2007), "Verification & Validation of Selected Fire Models for Nuclear Power Plant Applications, Volume 1: Main Report," Salley, M. H. and Kassawara, R. P., NUREG-1824, Final Report, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Washington, D. C., May, 2007.

NUREG 6850 (2005), "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities Volume 2 Detailed Methodology," Electric Power Research Institute (EPRI) 1008239 Final Report, NUREG/CR 6850, Nuclear Regulatory Commission (NRC),

Rockville, MD, September, 2005.

NUREG/CR 7010 (2010), "Cable Heat Release, Ignition, and Spread in Tray Installations During Fire (CHRISTIFIRE) Volume 1: Horizontal Trays," Draft Report for Comment, McGrattan, K., Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC, October, 2010.

NUREG/CR 7102 (2011), "Kerite Analysis in Thermal Environment of FIRE (KATE-fire)

Test Results," Final Report, Nowlen, S. P. and Brown, J. W., Office of Nuclear Regulatory Research, Nuclear Regulatory Commission, Washington, DC, June, 2011.

SFPE (1999), "Assessing Flame Radiation to External Targets from Pool Fires," SFPE Engineering Guide, Society of Fire Protection Engineers (SFPE), Bethesda, MD, 1999.

Yokoi, S., "Study on the Prevention of Fire Spread Caused by Hot Upward Current,"

Report Number 34, Building Research Institute, Tokyo, Japan, 1960.

Yuan, L. and Cox, F., "An Experimental Study of Some Line Fires," Fire Safety Journal, 27, 1996.

Page 8 of 8 Rev D.D. Page 8 of 8

Suite 817 Baltlmore.MD 21227 A HUGHES ASSOCIATES COMPANY January 23, 2008 United States Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Transmittal of Generic Fire Modeling Treatments, Revision 0 Ladies and Gentlemen:

Kleinsorg Group Risk Services, LLC is providing a copy of Hughes Associates, Inc. engineering report entitled, Generic FireModeling Treatments, Revision 0 dated January 23, 2008 to assist in the NFPA 805 Pilot Observation reviews. The document provided is subject to future updates as the pilot initiative progresses.

Hughes Associates, Inc. requests that the document included as Attachment 2 to this letter be withheld from public disclosure pursuant to Title 10 of the Code of Federal Regulations Section 2.390 as it contains proprietary work product. An affidavit dated January 22, 2008, executed by Philip J. DiNenno, Hughes Associates, Inc., is included as Attachment 1.

If you have any questions regarding this submittal, please contact me at 704.651.5548.

Yours truly, Elizabeth A. Kleinsorg Managing'Partner Kleinsorg Group Risk Services, LLC Attachments:

1. Affidavit

2. Generic Fire Modeling Treatments, Revision 0 Cc: Paul Lain (NRR) w/o Attachments vex 415.644.0544 - ekleinsorg@hoifire.com fx415.276.6891

Hughes Associates, Inc. Affidavit Regarding Withholding Generic Fire Modeling Treatments Revision 0, from Public Disclosure 1 of 4

AFFIDAVIT STATE OF MARYLAND )

)SS.

.CO UNTY O F BALTIMO RE ) ............ .... --............... .... ..

1. My name is Philip J. DiNenno. I am President of HUGHES ASSOCIATES, INC.

and as such I am authorized to execute this Affidavit.

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3. I am familiar with the HUGHES ASSOCIATES, INC information contained in the report Generic Fire Modeling Treatments, Revision 0, dated January 2008, and referred to herein as "Document." Information contained in this Document has been classified by HUGHES ASSOCIATES, INC as proprietary in accordance with the policies established by HUGHES ASSOCIATES, INC for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by HUGHES ASSOCIATES, INC and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from 2 of 4

disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information".

6. The following criteria are customarily applied by HUGHES ASSOCIATES, INC to determine whether information should be classified as proprietary:

1. (a) The information reveals details of HUGHES ASSOCIATES, INC's research and development plans and programs or their results.

2. (b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

3. (c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for HUGHES ASSOCIATES, INC.

4. (d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for HUGHES ASSOCIATES, INC in product optimization or marketability.

5. (e) The information is vital to a competitive advantage held by HUGHES ASSOCIATES, INC, would be helpful to competitors to HUGHES ASSOCIATES, INC, and would likely cause substantial harm to the competitive position of HUGHES ASSOCIATES, INC.

6. The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(d) above.

7. In accordance with HUGHES ASSOCIATES, INC s policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside HUGHES ASSOCIATES, 3 of 4

INC only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. HUGHES ASSOCIATES, INC policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

Signature STATE OF / ,_,,yl), COUNTY OF /ý- ,.- ý I HEREBY CERTIFY, that on this .02c day of ,- , 2008, before me, the subscriber, a Notary Public of the above State and County, personally appeared S, "n 5-Sr 'L, k,//-',4do know to me to be the person whose name is su6scribed to the within Affidavit, and after being swom, he made oath in due form of law that the matters and facts set forth in the foregoing statement are true and correct as therein stated.

AS WITNESS my hand and Notarial Seal.

Commsin cEpir Commission Expires / .,- '/

4 of 4

'insorg

__ .. ... Risk SerVICOS EL.C.

A HUGHES ASSOCIATES COMPANY Generic Fire Modeling Treatments Prepared for:

=aMN Erin Engineering and Research, Inc.

Project Number 1SPH02902.030 Revision:

Preparer: *c***.Name Date Preparer: < 01/23/2008 Sean P. Hunt Reviewer: 01/23/2008 John Cutonilli 01_23_2008 Review Method: E esig Review Alternate Calculation Approved by: - 01/23/2008 Craig L. Beyler Form 3B, Revl Page 1 of 1 Revision 01/05/2007

RAI - Fire Modeling 4 DAEC RAI FM RAI 4 NFPA 805, Section 2.7.3.3, "Limitations of Use," states:

"Acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have been subject to verification and validation. These engineering methods shall only be applied within the scope, limitation, and assumptions prescribed for that method."

Section 4.5.1.2 of the Transition Report states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). Details pertaining to the limitations of use of fire models are provided in Attachment J.

Identify uses of the Generic Fire Modeling Treatments outside the limits of applicability of the method and for those cases explain how the use of the Generic Fire Modeling Treatments approach was justified.

RAI RESPONSE:

Attachment J of the enclosure to the License Amendment Request (ML11221A280)

(LAR) provides a detailed list of the range over which the empirical and algebraic models are applicable. The "Generic Fire Modeling Treatments" report incorporates these limitations directly through the compilation of tables under which the empirical model limits are maintained (see RAI FM-03). For example, the limiting heat release rate per unit area for the plume and flame height equation ranges from about 100 kW/m 2 (8.8 Btu/s-ft2) to over 3,000 kW/m 2 (264 Btu/s-ft2 ). The transient, electrical panel, and combustible liquid ignition source Zone of Influence (ZOI) tables are either tabulated within this heat release rate per unit area range or reflect the maximum value over a heat release rate per unit area range bound by the limiting range. As such, when applying the tabulated results of the "Generic Fire Modeling Treatments", the underlying correlations will be used within their limits of applicability, except as described for RAI FM-03, provided the tables are appropriate for the configuration.

In order to determine whether a table is appropriate, each ignition source group contains a list of limits of applicability in the "Generic Fire Modeling Treatments" report.

These limitations may be in Sections 3.2, 4.2, 5.2, 6.2, and Section 7.2 of the report.

Many of the application limits are implemented automatically through the use of the NUREG/CR 6850 guidance. An example would be the characteristics of the transient fuel package (Section 3.2), which requires a unit heat release rate less than 1,000 kW/m 2 (88.1 Btu/s-ft2) and the characteristic plan dimension ratio (-1). Both constraints are met by the non-wood crib transient ignition sources that were used to develop the NUREG/CR 6850 [2005] peak heat release rate condition probability (Table E-7 of NUREG/CR 6850 [2005]).

Several applications within the DAEC FPRA have been identified in which the tabulated data in the "Generic Fire Modeling Treatments" may be applied outside the specified Page 1 of 6 Rev A.

A. Page 1 of 6

RAI - Fire Modeling 4 limits of applicability as described in Sections 3.2, 4.2, 5.2, 6.2, and Section 7.2. These limits are as follows:

Flame heights should be lower than the ceiling height. This is a general constraint on the use of the transient, cable tray, electrical panel, and combustible liquid ignition source ZOI tables.

" The maximum panel dimension for which the electrical panel ZOI tables are applicable is 0.9 x 0.6 x 2.1 m (3 x 2 x 7 ft) tall.

The room floor plan aspect ratio should not exceed five when using the hot gas layer tables (Section 6.2 of the "Generic Fire Modeling Treatments" report).

Areas in which the first application limit identified above may be exceeded are those that contain electrical equipment ignition sources and have low ceilings, such as the Essential Switchgear Rooms. In these areas, the ZOI above the panels was developed for the FPRA ignition source scenario even though the flame height was greater than the ceiling height. The original basis for this limitation is the shape of the flame emitting surface changes and could cause the ZOI to extend laterally either above or below the panel. Because there is little published data available on the heat flux profiles about such a configuration, it was opted to place a limitation on the application rather than pursue a generic solution in the "Generic Fire Modeling Treatments" report.

Nevertheless, the situation does arise periodically. The expectation is that the current implementation of the ZOI tables conservatively bounds the configuration in which the flames impinge on the ceiling end extend laterally from the point of impingement for the DAEC applications, though more generally there could be extreme applications that are not bound by the horizontal ZOls. Such extreme applications would consist of very low clearances between the fire base and the ceiling (- 0.6 - 0.7 m [2 - 2.3 ft]) such that the flame extension approaches or exceeds the ZOI dimension. The reason the cases at DAEC are believed to be bound by the horizontal ZOls developed under the assumption that there is no flame impingement is that the radiant heat flux from the flame extension is projected downward while at the same time the horizontal component is reduced due to the shorter vertical flame segment. In contrast, the horizontal ZOI dimension without consideration of flame extensions is based on the maximum horizontal heat flux component from the unobstructed flame height.

A simple application of Alpert's ceiling jet correlation can be used to estimate the transition from a conservative ZOI to a non-conservative horizontal ZOI when the flame height exceeds the ceiling height [Alpert, 2008]. The ceiling jet temperature, which is based on a range of fire sizes and ceiling height to flame height ratios that encompass those considered, is given as follows [Alpert, 2008]:

Page 2 of 6 Rev A.A. Page 2 of 6

RAI - Fire Modeling 4 s' ~(1) where AT is the maximum temperature within the ceiling jet (°C) at a distance r from the centerline of the fire (m), Q is the total heat release rate of the fire (kW), and H is the height of the ceiling above the fire base (m). A conservative estimate of the fire centerline is at the edge of the panel, which also lines up with the reference point for the ZOI table. Table 1 summarizes the 9 8 th percentile peak heat release rate fire characteristics for a transient ignition source and a multiple bundle electrical panel containing qualified electrical cables, including the ZOI dimensions used in the FPRA.

The distance at which the ceiling jet temperature equals the critical temperature increase for qualified cables (309'C [556 0 F]) as computed using Equation 1 is also shown in Table 1. The results indicate that the horizontal ZOI dimension is conservative provided the base of the ignition source is located more than 0.6 m (2 ft) from the ceiling for a transient fuel package and 0.7 m (2.3 ft) for an electrical panel fire. There are no instances for which this condition applies in the DAEC FPRA. Thus, though the flame heights for some ignition sources do exceed the application limits specified in the "Generic Fire Modeling Treatments" report, the configurations for which this occurs are bound by the ZOI dimensions assumed when developing the FPRA fire scenarios.

Table 1. 9 8 th Percentile Ignition Source Fire Characteristics.

Minimum Ceiling Height above the fire Horizontal ZOI Base for Ignition Peak Heat Flame Heightt Dimension Sgntio Release Rate used in the which the Source (kW [Btu/s]) (m [) used (m the Horizontal ZOI FPRAt (m [ft]) Dimension is Conservative (m [ft])

Transient 317 (300) 1.7 (2.3) 1.6 (5.2) 0.6 (2)

Multiple Cable 702 (665) 2.65 (8.7) 2.77 (9.1) 0.7 (2.3)

Bundle Panel t~rom the generic Fire Modeling Treatments report.

Areas in which the second application limit identified above may be exceeded are those that contain unusually large electrical panels where the dimensions exceed 0.9 x 0.6 x 2.1 m (3 x 2 x 7 ft) tall. On such panel was identified in each of the Essential Switchgear Rooms, though there are likely instances in other plant areas. The rationale for these applications involved a modification to the way in which the "Generic Fire Modeling Treatments" ZOI data are implemented when developing the FPRA fire scenarios. The ZOI for an electrical panel as developed in the "Generic Fire Modeling Treatments" report divided the overall ZOI into a region above the panel and region Rev A. Page 3 of 6

RAI - Fire Modeling 4 below the panel, each with entirely different exposure mechanisms. This led to a five parameter ZOI: four lateral dimensions corresponding to the narrow and wide panel dimensions above and below the panel top and one vertical dimension above the panel top. To minimize the complexity of implementing the ZOI, the FPRA fire scenarios were developed using the largest lateral ZOI dimension and the vertical ZOI dimension. The largest lateral ZOI dimension for the severe and non-severe panel fires corresponds to the lower lateral dimension adjacent to the wide side. This ZOI dimension is defined under the conservative assumptions that, the fire is located at the panel base the heat flux to the internal panel boundary is 120 kW/m 2 (10.6 Btu/s-ft2 ), the internal fire is adjacent to one boundary, and all energy is directed out the boundary in which the flames are adjacent to. The total energy emitted is also constrained to be less than or equal to the heat release rate of the source fire. The method could be extended to even larger panels; however, it was developed under the assumption that the exposure below the panels would be driven by localized internal effects. There is thus a point at which the treatment of the panel fires under as localized internal heat transfer phenomena becomes overly conservative. This is because the heat losses from other boundaries can no longer be ignored and the potential for a boundary to fully open becomes increasingly likely given the large internal heat release rate and the large plane surface area of the boundary panels. A reasonable upper limit for the localized fire exposure treatment of the internal panel fire would be if the panel boundary were fully open. In this case, the maximum heat transferred across one boundary would be given as follows:

,*.,, - AmB (2) where '"mw, is the maximum heat that can be transferred across a vertical boundary of an electrical panel (kW [Btu/s]), Ab is the area of the boundary (M 2 [ft 2]), and E is the flame emissive power (kW/m 2 [Btu/s-ft2]). Assuming the maximum flame emissive power is 120 kW/m 2 (10.6 Btu/s-ft2 ) based on Beyler [2008] and Muhoz et al. [2004], the maximum heat that could be transferred across a vertical boundary via thermal radiation is about 235 kW (227 Btu/s) if the heat transferred across an open boundary is considered to be an upper limit on the boundary heat losses in any one direction. To link this heat loss to the postulated fire size, the radiant fraction is used, which is reasonably approximated as 0.3 for enclosure fires [McGrattan et al., 2008]. Although specific fuels have been shown to have higher radiant fractions [Tewarson, 2008], such radiant fractions were obtained under oxygen rich environments and are not directly applicable to the configuration considered. Data for full scale open burn fires suggests the radiant fraction would be much lower, on the order of 0.2 [Beyler, 2008; SFPE, 1999]. Dividing the maximum boundary heat loss of 235 kW (223 Btu/s) by the radiant fraction (0.3) results in the largest fire size for which the lateral ZOI dimensions would be conservative, or 783 kW (742 Btu/s). This value exceeds the severe fire heat release rate used to characterize both the multiple bundle (717 kW [680 Btu/s]) and single bundle (211 kW [200 Btu/s]) electrical panels. This result is based on a radiant fraction of 0.3; if a value at the upper end of the often cited range 0.3 - 0.4 is assumed

[McGrattan et al., 2008], the largest fire size for which the lateral ZOI dimensions would be conservative, or 588 kW (557 Btu/s). However, this would be based on all heat Rev A. Page 4 of 6

RAI - Fire Modeling 4 losses being directed toward the target. The internal temperature during a fully developed enclosure fire would be greater than 600'C (1,112'F), which suggest the heat losses from all boundaries, except the open boundary, would be on the order of 110 kW (104 Btu/s). This means that the maximum total energy that could radiate toward the target via thermal radiation would be about 600 kW (253 Btu/s) x 0.4 or 240 kW (227 Btu/s). This is comparable to the maximum boundary heat loss via thermal radiation (235 kW [223 Btu/s]), which indicates the conclusion applies over a wider range of radiant fractions when the additional boundary heat losses are included.

The discussion above does not consider the flame extensions that would result if the boundary were actually open; however, the heat losses in directions other than that in which the target is located are also not considered. In the limit, the fire could be considered to be entirely open, in which case the ZOI dimensions obtained for transient fuel package fires would be nominally applicable. Based on the Table 3-2 in "Generic Fire Modeling Treatments", it is seen that the horizontal dimension of the ZOI for a 717 kW (680 Btu/s) source fire relative to a thermoset cable target would be approximately 2.3 m (7.6 ft). This ZOI dimension is relative to the center of the fire, so the ZOI dimension relative to a panel edge would be about 1.5 m (5 ft) for a 0.9 x 0.6 x 2.1 m (3 x 2 x 7 ft) tall electrical panel under these assumptions. This is bound by the ZOI dimension used in the FPRA, 2.77 m (9.1 ft), by a significant margin. The heat release rate per unit area for this configuration (1,280 kW/m 2 [113 Btu/s-ft2 ]), as obtained by dividing the peak heat release rate by the plan area of the panel (717 kW/0.56 m 2 [680 Btu/s/6 m2]), falls outside the range for which the "Generic Fire Modeling Treatments" report Table 3-2 was compiled. However, it can be seen in Figure 3-10 that the horizontal ZOI component is a decreasing function of the heat release rate per unit area for values above 200 kW/m 2 (17.6 Btu/s-ft2 ), so the estimate is applicable.

Given these considerations, it is clear that the application of the horizontal ZOI dimensions to panels larger than 0.9 x 0.6 x 2.1 m (3 x 2 x 7 ft) tall falls outside the application limit specified in the "Generic Fire Modeling Treatments" report. However, the applications at DAEC remains conservative for panel fire sizes up to about 783 kW (742 Btu/s) based on the way in which the ZOI is implemented. Because there are no situations in the DAEC FPRA for which panel fires larger than 717 kW (680 Btu/s) are postulated, the methods used are conservative.

A single area at DAEC has been identified for which the hot gas layer tables in the "Generic Fire Modeling Treatments" report were applied in an area where the floor area aspect ratio exceeded five (the Battery Room Corridor). A simple means to adapt the geometry to the underlying fire modeling basis as recommended in NUREG 1934 [2011]

will be used to reevaluate the fire scenarios in this area. The geometric adjustment essentially requires the volume of the enclosure to be reduced to the volume of a segment that meets the aspect ratio requirements. The HGL analysis in Appendix C of the Fire Scenario Report will be updated. This change will be incorporated in the updated FPRA model in response to RAI PRA-01.

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RAI - Fire Modeling 4 References Alpert, R. L. (2008), "Ceiling Jet Flows," Section 2-2, The SFPE Handbook of Fire Protection Engineering,4 th Edition, Society of Fire Protection Engineers, Bethesda, MD, 2008.

Beyler, C. L. (2008), "Fire Hazard Calculations for Large, Open Hydrocarbon Pool Fires," Section 3-10, The SFPE Handbook of Fire Protection Engineering,4 th Edition, Society of Fire Protection Engineers, Bethesda, MD, 2008.

NUREG 1934 (2011), "Nuclear Power Plant Fire Modeling Application Guide," Draft for Public Comment, Salley, M. H. and Kassawara, R. P., NUREG-1934/EPRI-1019195, U.S. Nuclear Regulatory Commission, Office of Nuclear Reactor Research, Washington, D. C., 2011.

NUREG/CR 6850 (2005), "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities Volume 2 Detailed Methodology," EPRI 1011989 Final Report, NUREG/CR-6850, Nuclear Regulatory Commission, Rockville, MD, September, 2005.

McGrattan, K. and Miles, S. (2008), "Modeling Enclosure Fires Using Computational Fluid Dynamics (CFD)," Section 3-8, The SFPE Handbook of Fire Protection Engineering,4 th Edition, P. J. DiNenno, Editor-in-Chief, NFPA, Quincy, MA, 2008.

Muhoz, M., Arnaldos, J., Casal, J., and Planas, E. (2004), "Analysis of the Geometric and Radiative Characteristics of Hydrocarbon Pool Fires," Combustion and Flame, No.

139, pp. 263-277, 2004.

Tewarson, A. (2008), "Generation of Heat and Chemical Compounds in Fires," The Society of Fire Protection Engineers (SFPE)Handbook of Fire Protection Engineering, 4 th Edition, Section 3-4, Society of Fire Protection Engineers, Bethesda, MD, 2008.

SFPE (1999), "The SFPE Engineering Guide for Assessing Flame Radiation to External Targets from Pool Fires," NFPA, Quincy, Mass., June, 1999.

Page 6of6 Rev A.A. Page 6 of 6

RAI - Fire Modeling 5 FM Question # 05 NFPA 805, Section 2.7.3.4, "Qualification of Users," states:

"Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations."

Section 4.5.1.2 of the Transition Report states that fire modeling was performed as part of the F PRA development (NFPA 805,Section 4.2.4.2). This requires that qualified fire modeling and PRA personnel work together.

Regarding qualifications of users of engineering analyses and numerical models:

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

b. Describe the process/procedures for ensuring adequate qualification of the engineers/personnel performing the fire analyses and modeling activities.

c. Explain how the necessary communication and exchange of information between fire modeling analysts and PRA personnel was accomplished.

RESPONSE

a) The development of a FPRA involves the integration of a variety of technologies.

These technologies include both fire modeling analyses and probabilistic risk assessment techniques. The qualifications that are required for the staff and consulting engineers that use and apply these technologies depends in part on their specific assigned role on the project. In general, the qualification requirements for those that are technical leads in the preparation of technical tasks are consistent with and often exceed those articulated in NEI 07-12 for qualification of Peer Reviewers. Given the magnitude of the technical activity being performed, the technical leads are sometimes assisted by support staff. There are no specific qualifications for those in a support role as the assigned technical lead would retain overall technical responsibility for the entire body of work. The overall acceptability of the resulting body of work is established through the review and approval process of the associated analysis documentation.

b) The process/procedure used for ensuring adequate qualification of the engineers/personnel performing the fire analysis and modeling activities varied. In the specific case of fire modeling, since the activity involved a skill set that was not inherently part of the capabilities for a risk assessment or a deterministic safe shutdown analyst, acknowledged industry experts were used exclusively for this task. In the case of modeling, a series of risk analyst qualification guides were used. These qualification guides are similar to those already used by the risk assessment group but were Rev A. Page 1 of 2

RAI - Fire Modeling 5 expanded to address the unique aspects associated with fire PRA. The qualification guides include both knowledge and practical elements.

c) The coordination of technical activities between the fire analysis individuals and the risk modeling individuals was facilitated by the availability of a detailed generic fire modeling analysis. Unlike an approach where individualized fire modeling analyses are prepared for each fire initiating event, the generic treatment serves to establish a standardized solution. In order to develop such a standardized solution, the fire modeling analyst established a prescriptive set of boundary conditions for which the generic solution would apply. These boundary conditions are stipulated in a detailed technical report. This detailed technical report forms the foundation of the means for coordinating the technical activities. In addition, the fire modeling analyst and the risk modeling individuals were integrated into a single project team which further facilitated and streamlined the communication and exchange of information.

Page 2 of 2 A.

Rev A. Page 2 of 2

RAI - Fire Modeling 6 DAEC RAI FM 6 NFPA 805, Section 2.7.3.5, "Uncertainty Analysis," states:

"An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met."

The MCR abandonment fire modeling report and the Generic Fire Modeling Treatments document and supplements include a discussion of the sensitivity of the calculations to variations in the input parameter values.

Regarding the uncertainty analysis:

a. The staff requests an explanation regarding if and how the results of these sensitivity analyses were used in satisfying the requirements of NFPA 805 Section 2.7.3.5.

b. In cases where the results of the sensitivity analyses were not used for this purpose (see a. above), explain how the requirements of NFPA Section 2.7.3.5 were met.

c. If necessary, revise elements 6 and 7 in Section 1.3.1 of the FPRA Fire Scenario Report (Report No. 049308001.003, Rev. 2).

RESPONSE

a. The performance criteria shall be met to demonstrate reasonable assurance that in the event of a fire the plant is not placed in an unrecoverable condition. With the exception of the Main Control Room (MCR), the Generic Fire Model Treatments were used for fire scenario development. The Generic Fire Model Treatments Report assesses the model results sensitivity to unknown parameters to ensure that bounding assumptions are used when calculating the critical separation distances for transient fires (Section 3.3.6), small liquid pool fires (Section 4.3.3), electronic cabinets (Section 5.3.3), and cable tray fires (Section 7.3). Appendix B of the Generic Fire Model Treatments describes that the hot gas layer fire model inputs and assesses the results sensitivity and concludes the inputs are expected to produce the most severe fire environment given a heat release rate, enclosure volume, and opening area percent.

Given these sensitivity assessments, the Generic Fire Model Treatments provide conservative input when identifying fire scenarios.

The MCR abandonment time fire model report assesses the results sensitivity to several different input parameters in Appendix B. Section B.13 of the MCR abandonment time fire model report summarizes the sensitivity analysis concluding that most of the input parameters were selected conservatively. Two parameters, the fire growth rate and burning regime, were identified as potentially influencing the predicted MCR abandonment times. However, these parameters were selected based on industry guidance and what is representative of the MCR. Therefore, given the sensitivity assessment, the MCR abandonment time fire model results provide conservative input when identifying fire scenarios.

Rev A. Page I of 2

RAI - Fire Modeling 6 The sensitivity assessments in the Generic Fire Model Treatments and the MCR abandonment fire model report conclude that conservative inputs were selected and result in conservative critical separation distances, time to hot gas layer estimates, and time to MCR abandonment estimates. These conservative results are used as the input for identifying fire scenarios. As such, the requirements of NFPA 805 Section 2.7.3.5 have been met.

b. The sensitivity analysis in the Generic Fire Model Treatments report and the MCR abandonment fire modeling report were used to meet the requirements of NFPA 805 Section 2.7.3.5 as described in part (a) of the response above.

c. Element 6 in Section 1.3.1 of the FPRA Fire Scenario Report states, "Fire modeling is based on the results of the Hughes report, Generic Fire Modeling Treatments which utilize NUREG/CR-6850 HRRs. The report identifies uncertainties in the fire model used. (SR FSS-C3, HLR FSS-D)."

As discussed in first part of the response, the Generic Fire Modeling Treatments assess the sensitivity of the input parameters and concludes the selected input parameters are conservative. The discussion in Element 6 will be updated to include the details of the discussion in first part of the response.

Element 7 in Section 1.3.1 of the FPRA Fire Scenario Report states, "Bounding fire modeling based on the results of the Hughes report, Generic Fire Modeling Treatments is generally applied. The fire modeling is refined to consider point estimate HRRs and fire growth and manual suppression when required. Electric panel fire growth is based on NUREG/CR-6850. The applied HRRs and fire growth are input parameters with uncertainty that could result in conservative or non-conservative fire risk. Manual non-suppression probabilities are based on NEI-04-02 FAQ 08-0050. Manual detection is assumed at t=0 when automatic detection is available and not credited otherwise. (SR FSS-C1, SR FSS-C2, SR FSS-C3)."

The refinements to the generic fire modeling are related to applying a fire growth rate to electronic cabinet fires and manual non suppression probabilities. These refinements use parameters selected based on accepted industry guidance and do not affect the sensitivity assessment in the Generic Fire Modeling Treatments report. The discussion in Element 7 will be updated to clarify that the refinements do not change the conservative assumptions used in the input parameters of the fire models.

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RAI - Safe Shutdown Analysis 3 DAEC RAI SSA 3 Instrument Tubing - The LAR instrument tubing guidance of NEI 00-01 Sections 3.2.1.2, 3.2.1.7, and 3.4.1.8 refers to referenced calculation NSCA-FPLDA013-PR-007 Revision 0 and R98-0001 Revision 3. The alignment basis for tubing failure indicates "Instrument tubing credited for safe shutdown indication and bi-stable actuation that could adversely affect safe shutdown at DAEC was identified and evaluated. Conclusions of the evaluation are incorporated into DAEC fire area compliance assessments".

Provide the instrument tubing failure modes that were considered in the analysis.

Describe what method(s) of plant review were conducted to determine that certain tubing materials would not be affected by fire?

RESPONSE

Heat sensitive piping is assumed to fail resulting in loss of associated instrument function. Heat sensitive piping is identified as that constructed with brazed or soldered joints per the guidance of NEI 00-01 Section 3.2.1.2, Revision 2. Design documents including Field Sketch Drawings (FSKs) provide installation details and specify materials of construction for instrument tubing. The pertinent bills of material for instrument tubing within the scope of the NSCA show the tubing material as ferrous with joints that are not brazed or soldered.

Page 1 of I Rev A. Page 1 of 1

RAI - SSA 4 SSA Question #4 Safe and Stable - LAR Section 4.2.1.2 describes the capability for achieving and maintaining safe and stable conditions and the licensing basis is to achieve and maintain hot shutdown (Mode 3) conditions. The licensee described "following stabilization at hot shutdown, a long term strategy for reactivity control, decay heat removal, and inventory/ pressure control would be determined based on the extent of equipment damage."

Provide a qualitative description of the risk impact(s) of any "long term actions" required to maintain safe and stable conditions including resource availability and the "routine" nature of the actions that may be required to maintain safe and stable (such as water inventory control and diesel fuel management).

RESPONSE

In the hot shutdown state, decay heat removal is accomplished by allowing steam to flow from the reactor pressure vessel (RPV) to either the main condenser or the suppression pool. If the main condenser is used, heat is rejected to the atmosphere via large cooling towers; if the suppression pool is used, heat is rejected to the Cedar River via one or both of the RHR heat exchangers. The first method relies partially on non-safety related equipment, such as circulating water, condensate, and feedwater pumps, while the second method relies solely on safety-related equipment. In the event that off-site power sources are not available, the second method is used since the safety-related equipment is powered by on-site emergency diesel generators.

The RPV inventory function can be met with either high pressure or low pressure injection systems. Viable high pressure systems are Feedwater, HPCI, RCIC, and CRD, while viable low pressure systems are Condensate, LPCI, and Core Spray.

The Cedar River serves as a reliable, constant source of water for supporting decay heat removal and component cooling needs. If the main condenser is used as the primary heat rejection mechanism, river water is pumped from the intake structure to the Circ Water Pit at a rate sufficient to make up for cooling tower water vapor losses. If suppression pool cooling is used as the primary heat rejection mechanism, river water is supplied at a rate equivalent to the flow rate of operating RHR Service Water and Emergency Service Water pumps.

The Condensate Storage Tanks (CSTs) provide a source of water for operation of the HPCI and RCIC systems. If the tanks become depleted, the suction source for both HPCI and RCIC can be transferred to the suppression pool. In this mode, water loss from the primary containment is minimized by directing steam from the RPV to the suppression pool. The CSTs also provide makeup to the hotwell in the event that the main condenser is being used for decay heat removal. Only a small amount of water is needed for this purpose however, since only water lost due to normal system leakage needs to be replaced.

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RAI - SSA 4 Pressure control of the RPV is accomplished by operation of Safety Relief Valves (SRVs).

The SRVs rely on pressurized nitrogen for their operation, which is supplied by the Drywell Pneumatics system. Four large accumulators provide pneumatic pressure for cycling SRVs. Nitrogen for makeup is available on-site and is provided to the Drywell Pneumatics system automatically by the Nitrogen Makeup System. The Nitrogen Makeup System is assumed in the PRA program to contain sufficient nitrogen for successful operation of drywell pneumatic loads for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In the longer term, additional nitrogen may need to be delivered to the site. Ordering and taking delivery of nitrogen is a routine activity at the DAEC. Ample time would be available for achieving this action in the event that supplies run low while operating in the hot shutdown mode.

If one or both of the standby diesel generators (SBDGs) are operated for on-site AC power supply, delivery of additional fuel oil would be needed if the normal seven day supply runs low. Ordering and taking delivery of diesel fuel oil is a routine activity at the DAEC and with a seven day supply already available on site, ample time would be available for achieving this action if needed.

Adequate negative reactivity for the prevention of recriticality while in the hot shutdown mode is provided by inserted control rods. In the unlikely event that one or more control rods fail to insert, the Standby Liquid Control (SBLC) system can be manually operated to maintain the fuel in a sub-critical condition. Additional supplies of boron are kept on site in the unlikely event SBLC fails to inject when called upon.

The risk impact of needing to obtain commodities such as nitrogen, diesel fuel oil, and boron is very low based on the long period of time before the depletion of such items becomes a concern, and on the routine nature of ordering and taking delivery of them.

For fire initiated events where the Emergency Response Organization is activated, additional personnel are available for initiating these tasks, further ensuring they will be reliably accomplished.

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RAI - SSA 5 DAEC RAI SSA RAI 5 Implementation Items - NEI 04-02, "Guidance for Implementing a Risk-informed, Performance Based Fire Protection Program Under 10 CFR 50.48(c)," Section 4.6 indicates that the LAR should contain a "discussion of the changes to Updated Final Safety Analysis Report (UFSAR) necessitated by the license amendment and a statement that the changes will be made in accordance with 10 CFR 50.71 (e)." Figure 4-8 of the LAR indicates that a revised UFSAR will be developed as a post-transition document representing the revised license condition. However, there appears to be no implementation item to update the UFSAR, nor is there a description of the changes that need to be made to the current UFSAR.

Provide this description and indicate its implementation in Attachment S or justify why this is not necessary.

RAI RESPONSE:

The action to update the UFSAR was not included in Attachment S of the enclosure to the License Amendment Request (ML11221A280) (LAR), since the revision of the UFSAR is required by regulation (10 CFR 50.71e) and part of the UFSAR update process. Therefore the need for an implementation item for NFPA 805 is not required.

Also, the specific NEI 04-02 guidance reference is under revision in FAQ 12-0062 UFSAR Content, Revision 0 dated March 13, 2012 (ML120790015). This FAQ does not suggest that the content of the UFSAR be included in the NFPA 805 LAR submittal..

The FAQ contains the following suggested change to the LAR template:

5.4 Revision to the UFSAR After the approval of the LAR, in accordance with 10 CFR 50.71(e), the [ENTER PLANT] UFSAR will be revised. The format and content will be consistent with FAQ 12-0062.

NextEra Energy Duane Arnold intends to use the following format and content, as outlined in the FAQ.

The proposed UFSAR revision would indicate appropriate "general references" documents, but would not "incorporate by reference" those documents that provide a more detailed description and basis for the risk-informed, performance-based fire protection program (based on the definitions of "General References" and "Incorporation by Reference" in NEI 98-03, Revision 1). After the approval of the LAR, in accordance with 10 CFR 50.71(e), the fire protection section(s) of the UFSAR will be revised. The fire protection section will include the following:

0 9.5.1 Fire Protection o Provide general discussion of the Fire Protection Program regulatory requirements.

0 9.5.1.1 Design Basis Summary o Provide a discussion of defense-in-depth Rev C Page I of 2

RAI - SSA 5 o Provide general discussion of the nuclear and radioactive performance criteria o Provide a general discussion of Chapter 2, 3, and 4 NFPA 805.

o Provide a discussion of codes of record

" 9.5.1.2 Systems Description o Required Nuclear Safety Capability Systems, Equipment and cables o Required Fire Protection System and Features o Required SSCs for radioactive release o Power Block Definition and Structures

" 9.5.1.3 Safety Evaluation (Fire Safety Analysis) o Point to and describe fire protection program design basis document(s)

" 9.5.1.4 Fire Protection Program o Point to and describe fire protection program plan document(s) that describe organization, responsibilities, processes/procedures, and qualifications.

UFSAR sections that currently discuss/refer to the fire protection program will be revised to coordinate with the new NFPA 805 Fire Protection Program sections.

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RAI - Safe Shutdown Analysis 8 DAEC RAI SSA 8 LAR Table B-3, Fire Area CB1, under "Required Fire Protection Systems" states that fire brigade response "could be challenging" as the reason for requiring fire detection systems as "Defense-in-Depth". This phrase is used in numerous entries of numerous fire areas.

Clarify the meaning of "could be challenging" and identify the criteria for making this determination.

RESPONSE

Each fire zone or physical analysis unit was evaluated by the site to a set of considerations to judge whether the installed fire detection or suppression system(s) are necessary to meet defense-in-depth in those Fire Areas meeting section 4.2.4.2 of NFPA 805. There are no specific criteria for "could be challenging," it is a judgment made by the site Fire Marshal, Fire Protection Program Owner and NSCA engineer.

The considerations included the following: Class B fire that potentially could spread to multiple elevations, high heat release rate fire, limited or constricted avenues of attack, and fire fighting environments that may be hot, humid or have radiological concerns.

The installed detection or suppression system will provide early warning to plant operators to allow time for the fire brigade to arrive on the scene and effectively manage the challenges to fire fighting and rapidly extinguish the fire.

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RAI - Safe Shutdown Analysis 9 DAEC RAI SSA 9 Recovery Actions - LAR Attachment G, Recovery Actions, (page G-5, Table G-1) identifies for fire area CB1, two recovery actions for the 1A Switchgear Room Air Supply and Exhaust Fans which are to block open doors and dampers and energize fans in the 1A4 switchgear room.

Identify the dampers and describe the process of blocking open fire dampers.

RESPONSE

The referenced recovery actions are part of the steps required to establish alternate switchgear room ventilation. The inclusion of damper operation is a typographical error in the VFDR (SSA-CB1-09) which then carried through into Attachment G of the enclosure to the License Amendment Request (ML11221A280) (LAR.) No dampers require operation to establish alternate switchgear room ventilation.

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RAI - PRA 2 PRA Question # 2 Confirm that FPRA modeling changes, cable selection analyses, and evaluations which were not complete at the time of the June 2010, FPRA peer review (e.g., F&Os 2-9, 2-3, 2-7, 2-8, 4-17, 5-28), were completed for the LAR FPRA model. In addition, confirm that the final results reported in the LAR are based on a FPRA that is built upon the internal events models that was reviewed in March 2011 including the resolution of the internal events F&Os reported in Appendix U. Insofar as not clarified in the PRA timeline requested in another RAI, provide a summary of how the results of the 2011 internal events review were incorporated into the FPRA that was reviewed in 2010.

RESPONSE

A full scope peer review of the Fire PRA was conducted in June of 2010. Some of the F&Os from this review identified weaknesses in the incorporation of fire related elements into the internal events PRA. These weaknesses were addressed in the Fall of 2010 and in February 2011, a revised PRA model was made available to the NFPA 805 project team. All fire scenarios were re-quantified in FRANC using this revised model; delta CDF and LERF for Variances From Deterministic Requirements (VFDRs) were evaluated using it; and, the XINITS model used for final quantification was derived from the revised model and re-quantified FRANC results.

A focused peer review of the internal events PRA was conducted in March 2011 based on the model and documentation that had been finalized in February. Twelve findings were identified, which are listed in Table U-1 of the enclosure to the License Amendment Request (ML11221A280) (LAR). Each finding is assessed with respect to its potential impact on quantified results of the Fire PRA. All twelve findings were found to have little or no impact on the Fire PRA results.

Findings from the June 2010 Fire PRA peer review were resolved for the LAR FPRA model and their dispositions with respect to the NFPA 805 application are summarized in Table V-3 of the LAR. Included in Table V-3 are the F&Os cited in Question 2 with the exception of F&O 5-28, which is a suggestion F&O rather than a finding.

F&Os 2-7, 2-8, 2-9, and 4-17 involve a lack of evidence that systems were added to the internal events PRA in a manner consistent with the standard's High Level Requirements (HLRs) for Systems Analysis (SA). Table E-1 of the Fire Model Development Report (493080001.02) provides a list of additions to the PRA model in consideration of fire initiated events. Nineteen of the forty-six items are added to incorporate Multiple Spurious Operations (MSOs) identified in the DAEC report of the Expert Panel for Addressing MSOs, while the remaining 27 items are added to incorporate components contained in the Nuclear Safety Capability Analysis (NSCA) that were not previously in the PRA model. Fault tree modeling of most newly added components is simplified since the contribution to plant risk of random failures is insignificant compared to the contribution from the effects of fire. This treatment is consistent with supporting requirement SY-A15, which states, in part, that one or more failure modes for a component may be excluded from the systems model if the Rev A. Page 1 of 2

RAI - PRA 2 contribution of one of them to the total failure rate or probability is less than 1% of the total failure rate or probability for that component.

F&Os 5-28 and 2-3 are related to the cable selection process. Suggestion F&O 5-28 notes that draft fire area Compliance Assessment Summaries were used in the cable selection analysis for NFPA 805; considering they could be out of date, the Fire PRA should reflect the final versions. The unresolved items in the draft summaries were related to items that were resolved during the VFDR identification process and did not have limitations regarding cable selection or circuit analysis. The cable selection and circuit analysis in the FPRA reflects the final version of each Compliance Assessment Summary. Table 4.0-1 of the Fire Scenario Report and the fire scenario database will be updated to properly reference the final Compliance Assessment Reports.

Finding F&O 2-3 identifies a lack of evidence that conductor-to-ground and conductor-to-conductor shorts are included as potential cable and circuit failure modes as required by supporting requirement CS-A5. Section 4 of the Fire Model Development Report documents the FHA-500 methodology used for the cable selection task. The methodology selects all cables for a component which includes cable conductor-to-ground and conductor-to-conductor shorts.

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RAI - PRA 3 DAEC RAI PRA 3 Provide a comprehensive timeline showing the development of the full power internal events (FPIE) PRA and the FPRA for the NFPA 805 application including the PRA model versions and reviews.

RESPONSE

A full scope peer review of the FPIE PRA was conducted in December 2007. The model of record at that time was Revision 5B. In June 2008, a project was started to update the FPIE PRA to Revision 6 with a goal of bringing it into conformance with NRC Regulatory Guide 1.200, Rev 2. At the same time, the Fire PRA project was begun for the NFPA 805 application. Ideally, the FPIE PRA model should be completely finished before being supplied to the NPFA 805 project as a base model to which fire specific elements are added. This was not practical however, given project constraints for completing the Fire PRA prior to the originally scheduled NFPA 805 LAR submittal date.

In July 2009, a preliminary version of the FPIE PRA model was delivered to the NFPA 805 project while work continued to complete it. Plant equipment identified as being needed to support the NFPA 805 Nuclear Safety Capability Analysis (NSCA) were added to the PRA model by the FPIE project personnel while logic associated with fire initiated scenarios were added by the FPRA project personnel. The model additions were tracked with the use of a spreadsheet and are listed in Appendix E of the Fire PRA Model Development Report.

A full scope peer review of the Fire PRA was conducted in June of 2010. Some of the F&Os from this review identified weaknesses in the incorporation of fire related elements into the internal events PRA. These weaknesses were addressed in the Fall of 2010 and in February 2011, a revised PRA model was made available to the NFPA 805 project team. This model became the basis for the Fire PRA. The June 2010 peer review is considered to be the peer review of record for the Fire PRA. Findings from the review, and their dispositions, are contained in Table V-3 of the LAR.

A focused peer review of the FPIE peer review was conducted in March 2011. Its purpose was to review the technical elements of the FPIE PRA with a focus on DAEC's disposition of F&Os resulting from the full scope peer review conducted in 2007. All supporting requirements for the Human Reliability Analysis technical element were reviewed since it involved a broad based change in methodology for calculating human error probabilities. Other upgrades received enhanced attention by the review team, but did not require re-review of all supporting requirements within the respective technical element category.

The March 2011 peer review is considered to be the peer review of record for the FPIE PRA.

Four supporting requirements were assessed as 'Not Met' and three were assessed as only meeting capability category I. Twelve findings were identified, which are listed in Table U-1 of the LAR. Each finding was assessed with respect to its potential impact on Rev A. Page 1 of 2

RAI - PRA 3 quantified results of the Fire PRA. All twelve findings were found to have little or no impact on the Fire PRA results.

The FPIE PRA model that was the subject of the focused peer review in March 2011 is the same model that serves as the bases of the Fire PRA used in performing PRA evaluations for the LAR. Findings associated with the FPIE PRA model, and their impact on fire evaluation results reported in the LAR, are summarized in Table U-I of the LAR. Likewise, findings from the Fire PRA peer review held in June 2010 and their disposition with respect to the NFPA 805 application are summarized in Table V-3 of the LAR. Together, these represent a comprehensive review of all aspects of the PRA that are relevant to NFPA 805 risk reviews.

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RAI - PRA 4 DAEC RAI PRA 4 F&O 4-34. Describe the process for dividing up the transient frequency within physical analysis units (PAUs). Discuss the use of the area weighting factor in evaluating the importance of transients to a PAU. Define localized transients and describe how the area weighting factors were applied to both generalized and localized transients. Clarify if the area weighting factor was applied to incorporate the entire length of cable trays and any other targets not limited to a single location and, if not, provide justification for not doing so. Also clarify the use of 100 square feet as an area factor and its application to only localized transients.

RESPONSE

Walkdowns were performed in PAUs to identify targets that potentially could be damaged by transient fires. Transients were postulated in plant locations where a specific target set was located (e.g., a group of risers). The PAU transient frequency was divided up for each postulated transient and target set based on the transient zone of influence. The zone of influence area for the postulated transient was divided by the overall PAU area to calculate an area weighting factor that was applied to the PAU transient frequency resulting in a frequency for the postulated transient.

In many cases, the postulated transient target set was limited to a group of risers or conduits without secondary combustibles. These transients were referred to as localized transients. Given the small target set a zone of influence area of 100 square feet was generally used. For these types of postulated transients, the use of 100 square feet is conservative given that 100 square feet exceeds the zone of influence area of the 9 8 th percentile transient fire.

In the cases where a postulated transient included a larger target set and/or secondary combustibles (e.g., cable trays) a more appropriate area weighting factor was applied based on the entire area of the postulated transients that could damage the target set (e.g., the entire length of a cable tray).

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RAI - PRA 9 DAEC RAI PRA 9 F&O 3-8 stated that an assessment of the effectiveness, reliability, and availability of credited passive fire barriers was not performed. The response states that the, "[f]ire protection program provides the assessment." Clarify how the fire protection program assess the effectiveness, reliability, and availability of credited passive fire barriers.

RESPONSE

The Fire PRA development made the following assumptions and considerations:

Fire barriers were credited in the multi-compartment analysis (MCA) consistent with the Appendix R requirements in the FHA-400.

The MCA review was completed using the FHA. The boundary is described along with any adjoining compartments in Appendix C of the DAEC Fire Scenario Report.

The Fire Protection Program defines the adequacy of passive fire barriers in the Fire Hazards Analysis (FHA-400). The information in FHA-400 includes:

" FBIM - Fire Barrier Identification Matrix (Table 3-1)

" Fire Zone and Fire Area Selection

Fire Hazards Identification

Fire Zone Occupancy Classification

" Adequacy of Fire Barriers Fire Zone designations are from the original plant partitioning in analyzing the Fire Protection Program under BTP 9.5-1. Fire Areas, which may include multiple Fire Zones, were created to facilitate analysis of fire protection under Appendix R. The FBIM is a listing of fire barriers in relation to their associated fire zones and fire areas. The FBIM identifies which barriers are required to meet Appendix R and those with requirements associated with NEIL insurance requirements as well as other requirements associated with requirements such as building codes.

The Fire Hazards Analysis for each Fire Zone includes an inventory of fire hazards and combustible materials. The analysis of the hazards in FHA-400 in conjunction with a Fire Protection Evaluation (FPE) determines whether or not the barrier is adequate for the hazard. The following considerations are described in detail in FHA-400:

Calculation Methods o Combustible Content of installed Combustibles o Burn Rate of Surface Limited Combustibles o Evaluation of Hydrogen Hazard o Severity of Oxygen Limited Fires o Cable Tray Loading and Heat Release o Occupancy Classification o Fire Duration o Transient Combustible Estimating Rev A. Page 1 of 2

RAI - PRA 9 Adequacy of Barriers o Regulatory basis o Fire Area and Fire Zone Boundaries o Barrier Evaluations o FBIM Designations o Methodology for Evaluating Barriers The FPE then is the final step in the analysis of a barrier where the barrier wall and penetrating elements (HVAC ducts and dampers, pipes, doors and penetrating elements with seals) are assessed. The FPE's are listed in the FBIM. A summary of each FPE is included in LAR Attachment C.

The Fire Plan includes surveillance requirements for fire doors, fire dampers and fire barriers. Each is inspected on a periodic basis and maintained. When fire doors, fire dampers and fire barriers are found to be non-functional, fire protection impairments are initiated, compensatory measures are established and the impairment is tracked until resolved. The Fire Protection Program has a health report that is updated quarterly and presented to Plant Health committee as needed. The health reporting process includes trending evaluation to identify ineffective maintenance. The Fire Marshal conducts periodic inspections to observe and verify plant conditions. These observations are captured in the plant corrective action process and tracked to conclusion when changes or issues are found.

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RAI - PRA 10 DAEC RAI PRA 10 F&O 4-35. SR FSS-D7, regarding credited fire suppression systems, has three elements that must be addressed for CC-Il. Relative to these elements, address the following:

a. Confirm that 1) the credited systems are installed and maintained in accordance with applicable codes and standards and 2) the credited systems are in a fully operable state during plant operation or provide justification why such confirmation is not necessary.

b. Provide a discussion of the basis for your statement that "DAEC have not experienced outlier behavior in the past."

RESPONSE

Duane Arnold has maintained that the Fire PRA satisfied CC I. The following response was written in regard to resolving F&O 4-35:

FSS-D7, CC I While plant specific data was not reviewed, it is not believed that DAEC systems have experienced outlier behavior and generic values provided by NUREG/CR-6850 are appropriate.

Credited suppression systems will be evaluated as part of the NFPA 805 Monitoring Program as described in Section 4.6. Within this program, reliability and availability performance criteria will be established for equipment and programmatic elements important to the fire protection program. Attributes of existing suppression systems with regard to installation, maintenance, and operational history will be determined during development and implementation of the monitoring program.

a) The design requirements for each of the credited suppression systems called out conformance to NFPA standards. The following table lists the systems and identifies specific documentation of recent code evaluations if applicable. Each of the systems is in service and is monitored for impairments under the site common procedure, ACP 1412.4. Periodic surveillance procedures as noted are performed to ensure corrective maintenance is completed to maintain the systems in a fully operational state.

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RAI - PRA 10 Fire Description Code Evaluation Surveillance Zone Procedure 07A Feed Pump Deluge Systems NS130005 3 &4 FPE-S07-003 07C Turbine Lube Oil Cooler NS130005 Deluge System 7 FPE-S07-004 (Note 2) 08D Hydrogen Seal Oil Deluge System 6 (Note 1) NS130005 07C Turbine Lube Oil Tank Wet Pipe Sprinkler System 1 (Note 2) NS130003 Heater Bay Wet Sprinkler System 4 and Condenser DCP 1510, DCP 1549,Cal- NS13C003 07F Bay Wet Pipe Sprinkler 465-M-006 and M104A-10.

System 16 Turbine Building RR Bay Wet 08D Pipe Sprinkler System 9 FPE-S07-002 (Note 1) NS13C003 "B" EDG Room and Day 08F/G Tank Room Pre-action FPE-S06-006 (Note 3) NS13C011 Sprinkler 3 "A" EDG Room and Day 08H/J Tank Room Pre-action FPE-S06-006 (Note 3) NS13C011 Sprinkler 2 NOTES -

1. The Deluge on the Hydrogen Seal Oil System does not have a specific code evaluation. The deluge covers just the area of the Hydrogen Seal Oil Equipment which is underneath the Fire Zone 08D Wet Pipe Sprinkler 9 which covers the entire Fire Zone, including the Hydrogen Seal Oil Equipment skid. Fire Zone 08D Wet Pipe Sprinkler 9 has a specific code evaluation as noted above.

2. The Deluge System on the Turbine Lube Oil Tank is evaluated but the Sprinkler over the entire Turbine Lube Oil System and Reservoir has not been. The Sprinkler was designed in accordance with NFPA 805 standards.

3. The pre-action system for each EDG Room and its associated Day Tank Room is a system protecting both fire zones. Only the protection for the EDG Rooms, 08F and 08H are modeled in the Fire PRA.

b) Surveillance tests are performed on all systems as noted. When corrective maintenance is identified appropriate compensatory measures are taken and maintenance is scheduled. An examination of the impairment log, which has been electronic since 2008, reveals that the vast majority of impairments on these systems are not for corrective maintenance but for other work such as scaffolding Rev A. Page 2 of 3

RAI - PRA 10 installations blocking small portions of the systems spray pattern. Duane Arnold has established System and Program Health reports to trend and monitor equipment and system performance. During the transition to NFPA 805 the monitoring program will be enhanced to satisfy new requirements.

Therefore, while no specific analysis exists to analyze the reliability and unavailability of the equipment there are administrative controls in place that capture corrective maintenance issues and monitoring programs to trend equipment and system performance to ensure operational status.

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RAI - PRA 11 PRA Question # 11 F&O 2-14. For postulated oil fires in the Turbine Building that result in damage to structural integrity address the following:

a. Section 5.1.8.4 and Appendix A of the Fire Scenario Report (FSR) are inconsistent in how the Turbine Building catastrophic fire analysis (i.e., scenario TBO01) is described.

Clarify the assumptions and results for this analysis.

b. This evaluation to satisfy SR FSS-F1 means that SR FSS-F3 is now applicable.

Accordingly, update the self-rating for this SR in LAR Table V-I, "DAEC Fire PRA Quality Summary," or provide justification for why SR FSS-F1 and F3 are not applicable to DAEC. If the self-rating is CC-I, provide justification for why CC-I is adequate for the NFPA 805 LAR.

RESPONSE

a. Fire scenario TBO01 postulates a catastrophic turbine generator oil fire that results in wide spread damage inside the building and damage to the structural integrity of the building. The fire scenario was modeled using the guidance in NUREG/CR-6850, Appendix 0. Table 0-2 provides three conditional probabilities for catastrophic fires; 0.025, 5E-4, and 1E-5.

From Section 0.2.4, the conditional probability of a catastrophic fire is 0.025. Per Section 0.2.4, last sentence of the first paragraph, automatic fire suppression systems and fire brigade can be credited to prevent wide spread damage. Table 0-2 provides two additional conditional probabilities, 5E-4 and 1E-5. Section 0.2.4 does not provide specific clarification for the basis of the two additional conditional probabilities.

Therefore, when applying credit for automatic suppression systems and fire brigade the conditional probability of a catastrophic fire and suppression failure should be estimated between 5E-4 and 1E-5.

Based on the guidance, the following assumptions were used for the analysis:

" The conditional probability of a catastrophic fire is 0.025

The turbine building areas have two automatic suppression systems. Deluge systems protect equipment where a large amount of oil is present, and automatic wet pipe systems protect large areas in the turbine building. From NUREG/CR-6850 Appendix P, the unreliability value for deluge/preaction sprinkler systems is 0.05 and the value for wet pipe sprinkler systems is 0.02. Both of these systems rely on the fire water system, so there is a dependency between the systems that must be accounted for. Given the difficulties associated with assigning an appropriate common cause factor, only the wet pipe system unreliability value will be applied.

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RAI - PRA 11

Fire brigade response was credited based on the average T/G fires duration from Table P-2. Table P-2 identifies 21 events used in the suppression curve for a total duration of 749 minutes. The average calculates to be 749/21 which equals 35 minutes. From Table P-3, the failure to suppress probability for T/G fires at 35 minutes is 0.37. A value of 0.4 will be used.

The non suppression probability used for the automatic suppression systems and fire brigade is calculated to be the product of 0.02, and 0.4 which equals in 0.008.

Using the assumptions above based on guidance from NUREG/CR-6850, the conditional probability of a catastrophic fire of 0.025 was applied and a non suppression probability for automatic suppression systems and fire brigade of 0.008 was applied.

These probabilities result in a probability of a catastrophic fire and failure to suppress of 2E-4. This value is between the recommended values of 5E-4 and 1E-5 in Table 0-2 of NUREG/CR-6850.

Section 5.1.8.4 and Appendix A of the Fire Scenario Report will be updated to reflect the discussion above.

b. Fire scenario TBOO1 was postulated consistent with the requirement of SR FSS-F1 CC-I/Il and included and quantified consistent with the requirement of FSS-F3 CC-Il/Ill. The self rating in Table V-1 of the LAR should be changed as follows:

Table V-1 DAEC Fire PRA Quality Summary Supporting Peer Review DEAC Final Capability Requirement Capability Assessment Comment Assessment FSS-F3 NA CC Il/111 Addressed finding 2-14 Page 2 of 2 Rev A.A. Page 2 of 2

RAI - PRA 13 DAEC RAI PRA 13

a. F&O 6-3. Table 2.2-2 of the "Plant Partitioning and Fire Ignition Frequency Development" report identifies credited spatial separation areas and provides justification based on hot gas layer (HGL) formation and postulated combustibles. Expand the justification to address all of the criteria in Section 1.5.2 of NUREG/CR-6850, including the potential presence of ignition sources and the potential for damage or fire spread from flame height and plume effects.

b. F&O 6-4. Section 2.2 of the "Plant Partitioning and Fire Ignition Frequency Development" report states that non-rated walls, ceilings, and floors were credited for several PAUs. Describe the evaluation used to justify the use of partitioning elements lacking fire resentence rating.

RESPONSE

a) Section 1.5.2 of NUREG/CR-6850 guidance defines a physical analysis unit as a well defined volume where the credited partitioning features should substantially contain fire behaviors. The guidance acknowledges that smoke from a fire and the fire itself may spread beyond a physical analysis unit which is treated in the multi compartment analysis; however, the partitioning features should assure damage is highly unlikely. General guidance for spatial separation indicates that it should be credited in large volumes where the separation is free of ignition sources and combustibles. For vertical separation, the potential for fire spread due to flame spread and plume effects should be considered. SR PP-B3 of the ASME/ANS RA-Sa-2009 PRA Standard states, "If spatial separation is credited as a partitioning feature, JUSTIFY the judgment that spatial separation is sufficient to substantially contain the damaging effects of any fire that might be postulated in each of the fire compartments created as a result of crediting this feature." Spatial separation is credited in the Reactor Building, Turbine Building, Pumphouse, and Intake Structure.

The justification for the credited spatial separation is expanded to include each item from the guidance of NUREG/CR-6850. In the Reactor Building, only transient ignition sources may be postulated in some of the credited horizontal spatial separation elements. However, these areas are large and the only potential path for fire spread to the adjacent area is via cable trays located several feet outside the zone of influence of a 9 8 th percentile transient fire. The credited vertical separation consists of open hatches and do not contain ignition sources or combustibles. In the Turbine Building, only transient ignition sources may be postulated in some of the credited spatial separation elements. However, these areas are large and the only potential path for fire spread to the adjacent area is via cable trays located several feet outside the zone of influence of a 9 8 th percentile transient fire. The credited vertical separation consists of open Rev A. Page 1 of 11

RAI - PRA 13 hatches and grating that do not contain ignition sources or combustibles.

The exception in the turbine building is the potential for oil fires that may not be contained via the spatial separation. These fires are unique and are considered separately in the multi compartment analysis using the guidance in NUREG/CR-6850. In the Pumphouse, only transient ignition sources may be postulated in the credited spatial separation elements.

The separation is via stairways that do not contain combustibles. There is not potential for flame spread and plume effects would be negligible given the volumes and location of equipment and cables. In the Intake Structure, only transient ignition sources may be postulated in the credited spatial separation elements. No combustibles are present in the credited spatial separation elements. The separation is via ventilation openings.

The following table summarizes the spatial separation credited in the FPRA. The table identifies the PAU for which spatial separation was credited. In addition the table identifies if there is a potential for an ignition source and if there are combustibles in the credited spatial separation. For vertical separation, the potential for flame spread and plume effects are identified. Based on these considerations, the spatial separation was reviewed to determine if the spatial separation was sufficient to likely contain a fire. The potential for fire spread beyond the credited spatial separation was included in the multi compartment analysis. Section 2 and Table 2.2-2 of the Plant Partitioning and Fire Ignition Frequency report will be updated to reflect the expanded justification provided in this RAI response.

b) Under the original Duane Arnold Fire Protection Program the primary buildings; Intake, Pump House, Reactor Building and Turbine Building were sectioned by rooms which were designated as Fire Zones.

Consequently the walls, ceilings and floors were originally analyzed to be adequate for the hazard and included in the Fire Protection Program under BTP 9.5-1 Appendix A as qualified fire barriers. Over the years the program was revised and re-analyzed under Appendix R and the barriers re-base lined. These fire barriers were removed from the required Appendix R program to minimize barrier inspections. These fire barriers are inspected and have been maintained and repaired although they are not credited as part of the Appendix R Fire Protection ProgramTherefore, there is reasonable assurance that these fire barriers will prevent the spread of fires and the potential failure of these barriers are considered in the multi-compartment analysis.

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RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 01AN Horizontal Transients only No combustibles No Large Torus Room area where FHA-M-01 separation to 1AS. HGL is not a concern. Postulated Vertical separation transients would not result in fire to 02D via open spread to 1AS or 02D.

hatches.

01AS Horizontal Transients only No combustibles No Large Torus Room area where FHA-M-01 separation to 1AN. HGL is not a concern. Postulated transients would not result in fire spread to 1AN.

02A Horizontal Transients only Exposed cable N/A - horizontal Large general Reactor Building FHA-M-02 separation to 02B. trays located 20 separation only area where HGL is not a concern.

feet above the Postulated transients would not floor. result in fire spread to 02B.

02B Horizontal Transients only Exposed cable No Large general Reactor Building FHA-M-02 separation to 02A. trays located 20 area where HGL is not a concern.

Vertical separation feet above the Postulated transients would not to 03B via an open floor. result in fire spread to 02A or 03B.

hatch.

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RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 02D Vertical separation Transients only No Combustibles No RHR valve room where HGL is FHA-M-01 down to 01AN and not a concern. Postulated up to 02L. transients would not result in fire spread to 01AN or 02L.

02L Vertical separation Transients only No Combustibles No RHR valve room pipe chase FHA-M-via opening of the where HGL is not a concern. 03,04 pipe chase to 02D. Postulated transients would not result in fire spread to 02D.

03A Horizontal Transients only Exposed cable N/A - horizontal Large general Reactor Building FHA-M-03 separation to 03B. trays located 13 separation only area where HGL is not a concern.

feet above the Postulated transients would not floor. result in fire spread to 03B.

03B Horizontal Transients only Exposed cable No Large general Reactor Building FHA-M-03 separation to 03A. trays located 13 area where HGL is not a concern.

Vertical separation feet above the Postulated transients would not down to 02B and up floor. result in fire spread to 03A.

to 04B via an open Transients are not postulated in hatch. the open hatch and there are no combustibles.

Page 4 of 11 Rev A.

Rev A. Page 4 of 11

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 04A Horizontal Transients only No combustibles N/A - horizontal Large general Reactor Building FHA-M-04 separation to 04B. separation only area where HGL is not a concern.

Postulated transients would not result in fire spread to 04B.

04B Horizontal Transients only No combustibles No Large general Reactor Building FHA-M-04 separation to 04A. area where HGL is not a concern.

Vertical separation Postulated transients would not down to 03B and up result in fire spread to 04A.

to 05A via an open Transients are not postulated in hatch is. the open hatch and there are no combustibles.

04G Vertical separation No ignition No combustibles No Fuel Pool cooling pump area FHA-M-04 to 05B is via open sources where HGL is not a concern.

hatch. Transients are not postulated in the open hatch and there are no combustibles.

05A Vertical separation No ignition No combustibles No Large general Reactor Building FHA-M-04 down to 04B and to sources area where HGL is not a concern.

06A is via open Transients are not postulated in hatch. the open hatch and there are no combustibles.

Page 5 of 11 A.

Rev A. Page 5 of 11

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 05B Vertical separation No ignition No combustibles No Fuel Pool tank area where HGL is FHA-M-04 to 04G is via open sources not a concern. Transients are not hatch. postulated in the open hatch and there are no combustibles.

06A Vertical separation No ignition No combustibles No Large general Reactor Building FHA-M-04 down to 05A is via sources area where HGL is not a concern.

open hatch. Transients are not postulated in the open hatch and there are no combustibles.

07A Horizontal Transients only No combustibles No Large Turbine Building area. FHA-M-01 separation to 07H is Postulated transients would not via open corridor. result in fire spread to 07H or 08A.

Vertical separation to 08A is via an Spatial separation is used as a open hatch. partitioning element; however, large oil fires are treated as resulting in HGL and multi compartment interactions from these fires are included in the FPRA.

Page 6 of 11 A.

Rev A. Page 6 of 11

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 07C Horizontal Transients only Exposed cable No Large Turbine Building area. FHA-M-01 separation to 07E is trays located 15 Postulated transients would not via open corridor. feet above the result in fire spread to 07E or 08C.

Vertical separation floor.

to 08C is via an Spatial separation is used as a open grating. partitioning element; however, large oil fires are treated as resulting in HGL and multi compartment interactions from these fires are included in the FPRA.

07E Horizontal Transients only Exposed cable No Large Turbine Building area. FHA-M-01 separation to 07C is trays located 15 Postulated transients would not via open corridor. feet above the result in fire spread to 07C, 08C Vertical separation floor. or 08D.

to 08C is via an open grating. Spatial separation is used as a Vertical separation partitioning element; however, to 08D is via open large oil fires are treated as stairs. resulting in HGL and multi compartment interactions from these fires are included in the FPRA.

Page 7 of 11 A.

Rev A. Page 7 of 11

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 07F Vertical separation T/G Oil Potential from oil Condenser Heater Bay. FHA-M-to 09B is via large fires 01,02 openings. Spatial separation is used as a partitioning element; however, large oil fires are treated as resulting in HGL and multi compartment interactions from these fires are included in the FPRA.

07H Horizontal Transients only No combustibles N/A - horizontal 07H is a stairway. Postulated FHA-M-01 separation to 07A separation only transients would not result in fire via a corridor. spread to 07A.

08A Horizontal Transients only Exposed cable No Large Turbine Building area FHA-M-02 separation to 08C. trays located 15 where HGL is not a concern.

Vertical separation feet above the Postulated transients would not to 07A is via open floor. result in fire spread to 07A or 08C.

stairs.

08C Horizontal Transients only Exposed cable No Large Turbine Building area FHA-M-02 separation to 08A trays located 15 where HGL is not a concern.

and 08D. Vertical feet above the Postulated transients would not separation down to floor, result in fire spread to 07C, 07E, 07C and 07E is via 08A or 08D.

open grating.

Rev A. Page 8 of 11

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 08D Horizontal Transients only Exposed cable No Large Turbine Building area. FHA-M-02 separation to 08C. trays located 15 Postulated transients would not Vertical separation feet above the result in fire spread to 08C or 07E.

down to 07E is via floor. Transients are not postulated in open stairs. Vertical the open hatch and there are no separation up to combustibles.

09C is via an open hatch. Spatial separation is used as a partitioning element; however, large oil fires are treated as resulting in HGL and multi compartment interactions from these fires are included in the FPRA.

08K Vertical separation No ignition No combustibles No Demineralizer tank area where FHA-M-02 to 09C via an open sources HGL is not a concern. Transients hatch. are not postulated in the open hatch and there are no combustibles.

Page 9 of 11 Rev A.A. Page 9 of I I

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 09B Vertical separation T/G Oil Potential from oil Turbine Generator area. FHA-M-03 to 07F is via large fires openings. Spatial separation is used as a partitioning element; however, large oil fires are treated as resulting in HGL and multi compartment interactions from these fires are included in the FPRA.

09C Vertical separation No ignition No combustibles No Large Turbine Deck area where FHA-M-03 down to 08D and sources HGL is not concern. Transients 08K is via open are not postulated in the open hatches. hatch and there are no combustibles.

16A Vertical separation Transients only No combustibles No Postulated transients would not FHA-M-15 to 16F via open result in fire spread to 16F.

stairway.

16F Vertical separation Transients only No combustibles No Postulated transients would not FHA-M-1 5 to 16A via open result in fire spread to 16A.

stairway.

Page 10 of 11 A.

Rev A. Page 10 of 11

RAI - PRA 13 PAU Credited Spatial Ignition Source Installed Potential for Fire Justification Drawing Separation in Spatial Combustibles in Spread and Separation Spatial Plume Effects Separation 17A Vertical separation No ignition No combustibles No Transients are not postulated in FHA-M-17 to 17C separation sources the open hatch and there are no via heating and combustibles.

ventilating opening.

17B Vertical separation No ignition No combustibles No Transients are not postulated in FHA-M-17 to 17D separation sources the open hatch and there are no via heating and combustibles.

ventilating opening.

17C Vertical separation Transients only No combustibles No Postulated transients would not FHA-M-1 7 up to 17A result in fire spread to 17A.

separation via heating and ventilating opening.

17D Vertical separation Transients only No combustibles No Postulated transients would not FHA-M-17 up to 17B result in fire spread to 17B.

separation via heating and ventilating opening.

Page 11 oflI Rev A.

Rev A. Page 11 of 11

RAI - PRA 16 DAEC RAI PRA 16 Table 3.3-1 of the Fire Model Development Report describes the cues for operator action "D250DCENOPCB4023HE" as 1) loss of DC power on indicating lights and 2) reactor pressure vessel (RPV) level given high pressure core injection (HPCI) fail to start. Regarding this operator action address the following:

a. F&O 5-37. The second cue was added in response to the peer review finding on the first cue that it was inadequate since it would only occur when the batteries are depleted. Regarding the "RPV level given loss of HPCI" cue, explain the timing for the human failure event (HFE), describe how the operator knows to respond appropriately to this cue, and clarify why this cue is adequate for this operator action.

b. F&O 5-38. Clarify how the "RPV level given loss of HPCI" cue addresses the peer review finding that the four hour timeline plus the additional 10 minutes access delay (for ex-control room actions) is inadequate for station blackout (SBO) and loss of offsite power (LOOP) sequences when the battery charger is not available.

RESPONSE

The Internal Events version of the D250DCENOPCB4023HE-- action (OPERATOR FAILS TO ALIGN ALT 250VDC BATT CHGR W/I 4HOURS) uses the 250V DC CHARGER 1D43 TROUBLE annunciator as a cue. This cue is not credited for the Fire PRA because annunciators are not verified as protected for fire events.

The Internal Events HEP evaluation also includes "HPCI annunciator lights" as a secondary cue. The "RPV level given loss of HPCI" cue in the Fire PRA evaluation reflects RPV level indication inclusion on the Safe Shutdown Equipment List (SSEL) and therefore the availability of indirect HPCI failure indication for fire initiators.

However, it is true that HPCI failure would be expected after battery depletion which would occur too late to provide a valid cue for this action in either the FPIE or the FPRA action evaluations. Both HEP evaluations assume that a continuous DC supply is required for success and so the operators are required to align the standby charger prior to battery depletion.

For the FPRA, the operator action was assessed for Station Blackout conditions for which operators would be able to detect and interpret environmental conditions and follow procedure cues without the need for instrumentation. However, the operator action is not modeled in the FPRA for SBO sequences. As such, the timeline for the operator action would start after battery depletion and HPCI failure. There may be very little time to perform the action without the annunciator as a cue. Therefore, the operator action would be more appropriately not credited in the FPRA consistent with other operator actions potentially without an available cue.

The operator action is not a risk significant action in the FPRA. The battery chargers and cables are located in the essential switchgear rooms. The risk significant fire scenarios in the essential switchgear rooms consist of fire induced LOOP sequences Rev A. Page 1 of 2

RAI - PRA 16 that transfer to SBO sequences. Therefore, the operator action is not significant in these areas and not crediting the action in the FPRA would not result in a noticeable increase in CDF or LERF.

Table 3.3-1 of the Fire Model Development Report, the HRA, and the basic event probability will be updated to reflect the lack of a cue and that the operator action is not credited in the FPRA. In addition, the basic event probability will be applied in the updated FPRA model in response to RAI PRA 01.

Page 2 of 2 Rev A. Page 2 of 2

RAI - PRA 17 DAEC RAI PRA 17 F&Os 5-39 and 5-42. The resolutions to these F&Os do not completely address the peer review finding to assign dependency levels for recoveries of both cognitive and execution errors no lower than the minimum recommended by the Human Reliability Analysis (HRA) Calculator. Provide a description of the dependency analysis and how dependency values were established. If necessary provide justification for using dependency levels lower than the minimum recommended by the HRA Calculator.

RESPONSE

The dependency factors (DF) used for the cognitive and execution recoveries impact the associated independent HEP values. Therefore, the dependency analysis is only impacted by any change to independent HEP values. This finding applies to the selection of DF values within the individual HEP evaluations.

The HRA Calculator establishes the suggested DF value based on the Time Available for Recovery (Trec) as given in the following table. Note that Trec is the system window minus the cue delay and the manipulation time [Trec = Tsw - (Tdelay + Tm)].

Trec DF 0 - 15min HD 16 - 30min MD 31 - 60min LD

> 60 min ZD F&O 5-39 states: The dependency levels for recovery of cognitive errors in the CBDTM methodology have not been assigned. This is non-conservative. For example, for DSYSTM-NOP-302-1 HE--, the minimum level of dependence recommended by the calculator is medium. However, a "N/A" is assigned, which is equivalent to zero dependence (non-conservative).

Response (part 2):

The Cause Based approach to estimating the cognitive failure contribution, as documented in EPRI TR-100259, does not provide any time-based structure to modify the recovery factors allowed for the methodology's individual failure mechanisms (Pea Rev A. Page 1 of 10

RAI - PRA 17 through Pch). The recovery factors (RFs) included in Table 4-1 of ERPI TR-100259 were developed to account for dependence issues that were considered to exist between the error sources and the recovery mechanisms. Some freedom is granted to modify certain recovery factors (those that are assigned values of "X" in Table 4-1) based on the degree of dependence assumed for the recovery mechanism, but the time-based application of the THERP dependency equations for those values is a construct of the EPRI HRA Calculator and not required by EPRI TR-1 00259 in any way.

Use of the "not applicable" (N/A) option in the recovery analysis is, therefore, appropriate and consistent with the EPRI TR-100259 guidance.

Additionally, using an N/A is not equivalent to ZD as stated in the finding. N/A simply allows use of the CBDTM recovery factor at its face value as indicated in the following image from the HRA Calculator Pc panel:

Recoýwy Factors'Applied t..Pc .. ** Based on 195.00 Minutes for Recovery: DOependency should not be less than ZD Branch Initial HEP Self:RAie.0 ExtraCrew STKA.Review SNhilChange ERF Review DF MultiplyBy Override Value Final Value

_R ng.C jNaI NC N:~

q

~jf~ [eg.: - ~ ~ - FNe-Zi I1.0e-2 II/A 1 O.Qe-00 2 j' ra I 'ýNC NC 1.G- X 1.0e-i1 IN_/A7)) I1.0e-01 ._. .Oe+0 f 30e-03

. NC iei --.x rT~~oiFT 3.Oe-04

_p fj- t.~2 I1Oe1 I5O. NU iOeLi N/

ne " . NC 1.0e-I1 Ce-i S1:0e-01 lT.. 1 0.oeoo0 N.~, ' C...i 1.e-i NC 1.0es-1 [fNT rl lOe-1 I _ I__2.e-04 jr ~ .. ~ 6...,~

~~:.7-,.-X..*.

. 1 NC IN/Az] I i:Oe-C f 2 0eZOO

.:-r ....* iL ; : . .,*..:,!- NO Recalculate Sum fidb~- Pc I Recoverg,-6ýýýp.,ýedfte~

For Pcd above, the 0.1 recovery factor (RF) with a dependency factor (DF) value not set (i.e., not applied) allows the applications of 1E-1 as an RF, reducing the unrecovered Pcd value of 3E-3 to be reduced to 3E-4 via STA credit. If ZD were used as allowed by the Trec value as given in the top right of the image, the Pcd value would be substantially reduced as seen in this screen capture:

Page 2 of 10 Rev A. A. Page 2 of 10

RAI - PRA 17 Recovery FactorsApplied to Pc. . Based on 195.00 Minutes for Recoyery: Dependency should'not.be less than ZD, Branch. .... ri~iai.l.HE S~! R1.f,.*'

vel 'i'ew"STA ReviewShift-Chahge ERE Review OF MultiplyB*. OverrideValue FinalValue pimj ný- NC- -0a NC 5.e1IN/A. f 7re-35 pE fWjra egý 1.0e-1 Nt e-iie NAj .B'2 ___ _ 0:+0 jF' [ rN*C" ice-i X. 0e-1 N .=/A- .J0 0.0e,00

_ d r " 5.7'0 NC 5 1e-1

.ie-i1 X i.0e- 1 3.0e.03 5 T 9e-e06 pc :Jd : i *e-i5e i N1 X jj 11.0e-1 N/A z I 1.0e-01 5."Oe-04 f71 . - NC --. e. 1C **e-i j_0Cc-i-/A NC i~-iii 13 N NC--1 I/A Foa-oi.C~ ~ Ia~o B ealculateJ Sum of recovered Pda through Pch = Recovered Pc. 7.1e-C4 Although LD or ZD DFs are allowed by the HRAC for Trec values over 30 minutes, they are generally not used because the total Pc value can become too low to be reasonable. Note that the HRAC recommends the use of ZD for Trec values over 60 minutes as noted in the table above.

For Trec values less than 30 minutes, the HRAC recommends DF values of MD or HD.

HD is recommended for Trec of 15 minutes or less and MD is recommended for 15 min

<= Trec < 30 minutes. Application of the MD or HD DFs can increase the CBDTM results. The DF equations for MD and HD yield increased recovery multipliers which reduce the recovery credit.

Recovery Factors Applied.to Pc.. Based on 195.00 Minutes for Recovery: Dependency should not be tess than ZD Branch lIiitial;HEP Seltf-.Reiew: ES-kr.4t.w ýSTAReview Shi tlChange ERF Review DF Multiply By Override Value Final Value f-NC ý NC Q. j 5.e- I/*0 N/A -Oe.OO CCe-oC

.i"e-i

rJ7 F F 1.e.* * - N't-o 1.0e-1---*: _X 1.0 1iei-I NAJ

=N--

7.0e-02 I

f .. .

- 100e-00 OO

.E!Ej ra NC NC. 0.1 Xe- iOei I/Az FN F1i5e-C f* .Ce+O

_ .* .I . NC I 1.0e-i .Oe MD j] I 15e-01 f 4.5e-04 I -d 6E . fie-i5.0NC 1 C-e-i x NC 0e-1 NC iMe-Cl _02 . 0.0e

,c .e S2mo v: c.tro1 P v1

"_.ecalculate J Sum of recovered Pca thro'ugh Pch =Recovered Pc 11.2-03 Page 3 of 10 Rev A. A. Page 3 of 10

RAI - PRA 17 An MD DF is applied to the Pcd RF, increasing the original 0.1 RF to 1.5E-1 and increasing the final Pcd value to 4.5E-4. [MD = (1 + 6N)/7] = [MD = (1 + 6(3.OE-3))/7 =

0.15]

An HD DF further reduces recovery credit:

Recovery FactorsApplied toPc" l.. Based on 195.00 Minutes fbr Recovery:. Dep'eqdeoncy should not be less than ZD Br.nch . Init.. H"E*

Selt

....Re*ie r.CreW - SA.eview .. Shi-t.hange ERF Review DF Mult y. Override Value Final Value pc aT NC, e'INC. .. 5.e-;]JM O.O0 e pcbF neg fte~l NC 1.0Cei l e-i] [N /A 1.0e.-02 .e0 P.-J. F. -NC .. "N I e- x, 1.0e-1. / z I .0 i.e.Cl 0 fbrýi ~iJ NC :oei 50,M 111.e-1 X1.Ce-i I fH-D z Ece d " * " 1c0e-1 50e-1. -i 1e-I FN/A-:-] 5D4J FT pe...;i.arr ..~. .~i.e

. .',. . ' . Ac:;*

-"....... i a.e 1.

. c-i*' X NC "N -1 1.0e" N1 =/A- I .77 Q COe+O0 N1=/A- 2.0014 IN/A ..J 4 1 De-Di F-,77 F D.Oe.CD Recalculate f Sue of recovered Pca through Pch Recovered Pc f 22e-03 An HD DF is applied to the Pcd RF, increasing the original 0.1 RF to 5.OE-1 and increasing the final Pcd value to 1.5E-3. [HD = (1 + N)/2] = [HD = (1 + 3.OE-3)/2 = 0.5]

Again, the time-based application of the THERP dependency equations for those values is a construct of the EPRI HRA Calculator and not required by EPRI TR-100259 (CBDTM) in any way.

F&O 5-42 states: Recoveries for execution errors are assigned dependence levels that are lower than the dependence levels recommended by the HRA calculator, based on event timing. This can produce non-conservative results. For example, for DN2---

ANOPGRP3BYHE--, the dependence level recommended by the Calculator for execution error recoveries is high dependence, but medium dependence was assigned by the analyst. For DSYSTMNNOPRESTRTHE-- the dependence level recommended by the Calculator for execution error recoveries is high dependence but the dependency level assigned by the analyst is low.

Response (part 3):

All credited actions within the current version of the DAEC FPRA HRAC file were reviewed. Three actions with execution recovery DF values below those recommended by the Calculator were identified. The following table summarizes each action and provides the DF values and provides an increase ratio for the three actions that were Rev A. Page 4 of 10

RAI - PRA 17 assigned a DF below those recommended. The DF used for the cognitive and execution recoveries impact the associated independent HEP values. Therefore, the dependency analysis is only impacted by any change to the independent HEP values.

As discussed below, these three actions are not risk significant and use of the recommended DF would have a negligible impact on the results.

D125DCENOPLDSHEDHE-- (OP FAILS TO LOAD SHED DC BATTERIES PER AOP 301 DURING SBO): For the FPRA, offsite power recovery is not credited. Therefore, SBO sequences are assumed to result in core damage and the action to load shed is not risk significant.

DCRD-CNOPFTRNIIHE-- (OPERATOR FAILS TO MAXIMIZE CRD FLOW FOR IORV AND MLOCA-ST): IORV and MLOCAs above the core are not risk significant sequences in the FPRA. Therefore, the action to maximize CRD flow for these sequences is not risk significant.

DSYSTM-NOP-302-1 HE-- (OP FAILS TO PERFORM LOCAL STARTS PER AOP 302.1): The action credits operator action to manually close 4KV breakers given loss of the DC control power. The FPRA models fire induced failure of the DC control power resulting in failure of the 4KV bus. Therefore, the operator action to manually close 4KV breakers is not credited in the FPRA and the action is not risk significant.

Page 5 of 10 Rev A. Page 5 of 10

RAI - PRA 17 T(rec) HRAC Revised OriginalDFP Rai BEI i rcExe Comparison Result OrgnlDF Pf Ratio BE ID Mn Trec ePf D125DCENOPALNCHGHE-- 205 ZD LD,HD CONSERVATIVE PRACTICE D125DCENOPLDSHEDHE-- 30 MD ZD NONCONSERVATIVE 1.8E-02 2.6E-02 1.44 PRACTICE D250DCENOPCB4023HE-- 195 ZD LD,HD CONSERVATIVE PRACTICE DADS--ANOP15OPSIHE-- 51.5 LD LD AS SUGGESTED DADS--ANOP-LVL2-HE-- 67.75 ZD ZD AS SUGGESTED DADS--ANOPSISTRVHE-- 31.35 LD LD AS SUGGESTED DADS--ANOPS1-WA-HE-- 1 HD HD AS SUGGESTED DADS--ANOPS2-ST-HE-- 35.65 LD LD AS SUGGESTED DADS--ANOPS2-WA-HE-- 8.35 HD HD AS SUGGESTED Page 6 of 10 Rev A. Page 6 of 10

RAI - PRA 17 T(rec) HRAC Revised BEID Mi Trec Exe Comparison Result Original DF Pf Ratio DF Pf DADS--ANOPTRANS-HE-- 37.85 LD LD AS SUGGESTED DADS--NOPAFLVL2HE-- 105.65 ZD ZD AS SUGGESTED DCBHV-NNOPDORFANHE-- 79.8 ZD LD CONSERVATIVE PRACTICE DCBHV-NNOPFTSHV-HE-- 35 LD LD AS SUGGESTED DCNDSRCNOPMECVACHE- 49 LD LD AS SUGGESTED DCNDSRCNOPPRESCTHE- 589.5 ZD ZD AS SUGGESTED DCNDSTCNOP02 ---- HE-- 70 ZD ZD AS SUGGESTED DCNDSTCNOPTINJ--HE-- 45 LD LD AS SUGGESTED D-----

CNOPLL-ST-HE-- 1.5 HD HD AS SUGGESTED D-----

CNOPML-ST-HE-- 22 MD HD CONSERVATIVE PRACTICE Rev A. Page 7 of 10

RAI - PRA 17 T(rec) HRAC Revised Exe Orgna BE ID Min Trec DF Comparison Result fginal DF Pf Ratio D-----

CNOPML-WA-HE-- 2.5 HD HD AS SUGGESTED D-----

CNOPSL-WA-HE-- 21.6 MD MD AS SUGGESTED D-----

CNOPTRSLSTHE-- 33.9 LD MD CONSERVATIVE PRACTICE DCRD--CNOPFTRNI-HE-- 18.5 MD MD AS SUGGESTED DCRD--CNOPFTRNIIHE-- 3.75 HD MD NONCONSERVATIVE 9.7E-02 1.3E-01 1.34 PRACTICE DCRD--CNOPFTRNILHE-- 130.99 ZD LD CONSERVATIVE PRACTICE DFEED-CNOPSTRT--HE-- 32.75 LD MD CONSERVATIVE PRACTICE DHPCI-CNOP15---- HE-- 34.5 LD N/A NO PE RECOVERY APPLIED DHPCI-CNOPOPENDRHE-- 25 MD MD AS SUGGESTED Page 8 of 10 Rev A. Page 8 of 10

RAI - PRA 17 T(rec) HRAC Revised OriginalDFP Rai BEIDMi TecDIF Exe Comparison Result Pf DF Pf Ratio ON2---ANOPGRP3BYHE-- 29.5 MD MD AS SUGGESTED DPHVACNNOPPHDORSHE- 75 ZD ZD AS SUGGESTED DRCIC-CNOP2LPTRIHE-- 168 ZD ZD AS SUGGESTED DRHR--CNOPSPCELYHE-- 198 ZD ZD AS SUGGESTED DRHR--CNOPSPCNATHE-- 1175 ZD ZD AS SUGGESTED DRHR--CNOPSPRYM-HE-- 18.33 MD MD AS SUGGESTED DRHR--CNOPSPRYS-HE-- 18.5 MD MD AS SUGGESTED DRHRSWDNOPLTINJ-HE-- 10 HD HD AS SUGGESTED DSBGT-CNOP-VENT-HE-- 432 ZD ZD AS SUGGESTED DSYSTMNNOPRESTRTHE-- 42 LD LD AS SUGGESTED DSYSTM-NOP-302-1 HE-- 25 MD LD NONCONSERVATIVE 1.5E-02 1.8E-02 1.20 PRACTICE Rev A. Page 9 of 10

RAI - PRA 17 T(rec) HRAC Revised BE ID Min Trec Exe Comparison Result DF Pf Ratio DF Pf DSYSTM-NOP-PCFLDHE-- 235 ZD ZD AS SUGGESTED DTSC--ENOPALNTSCHE-- 95 ZD ZD AS SUGGESTED DWELLWDNOPELLWTRHE- 267 ZD LD CONSERVATIVE PRACTICE FMCR-HVAC-PURGE-HE-- 5 HD N/A NO PE RECOVERY APPLIED FMSIV--CLOSURE--HE-- 12 HD N/A NO PE RECOVERY APPLIED FRHR-SPC-M02010-HE-- 105 ZD ZD AS SUGGESTED FRPS-SCRAM-20 MIN 5 HD N/A NO PE RECOVERY APPLIED FSBDG1G21-START-HE-- 10 HD HD AS SUGGESTED Page 10 of 10 Rev A. Page 10 of 10

PRA Question 21 PRA Question 21 According to the disposition of F&O 4-32, an unreviewed method using a severity factor of 0.08 was removed for transient fire scenarios. Is this evaluation now consistent with NUREG/CR- 6850? If not, but, instead, another alternate approach or factor was used, describe that approach. If there remains a deviation from NUREG/CR-6850, perform a sensitivity study using NUREG/CR-6850 or NRC-endorsed frequently asked questions (FAQs).

RESPONSE

The transient fire severity factor is treated consistent with NUREG/CR-6850. An alternate approach or factor was not used. Given the severity factor was removed and an alternate approach or factor was not used, there is not a deviation from NUREG/CR-6850.

Page 1 of I Rev A.A. Page 1 of 1

RAI - PRA 22 DAEC RAI PRA 22 With respect to F&Os 3-4 and 4-43 concerning the screening of fire events, fire event action request (AR) number 00306845 was listed as a non-potentially challenging fire for purposes of the update of the generic fire frequency. This fire occurred near the exhaust line by the turbocharger associated with the emergency diesel generator (EDG). Lube oil was absorbed by the insulation and, according to the report, 60 to 80

% of the over piston oil leaked past the gasket onto the floor of the room with a good probability. According to plant personnel, a few gallons of oil could have leaked on the floor. The report also states that the fire was of sufficient size to consume the lube oil absorbed by the insulation. The AR report describes an operator extinguishing the fire with water mist, but later states the fire self-extinguished. The following additional information is necessary to understand the screening of this event as non-challenging (in the response to each, be clear with respect to whether the factors are being judged solely on the basis of the event description in the AR Report, from other records, or discussions with personnel familiar with the event):

a. Provide a specific discussion of this event relative to the NUREG/CR-6850 criteria for potentially challenging and undetermined events to determine if this fire should be counted in the Bayesian update from DAEC experience. In this regard, describe whether this fire caused damage to equipment or cables in the EDG room, ignited secondary combustibles, or affected the function of the EDG.

Also, indicate if any fixed suppression system was actuated as a result of this fire, or if the fire brigade contributed to the suppression of the fire. Also, indicate if more than one portable extinguisher was used to extinguish the fire. In summary, this analysis should address the objective classification criteria identified in Section C.3.3 of NUREG/CR-6850.

b. To support the more subjective criteria of Section C.3.3., further assessment should be done. Provide confirmation regarding the amount of oil that could have leaked on the floor. Also, to determine if such a fire would be extinguished regularly, indicate if this event could have occurred during a normal automatic start and load of the generator when an operator may not be present to immediately extinguish the fire. Indicate whether this operator is trained in the use of a fire extinguisher and describe his/her other duties besides monitoring the situation for a fire. In summary, the ability to regularly extinguish a fire of the type cited in the event is important to establishing whether the fire is potentially challenging.

Furthermore, NUREG/CR-6850 credits prompt suppression by plant personnel via the fire suppression curves which include short duration fires such as this event. Except for cases involving a designated fire watch, fire duration was not a factor in determining whether or not an event was potentially challenging.

Hence, if such a fire could have occurred without the presence of an operator trained in the use of a fire extinguisher (i.e., having equivalent qualifications to a fire watch) and having no significant distractions from identifying and extinguishing the fire, describe the potential extent of fire propagation or damage Rev A. Page 1 of 8

RAI - PRA 22 in the absence of the operator (no prompt suppression). Indicate if the oil pool described in the incident report could have caught fire, and potential damage and fire fighting and systems response under those conditions.

It should be noted that a fire would not be potentially challenging if it can be established that the fire would not have propagated beyond the EDG or caused substantial damage to the EDG (e.g., prevents the EDG from operating) in the absence of an operator. If the fire could have caused substantial damage to the EDG, become more involved through igniting the oil pool, or propagated beyond the EDG through some other means, this fire would have been potentially challenging.

c. Provide an evaluation of each of the other EDG fires which occurred during the period of the Bayesian update in light of the considerations discussed above, and characterize each as either potentially challenging, not challenging, or undetermined. Clarify that those characterized as potentially challenging or undetermined were appropriately included in the update.

RESPONSE

a. According to Appendix C of NUREG/CR-6850 "for diesel generators: a manifold fire (very common) is potentially challenging if it spread to a secondary fuel or if the initial fuel source is continuous... In contrast, a fire that causes substantial damage to the diesel generator but does not damage or spread to any other components should be classified as potentially challenging." The fire associated with 00306845 happened due to oil leaking past the pistons accumulating in the exhaust piping. This lubrication oil leaks out to the environment through gaskets while the exhaust piping temperatures were not high enough to vaporize the oil.

Some of the oil in the exhaust piping that leaked passed the gaskets was absorbed by the exhaust piping insulation (lagging). Also according to the report some of this lubrication oil leaked onto the floor of the room. In the evaluation of these events, exhaust fires are a known potential condition for Fairbanks-Morse Opposed Piston Diesel Engines. The oil leaked onto the floor would reduce the amount of oil trapped in the insulation, and therefore would reduce the time required for the fire to self extinguish. In fact these fires have been eliminated at other plants by allowing the oil to drip away from the insulation. Evaluations of these types of fires have found they are of sufficient size to consume the lube oil absorbed by the insulation, and after the exhaust temperatures are high enough to vaporize the exhaust lubrication oil as the diesel ran the fire would have self extinguished if allowed. This fire did not have a continuous fuel source, did not spread to secondary fuel sources, and did not damage the diesel generator, so the fire was classified properly as not potentially challenging in accordance with NUREG/CR-6850 guidance. This event was not counted in the Bayesian update from DAEC experience since it was not a potentially challenging fire.

No evidence was presented in the corrective action to indicate this fire caused any damage to the EDG including equipment or cables in the EDG room, and the EDG was considered operable with no EDG function affected. This is consistent Rev A. Page 2 of 8

RAI - PRA 22 with the fact that these fires occur in an area completely surrounded by metal with no other combustibles present.

There was no indication in the corrective action documentation that a fixed suppression system actuated as a result of this fire and it does not appear the fire brigade was called out or contributed to the suppression of this fire. This consideration aligns with the words in the corrective action document where the operator characterized the fire as 'small'.

The number of portable extinguishers used to extinguish the fire was not provided in the corrective action documentation. However the corrective action documentation points to the consideration it was probably only one. The fire brigade did not appear to have been called. The operator considered the fire to be small and was able to promptly extinguish the fire.

There was no indication in the corrective action documentation that a fixed suppression system actuated as a result of this fire. This consideration aligns with the words in the corrective action document where the operator characterized the fire as small in nature and was able to promptly put the fire out.

Another consideration is the heat generated as a result of lubrication oil flames is negligible with respect to the engine performance. The normal operating temperature of the manifold exhaust is between 800 and 1000 degrees Fahrenheit, which are temperatures well above the flash point of lubrication oil.

It appears only one fire extinguisher was used and no fixed fire suppression systems were used to fight the fire. No components outside the boundaries of the fire source were affected, and no combustible materials outside the boundaries of the fire ignition source were ignited. The corrective action documentation did not indicate an actuation of the automatic detection system.

The plant did not experience a trip. There was no appreciable damage incurred to the diesel engine or the surrounding area. The fire was promptly suppressed so in accordance with the objective classification criteria of Section C.3.3 of NUREG/CR-6850 this fire should be classified as a non-potentially challenging fire.

b. During normal operation an engine will consume one to one and a half gallons an hour of lube oil. Under no load or low load operation the exhaust temperatures are not hot enough to burn the oil, so it collects in the exhaust manifolds.

Experience has shown that the gaskets in the connections between the manifold and the engine can lose their compression due to uneven thermal expansion.

This loss of compression allows the oil in the exhaust to be forced out by the exhaust pressure and leak onto the heat shield. After considering the amount of lubrication oil the engine normally burns and the amount of oil in the manifold that could be pushed past the gaskets, NextEra Energy Duane Arnold estimates that during an engine start it is unlikely the amount of oil that could leak onto the floor is more than a pint. The oil on the floor does not communicate with the flames created by the lube oil trapped in the heat shield lagging, so the oil does not Rev A. Page 3 of 8

RAI - PRA 22 present another communication method for the flames to spread. Also oil leakage to the floor reduces the amount of oil trapped in the lagging, and would reduce the time required for the fire to self extinguish.

The ignition source in this event was extinguished by the operator but could have been allowed to burn until the fire self extinguished within two to five minutes later. The operator is a trained member of the Fire Brigade, and as such, attends two weeks of initial Fire Brigade training which includes extinguisher training, cyclic Fire Brigade training, and attends live fire school yearly. The operator would be performing other duties as assigned during the surveillance. This type of fire will self extinguish in an estimated 2 to 5 minutes, so the operator should not have to extinguish this fire and the fire was not potentially challenging.

The fire in this event was extinguished by the operator but could have been allowed to burn until the fire self extinguished within two to five minutes. This fire did not cause any damage to the EDG including equipment or cables in the EDG room, and the EDG was considered operable with no EDG function affected.

These fires occur in an area completely surrounded by metal with no other combustibles present. The oil on the floor does not communicate with the flames created by the lubrication oil trapped in the heat shield lagging, so the oil does not present another communication method for the flames to spread. These fires do not cause damage to the emergency diesel generators, and after these flame events occur the emergency diesel generators are found to be operable.

c. The following table contains a list of the EDG fires that occurred during the period of the Bayesian update from January 1, 2001 to March 31, 2008. The EDG fires were found to be not potentially challenging in accordance with NUREG/CR-6850 guidance "for diesel generators: a manifold fire (very common) is potentially challenging if it spread to a secondary fuel or if the initial fuel source is continuous... In contrast, a fire that causes substantial damage to the diesel generator but does not damage or spread to any other components should be classified as potentially challenging." With the above specific guidance for this type of fire, no fires were found to be potentially challenging, therefore the current Bayesian update was appropriate. The significant considerations for these fires were they will self extinguish if allowed and the fire is in an area surrounded by metal so the fire does not propagate beyond this area.

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RAI - PRA 22 Date (Corrective Action Number) Location Evaluation Characterization 12/4/03 A SDG The fire did not have a Not potentially (00300961) Manifold continuous fuel source, challenging Exhaust Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

1/9/04 A SBDG The fire did not have a Not potentially (00301270) Manifold continuous fuel source, challenging Exhaust Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

5/20/04 B SBDG The fire did not have a Not potentially (00302661) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

7/10/04 A SBDG The fire did not have a Not potentially (00303207) Manifold continuous fuel source, challenging Exhaust Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

8/8/04 A SBDG The fire did not have a Not potentially (00303522) Manifold continuous fuel source, challenging Exhaust Fire the fire did not spread to secondary source, and did not cause Page 5of8 Rev A.

Rev A. Page 5 of 8

RAI - PRA 22 Date (Corrective Action Number) Location Evaluation Characterization significant damage or affect the operability of the Standby Diesel Generator.

9/22/04 B SBDG The fire did not have a Not potentially (00304039) Manifold continuous fuel source, challenging Exhaust Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

4/13/05 B SBDG The fire did not have a Not potentially (00306845) Turbocharger continuous fuel source, challenging Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

11/21/05 B SBDG The fire did not have a Not potentially (00309964) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

1/8/06 B SBDG The fire did not have a Not potentially (00310594) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

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RAI - PRA 22 4/8/06 B SBDG The fire did not have a Not potentially (00312406) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

10/8/06 B SBDG The fire did not have a Not potentially (00315659) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

3/4/07 A SBDG The fire did not have a Not potentially (00319025) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

3/25/07 B SBDG The fire did not have a Not potentially (00319596) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

8/26/07 A SBDG The fire did not have a Not potentially (00322994) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

Rev A. Page 7 of 8

RAI - PRA 22 9/3/07 A SBDG The fire did not have a Not potentially (00323169) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

10/5/07 A SBDG The fire did not have a Not potentially (00323907) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

10/7/07 B SBDG The fire did not have a Not potentially (0323956) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

11/4/07 B SBDG The fire did not have a Not potentially (00324604) Exhaust continuous fuel source, challenging Manifold Fire the fire did not spread to secondary source, and did not cause significant damage or affect the operability of the Standby Diesel Generator.

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RAI - PRA 24 PRA Question # 24 F&O 1-7 concerned the counting of fixed ignition sources for determining the fire ignition frequency (FIF) and noted several potential discrepancies between the PAU ignition source data sheet (ISDS) and items observed in the peer review walkdown. The disposition refers to PI 07-06, in addition to NUREG/CR- 6850 and FAQ 06-0016, as the basis for screening various items when counting fixed ignition sources. This document includes in Section 2.3.2.2 the instruction that all motor operated valves (MOVs) and all transformer rates less than 45 kVA are noncountable items. FAQ 07-0031 Section 6.2.1 provides limitations on the exclusion of MOVs while Section 6.2.3 provides limitations on the exclusion of transformers that are not reflected in PI 07- 006. Provide a comparison of the criteria for inclusion or exclusion of various fixed ignition sources with the criteria in NUREG/CR-6850 and approved FAQs and justify any differences.

RESPONSE

PI-07-06, Fire Ignition Frequency Development, provides general project instructions to assist with the implementation of the guidance in NUREG/CR-6850 and approved FAQs and is not intended to repeat the guidance. Section 2.3.2 of PI-07-06 is titled "Items Not Generally Counted as FISs." The first sentence of the section states, "The following items are generally excluded as not significant and should not be included on the walkdown sheets unless there are extenuating circumstances that would cause these items to be more likely to cause a fire or that could damage high value targets." The purpose of the instruction is to ensure fire ignition frequencies are not diluted by counting components not included in the NUREG/CR-6850 generic frequencies.

Table B-1 of the FPRA Overview Report identifies that the clarifications in FAQ 07-0031 were incorporated into the fixed ignition source counting. Consistent with Section 6.2 of NUREG/CR-6850, Supplement 1 (based on FAQ 07-0031), MOV motors were not excluded if not totally enclosed and oil filled transformers were not excluded regardless of size.

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RAI - PRA 25 DAEC RAI PRA 25 F&O 4-16 concerns the apparent lack of modeling the fire induced opening of all 6 safety relief valves (SRVs) and that this scenario can be more limiting than 2 SRVs or atmospheric depressurization system (ADS), due to the thermal transient that results and this could result in a change to the success criteria, accident sequences, etc. that are presently modeled in the FPRA. The disposition states that spurious opening of all SRVs treatment is consistent with the FPIE PRA. As discussed during the audit, the opening of multiple SRVs (up to 4 for spurious ADS) are all modeled in the FPIE PRA as large steamline failures in containment. It was also stated during the audit that the worst case fire induced multiple-spurious operation (MSO) would lead to spurious opening of 5 SRVs. Document the maximum number of SRVs open due to MSO and provide a justification that the opening of this number of SRVs does not change the success criteria.

RESPONSE

The maximum number of SRVs postulated to open simultaneously due to multiple-spurious operation is six. This occurs for a fire scenario initiating within Fire Area CB1 as documented in the DAEC Fire Risk Evaluations report. The flow area of six SRVs is calculated to be approximately 0.83 ft2 based on a throat diameter of 5.03 inches for each valve. This can be compared to the flow area of a single main steamline, which is approximately 1.7 ft2 based on an inside diameter of 18 inches. Therefore, the rate of inventory loss from the reactor vessel when six SRVs open at once is less severe than the rate of inventory loss due to a double-ended break of a single main steamline.

Steam breaks within the drywell are evaluated for the DAEC using MAAP and are documented in the Level 1 MAAP notebook. A case with the identifier of DAEC819 shows that the fuel is successfully cooled with a single LPCI pump, even when a large rupture area of 3.14 ft2 is assumed. Therefore, a single LPCI pump would also be expected to successfully cool the fuel in the event that six SRVs open simultaneously. (Since the rate of inventory loss is relatively high for large steam breaks, low pressure injection must be initiated fairly quickly. This is demonstrated in the MAAP case with identifier DAEC818.

In this run, the same size rupture is assumed as in DAEC819, but with no injection by low pressure pumps. Onset of fuel damage is predicted to occur within only 14 minutes.)

Suppression pool temperature rise as a function of time for the spurious opening of six SRVs is expected to be almost identical to the profile for inadvertent actuation of the Automatic Depressurization System (ADS), where four SRVs open simultaneously.

This is due to the fact that approximately the same amount of steam is transferred to the pool for both cases in transitioning from a high pressure state to a low pressure state.

In addition, decay heat over time is the same in both cases. Therefore, following the depressurization phase, energy from the condensation of steam will be added to the suppression pool at the same rate. Since the rate of energy transfer to the suppression pool is essentially no different for the opening of six SRVs than it is for a normal depressurization event with ADS, the success criteria for cooling of the suppression pool by the RHR system is not changed. For the DAEC PRA, this is met with operation Rev A. Page 1 of 2

RAI - PRA 25 of a single RHR heat exchanger with one RHR Service Water pump and one RHR pump running.

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RAI PRA 26 DAEC RAI PRA 26 F&O 5-16 states that the sources of LERF model uncertainty and related assumptions have not been identified or documented. The disposition states that "[t]he quantification and summary report is updated to document the sources of LERF model uncertainty and assumptions." The resolution refers to two internal documents: DAEC-PSA-L2-15 and DAEC-PSA-QU-14. Provide a summary of the methodology used and the identified sources of LERF model uncertainty and assumptions described in these reports.

RESPONSE

The DAEC Fire PRA did not identify initial sources of LERF model uncertainty beyond those identified in the DAEC Full Power Internal Events PRA (FPIE PRA). The FPIE PRA process used to identify sources of model uncertainty was developed consistent with NUREG-1855 and the complimentary EPRI guidance TR EPRI 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments".

The FPIE PRA evaluation reviewed the generic list of modeling uncertainties provided in the EPRI document and performed an assessment of plant specific features and modeling approaches to determine if additional sources of model uncertainty and related assumptions should be incorporated into the list. The generic modeling uncertainties evaluated for a Mark 1 containment boiling water reactor like DAEC were:

The uncertainties associated with crediting core melt arrest in-vessel during Loss of Offsite Power events, during events where the vessel is at high pressure during the time between core damage and vessel failure, and during events where reactor vessel depressurization occurs after core damage but before the time of vessel breach. The conservative treatment for core melt arrest was determined to be such that it was considered to either not be a source of model uncertainty or changing the modeling method would not cause the risk metrics to approach any acceptance guidelines for applications.

" The uncertainties in modeling several phenomenological conditions that could lead to early containment failure which are dependent upon the vessel failure mode. These issues are: 1) RPV catastrophic failure, 2) direct containment heating, 3) ex-vessel steam explosion, 4) core-melt progression overwhelms vapor suppression capabilities or otherwise leads to containment failure, and 5)

Pedestal differential pressure causes structural failure and loss of containment integrity. The conservative treatment for modeling these phenomenological conditions was determined to be such that it was not considered to be a source of model uncertainty.

The uncertainties of crediting lower vessel head cooling external to the vessel by flooding the containment. The DAEC FPIE PRA model does not credit external vessel cooling or flooding containment without Core Spray keeping the core cooled while flooding is progressing. This modeling treatment was determined to be such that it was not considered to be a source of model uncertainty.

The uncertainties of characterizing the potential for core debris to contact the containment and challenge the integrity of the containment boundary. The DAEC FPIE PRA model explicitly models this phenomenon consistent with NUREG/CR-Page 1 of 2 Rev A. Page 1 of 2

RAI PRA 26 5423 and NUREG/CR-6025, therefore the modeling uncertainty is not expected to challenge any acceptance guidelines for anticipated applications.

The uncertainties of characterizing the potential for Interface System Loss of Coolant Accident, especially in relation to common cause ruptures of isolation valves between the RCS/RPV and low pressure piping. Given an overpressure condition in low pressure piping, there is uncertainty surrounding the failure mode of the piping. Fire initiating events that open these valves should govern the results for this application compared to valve generic failure rates including random common cause. Also this application does not credit the overpressure capability of the piping in the initiating event frequency, so it is not a source of model uncertainty.

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RAI - PRA 30 DAEC RAI PRA 30 The fire transition LAR utilizes a number of probabilistic computer codes in performing the risk assessment. Different codes were used for different purposes and the results sometime vary where they would be expected to agree. For total CDF and LERF the result on page W-4 of Attachment W of the LAR indicate that FRANC gives the highest value compared to XINITS. The benchmarking of XINITS in Appendix K of the Fire Risk Quantification Report indicates that the CCDP and CLERP for XINITS are higher than XINITS results on a scenario by scenario basis in Appendices L and M of the Fire Risk Quantification Report indicates that XINITS may give higher results for one scenario while FRANC may give higher results for another scenario. Provide a discussion of these differences in results, their cause and the impact of the differences on the FPRA results relative to the risk measures used for the NFPA 805 transition. Provide a discussion of the limitations of the software and their use and the use of the results from them.

Clarify which software results were presented in the LAR for transition to NFPA 805, and which software is (are) planned to be used post-transition.

RESPONSE

The FRANC software was used to develop and quantify the fire scenarios for the FPRA.

The FRANC software calculates the CCDP/CLERP for each fire scenario separately.

For the FPRA, the CCDP/CLERP was calculated at a truncation limit of 1E-9. The quantification results in a cutset file for each fire scenario. Subsequently, the FRANC software calculates the fire scenario CDF/LERF as the product of the CCDP/CLERP, non suppression probability (NSP), severity factor (SF), and fire ignition frequency (FIF).

While the quantification is sufficient to analyze fire scenarios individually, an integrated quantification that includes all of the fire scenarios is beneficial when identifying risk insights using importance measures and performing uncertainty evaluations, as well as other aspects of the quantification process.

The XINITS software was used as a method to obtain an integrated quantification that included all of the fire scenarios. The XINITS software inserts the fire scenarios from FRANC as initiators into the CAFTA fault tree logic. The fire scenario initiating frequency was included as the product of the NSP, SF, and FIF. The FPRA CDF/LERF was then quantified at a truncation limit of 1E-1 0/5E-1 1 to obtain a single cutset file consisting of the fire scenarios as initiating events.

The FRANC model and the CAFTA fault tree created from the XINITS software were quantified using the same quantification method. The FTREX software was used for quantification. The FTREX software calculation uses the minimum cut upper bound approximation. Differences in the results arise given the different risk metric and truncation limit which each was quantified. As discussed above, the FRANC model results were based on the minimum cut upper bound approximation of CCDP/CLERP for each fire scenario and the CDF/LERF was subsequently calculated. The XINITS Rev A. Page 1 of 2

RAI - PRA 30 CAFTA fault tree model results were based on the minimum cut upper bound approximation of CDF/LERF for all fire scenarios.

The results presented in Attachment W of the enclosure to the License Amendment Request (ML11221A280) (LAR) are from the FRANC model and the XINITS CAFTA fault tree model. The fire risk insight results presented in Tables W-1 and W-2 of the LAR are based on the XINITS CAFTA fault tree model. The fire risk evaluation results presented in Tables W-3 and W-4 of the LAR are based on the FRANC model.

The FRANC model quantification results are presented in Table W-4 of the LAR and have a CDF and LERF point estimate of 5.73E-5/yr and 3.76E-5/yr, respectively. The XINITS CAFTA fault tree model quantification results are presented in the FPRA Quantification Report and have a CDF and LERF point estimate of 5.71 E-5/yr and 3.58E-5/yr, respectively. The comparison above shows that the minimal cut upper bound approximation of the FRANC model provides a slightly larger CDF and LERF point estimate. While the CCDP/CLERP results for individual scenarios may be larger using the FRANC model results or the XINITS CAFTA fault tree model results, there is not expected to be a noticeable difference in the delta risk calculations performed as part of the fire risk evaluations. The fire risk evaluation results were based on the delta risk between a variant and compliant case for each fire scenario. The delta of the minimal cut upper bound approximation for the variant and compliant case is expected to be similar given the same model is used for each case.

The FRANC model results were used for the risk measures used for the NFPA 805 transition. The FRANC model provided the ability to easily assess individual fire scenarios during the fire risk evaluation process. A limitation of the FRANC software is that results for all fire scenarios are not integrated. An additional step to integrate the results is beneficial in completing the quantification process. As such, the FRANC and XINITS software are planned to be used post transition.

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RAI PRA 31 PRA Question 31 F&O 1-1 cites Appendix G of the Fire Model Development Report for the disposition of MSOs. For several MSOs (4b thru 4e) the resolution given is that containment overpressure was removed from the SSA and not required for the PRA. During the audit this was described as not requiring containment overpressure to insure that emergency core cooling system (ECCS) pumps taking suction from the torus had adequate net positive suction head (NPSH). Further it was stated that the internal events PRA did not take credit for the ECCS pumps for situations where containment over pressure is lost such as after containment venting. Provide a discussion of this inconsistency between the safe shutdown analysis (SSA) and the PRA and the potential impact on the results of the FPRA used for the transition. Also, describe the meaning of N/A in this appendix and the basis for it.

RESPONSE

Available NPSH for pumps of the ECCS systems (RHR, HPCI, CS) and for RCIC is evaluated in Extended Power Uprate (EPU) report T0406, "ECCS Net Positive Suction Head." For events associated with the Safe Shutdown Analysis, peak suppression pool temperature remains well below 212 0 F. With subcooled torus water, substantial margin to required NPSH exists, and there is no need to rely on increased torus airspace pressure for reliable RHR and Core Spray pump operation. Therefore, containment overpressure is not credited in the Safe Shutdown Analysis.

For the NFPA 805 project, more challenging containment conditions are encountered due to consideration of multiple spurious operations and due to consideration of equipment and operator failures beyond those associated with effects of the fire alone.

In the PRA model, adequate NPSH for operation of RHR, Core Spray, HPCI, and RCIC pumps is assumed to exist early in accident sequences, when the primary containment is still intact. This is a reasonable assumption even though EPU report T0406 concludes that containment airspace pressure needs to be approximately 20 psia to have sufficient NPSH for long term (> 10 minutes) injection in response to a large break LOCA. Tests performed at Browns Ferry and Monticello in the 1980s showed substantial margin for ECCS pumps of the type used at DAEC with respect to manufacturer NPSH limits. In addition, accident scenario calculations performed for licensing purposes typically over-predict severity of containment conditions since they employ conservative inputs compared to those used in the performance of PRA based calculations (Ref: DAEC Event Tree Notebook (DAEC-PSA-ET-01), Attachment 3-1.)

Since ECCS pumps are judged to be capable of operating in the absence of containment pressures greater than atmospheric, it can be said that "containment overpressure" is not credited in PRA analyses. Nonetheless, at the time at which the primary containment is vented, or, when it ruptures due to excessive pressure, the RHR and Core Spray systems are assumed to fail. Effects considered in the PRA include steam binding of pumps due to excessive flashing of saturated water and harsh environmental conditions with the reactor building. When the primary containment is no longer intact, credit is not taken in the PRA for throttling of flow to reduce required Rev A. Page 1 of 2

RAI PRA 31 NPSH, nor is alignment of Core Spray to the condensate storage tank included in the model (Ref: DAEC Event Tree Notebook (DAEC-PSA-ET-01), Attachment 3-1.) This treatment is conservative, yet reasonable given the adverse conditions present in accident sequences where normal heat removal systems are unavailable.

In summary, there is no inconsistency between the SSA and the PRA, and, resolutions provided in Table G-1 of the Fire PRA Model Development Report (0493080001.002) for BWROG multiple spurious operation items 4b through 4e are stated correctly.

Containment overpressure is not credited for ECCS pump operation in either the SSA or the PRA. Pump failure is assumed in the PRA following loss of containment integrity, but this is due to consideration of factors other than loss of NPSH alone, i.e., excessive steam formation and/or creation of adverse environmental conditions in the pump rooms.

Table G-1 of the Fire PRA Model Development Report (0493080001.002) contains the disposition of items in the BWROG Multiple Spurious Operation list. The abbreviation N/A is used in the column labeled "Final PRA Resolution" to designate items that are not applicable to the DAEC, and to designate items that are already considered in the PRA model and for which no further action is needed to include them.

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RAI - PRA 32 DAEC RAI PRA 32 Attachment G of the Fire Scenario Report identifies a recovery action (RA) to locally close cross-tie manual valve, V1 9-0048. The discussion in Section 6.2 of the Fire Quantification Report (p. 84/491) states that the cross-tie manual valve, V19-0048, is closed to isolate spuriously operating valves that could result in flow diversion and while this action (to close the valve) would prevent use of the cross tie which is credited in the PRA, it is not considered an adverse action as it would isolate flow diversion. Table E-3 of the FSR states that MO-201 0 (a valve that also isolates the cross-tie line) is assumed closed in the FPRA for fire areas RB1 and RB3 so as to not credit the cross-tie.

Discuss further the modeling associated with this cross-tie, its being credited or not in the FPRA, and how the operator closing the manual valve does not adversely impact the sequences where credit is taken for the cross-tie.

RESPONSE

As described in Section 6.2 of the Fire Quantification Report, current fire procedures were reviewed to determine if any proceduralized recovery actions could have an adverse impact on the fire PRA. In the current abnormal operating procedure for fire, AOP 913, control room operators are directed to close manual isolation valve V19-0048, which is located on the RHR A to RHR B cross-tie line. This action is taken for fire in areas RB1 and RB3. For fire area RB1, the Division 1 RHR and RHR Service Water systems are relied upon for suppression pool cooling. Closing manual valve V19-0048 ensures that water supplied by RHR Pumps A and C is not diverted to the reactor vessel or drywell via Division 2 RHR valves that may have spuriously opened. For fire area RB3, the Division 2 RHR system is relied upon for injecting water into the reactor vessel. Closing V19-0048 ensures that water supplied by RHR Pumps B and D is not diverted to the torus or drywell via Division 1 RHR valves that may have spuriously opened. The action to close V19-0048 was identified as not having an adverse impact since the potential for certain flow diversions are avoided.

In the Fire PRA, the RHR cross-tie line is assumed to be isolated for all fires initiated in RB1 and RB3. This treatment is consistent with actions specified in the present version of AOP 913, and has only a small impact on CDF as a result of the inability to supply water from one division of RHR to the other in these scenarios. Per the design basis of the DAEC, the RHR cross-tie line is needed for LOCA events so that water from at least three of the four RHR pumps can be delivered to the intact recirculation loop. However, thermal-hydraulic calculations using realistic inputs and assumptions show that a single RHR pump can inject enough water into the reactor vessel to prevent core damage, even for large break LOCA events. As such, the actual importance of the cross-tie line is much less than what would be expected if three pumps were needed to prevent core damage. The same is true with regard to the suppression pool cooling mode of RHR.

Only one RHR pump delivering water to the torus through a single RHR heat exchanger is needed to maintain integrity of the primary containment. It is unlikely the cross-tie would be needed to get water from one division of pumps to injection valves in the opposite division. The low importance of the RHR cross-tie line is reflected in the Page 1 of 2 Rev A. Page I of 2

RAI - PRA 32 internal events PRA model. When the cross-tie valves are assumed to always remain closed, CDF increases by only one percent.

Actions prescribed in AOP 913 to close cross-tie valve V1 9-0048 for fire in RRB1 and RB3 are not required recovery actions for NFPA 805. If the actions are preserved, their inclusion in the Fire PRA model as currently written is appropriate. Ifthey are deleted from AOP 913 for NFPA 805 implementation, the Fire PRA should be revised to accurately reflect the revised procedure.

Attachment G of the enclosure to the License Amendment Request (ML11221A280)

(LAR) identifies a recovery action to locally close V19-0048 upon control room abandonment in the event of fire in the control room. The Fire PRA treatment for control room abandonment does not explicitly require this action to meet the risk assessment criteria for performance-based evaluations. However, it is concluded in Attachment G of the LAR that although no recovery actions listed in Table G-1 are deemed necessary to meet the risk acceptance criteria, they are necessary to maintain a sufficient level of defense-in depth.

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RAI - PRA 33 DAEC RAI 33 F&O 4-12 concerns the impact of fire on the Level 2 model. The disposition states that events in the Level 2 model are generally conditional events, phenomenon events, and long term human actions and that none of these events would be considered impacted by a fire as discussed in the review. Provide the definition of "long term" and the basis for saying that long term human actions would not be impacted by a fire considering, in particular, the loss of instrumentation that might occur.

RESPONSE

NUREG/CR-6850, Section 12.5 provides guidance in assigning screening post fire human error probabilities. In Section 12.5.3.6, the guidance selects a time of one hour for when the fire can be assumed out and not continuing to be causing delayed spurious activity or other late scenario complicating activities. In addition, draft NUREG-1921 states: "Long term actions are those that are not required during the early stage (e.g.,

first hour) of a fire event and are not expected to be performed until approximately one hour after the fire initiation and trip of the plant or until the fire is out. Thus, short term actions are those required within the first hour of a trip."

For the FPRA, a review of DAEC-PSA-L2-15 was performed and identified that the Level 2 model credits additional operator actions in the Level 2 logic that were not in the Level 1 logic. Some of these actions were not credited in the FPRA. The other actions would be performed after the first hour and credit NSCA instrumentation. The instruments were assessed for each fire area and at least one instrument was determined to be available.

Section 6 of the Fire Model Development Report will be updated to include the details included in the RAI response.

Operator Action Basic Event Description DAEC-PSA-L2- Time Credited 15 Section Required Instrumentation DADS--ANOP-LVL2-HE-- Operator Fails to C.3 2 hr RPV Level Depress Before RPV Fails Given Operator Failed in Level 1 DSYSTM-NOPALTINXHE-- Operator Fails to Align C.4, C.7 1-3 hr RPV Level Alternate Injection Sources in Level 2 DSYSTM-NOPTERMINHE-- Operator Intervenes C.4, C.7 1-3 hr RPV Level and Terminates Injection Page 1 of 2 Rev A. Page 1 of 2

RAI - PRA 33 Operator Action Basic Event Description DAEC-PSA-L2- Time Credited 15 Section Required Instrumentation DPCONTNNOPCOINJ-HE-- Operator Restores C.6 1-3 hr RPV Level Coolant Injection After Control Rods are Melted DPCONTNNOPPHSLC-HE-- Failure to Inject SLC C.6 Not N/A w/Boron for Low Water Credited Level DRHR--CNOPSPRYSIHE-- OP Fails to Initiate DW C.7 Not N/A Sprays for Debris Credited Cooling (SI Node)

DPCONTNNOPHUSUSFHE- OPERATOR C.9 5 hr Torus Pressure SUSPENDS FLOODING BASED ON ERRONOUS INDICATION DPCONT-NOPHUCLVNHE-- Operator Fails to Close C.9 5 hr Torus Pressure Wetwell Vent DSYSTM-NOP-PCFLDHE-- Operators Fail to C.9 5 hr RPV Level Implement Primary Containment Flooding DCNDSRCNOPMSVOPNHE- Op Fails to Open an C.12 Not N/A MSIV and/or Bypass Credited Valve DSBGT--NOPL2VENTHE-- OP FAILS TO VENT C.13 6 hr Torus Pressure CONTAINMENT (EOP-2 Step PC/P-10) POST COREDAMAGE I _I Page 2 of 2 A.

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RAI - PRA 34 DAEC RAI PRA 34 F&O 4-7 states that the logic under gate HPCI-MSL-FLD appears to be incorrect, as developed, both a HPCI valve failure and Level 8 Failure are required and even if level 8 occurs, the valve failure can result in overfeed continuing. The disposition states that the logic has been corrected and the requirement for Level 8 failure removed. A review of the fire PRA Computer Aided Fault Tree Analysis (CAFTA) model during the audit showed that this gate did not include a valve failure but did include other failures that are expected to be less important than a valve failure. Explain this modeling and its relevance to the FPRA. Also, explain the correction and the technical basis for it.

RESPONSE

The logic under gate HPCI-MSL-FLD models reactor vessel overfill given failure of the Level 8 trip system in conjunction with failure of the operator to trip HPCI. The fire induced spurious HPCI logic is modeled under gate FIRE-107. Gate FIRE-107 represents spurious HPCI system operation with failure of steam supply isolation valves to close.

The basis for F&O 4-7 was that gate FIRE-107 was incorrectly located under gate HPCI-MSL-FLD. The modeling was corrected by removing gate FIRE-107 from under gate HPCI-MSL-FLD so that the Level 8 trip failure logic and the fire induced spurious HPCI logic are no longer dependent. As identified in F&O 4-7, the technical basis for the correction is that vessel overfill due to fire induced spurious HPCI would not necessarily be mitigated by successful Level 8 trip logic actuation. The revised fire PRA model only credits closure of HPCI steam supply isolation valves to mitigate a fire induced spurious HPCI.

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RAI PRA 36 DAEC RAI PRA 36 Confirm that the potential impact on the FPRA results of (a) all known outstanding plant changes that would require a change to the FPRA model and (b) all planned plant changes that would significantly impact the FPRA model are included in the change-in-risk results.

RESPONSE

As discussed in Section 4.8.2 of the enclosure to License Amendment Request (ML11221A280) (LAR), the Fire PRA model represents the as-built, as-operated and maintained plant as it will be configured at the completion of the transition to NFPA 805.

The Fire PRA model includes credit for the planned implementation of the modifications identified in Attachment S of the LAR. Following installation of modifications and the attendant installation details, additional refinements surrounding the modification may need to be incorporated into the Fire PRA model. However, these changes are not expected to be significant. No other significant plant changes are outstanding with respect to their inclusion in the Fire PRA model.

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RAI PRA 37 DAEC RAI PRA 37 FPIE finding DA-C10-01A (SR DA-C10) indicates that no evidence of failure mode information is provided. The DAEC disposition states that every basic event is coded to identify that it is tested and that this implicitly implies that the testing procedures have been reviewed to determine that the test covers the failure mode. Provide the following:

a. Discuss where this coding is documented and how this indicates that the review of test procedures to verify the failure mode is properly tested.

b. It is noted that for two important component failure types (a circuit breaker fails to open or close and an air operated valve fails to open or close) the data notebook gives the same number of demands for the normally open fail to close and the normally closed fail to open failure modes. Verify that the testing program (procedures, practices and plant population) for these components gives the same number of demands regardless of failure mode and that this determination meets the standard or other guidance requirements.

RESPONSE

a. Coding for plant specific component failure data is discussed in Appendix C of the DAEC Component Failure Data Notebook (DAEC-PSA-DA-08 Volume I,)

which was prepared for the FPIE PRA. For each basic event, an operation code was developed as part of the basic event naming scheme that required determining whether:

i. components are operated continuously or routinely during normal operation, cold shutdown or refueling operations; ii. components are in standby and are not operated routinely except for testing, iii. valves are routinely operated, or iv. valves are operated less than once a cycle or are not in a test program.

Since the operation code was determined for each appropriate basic event, existence of the code implies that testing procedures were reviewed to determine the correct operation code.

b. The plant data that was used as the bases for the 4160 volt circuit breakers and air operated valves at DAEC for the January 1, 2003 to June 30, 2008 timeframe was reviewed. The number of equipment demands used for these components was determined to be appropriate.

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RAI - PRA 39 PRA Question # 39 The resolution to F&O 5-41 does not completely address the peer review finding to use upper bound cognitive error probabilities, rather than nominal or lower bound, where appropriate. Describe how the FPRA HRA results were reanalyzed for appropriateness and the results of this review. Specifically discuss whether human error probability (HEP) values were changed as a result of this review, including whether upper bound HEPs were used.

RESPONSE

It should be emphasized that the cognitive error assessments for the DAEC FPIE and FPRA operator actions are conducted with the use of two HRA methods: CBDTM +

ASEP. While the CBDTM allows the detailed consideration of the plant indication/operator and operator/procedure interfaces, it lacks the ability to reflect increases in cognitive error when the time available for recovery (Trec) is relatively short. In order to capture the details provided by the CBDTM and still address the increased Pc expected for time limited actions, the ASEP Pc component is added to the Pc results obtained via the CBDTM when the time available for recovery is less than 60 minutes. This represents a conservative practice. The detailed fire HEP quantification method described in Appendix B of draft NUREG-1921 does not require the sum of two Pc evaluation methods; it describes the use of CBDTM or HCR/ORE.

Short term FPIE actions (i.e., those actions that have time available for recovery values less than 60 minutes) that are translated into FPRA actions typically have an increased cognitive failure probability via the following changes.

" The time available for recovery is reduced by 10 minutes for CR-executed actions and by 20 minutes for ex-CR executed actions. A reduction in Trec leads to increases in both the ASEP curve results (even for the lower or nominal curves) and CBDTM contributions (through the loss of recovery factors such as the STA).

" The loss of CR indications and other fire event impacts are addressed by selecting revised CBDTM endstates that have increased Pc values.

The cognitive error increases resulting from the above approach are considered to provide best estimates. The ASEP curve selections for the FPIE actions were allowed to remain for their translation in to FPRA actions. The lower bound curve is selected for FPIE actions with which the operators are very familiar - generally common and relatively simple EOP or AOP based actions. Fire impacts are predominantly reflected through potential cue impacts and the reduction in Trec which are addressed by CBDTM as described above. Even when the FPIE ASEP curve selection is maintained for FPRA, the ASEP contribution will increase due to the reduced Trec value applied for fire initiators.

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RAI - PRA 39 If the DAEC Pc evaluations used the ASEP method alone and not the sum of CBDTM +

ASEP, the use of the upper bound curve for unusually complex actions might be valid.

However, choosing the upper bound curve substantially increases the Pc and is judged to result in an overly conservative and not a best estimate result for EOP or AOP based actions. Additionally, it is notable that the upper bound ASEP curve allows no credit for actions with Trec values less than 10 minutes. None of the fire response actions or the FPIE actions retained for the Fire PRA were considered complex enough to justify the use of the upper bound ASEP curve.

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RAI - PRA 40 DAEC RAI PRA 40 Clarify if protective relaying cables 1X3 and 1X4 were considered in multi-compartment analyses for the FPRA. If not, discuss why not.

RESPONSE

Protective relaying cables for 1X3 and 1X4 were considered in multi-compartment analyses for the FPRA. Multi-compartment interactions considered all potential failures in adjacent compartments and did not exclude cable routing data.

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RAI - PRA 41 PRA Question 41 Describe how the failure of fire dampers, including potential smoke propagation, were considered in the FPRA and the Fire Risk Evaluations for multicompartment analyses involving risk-significant fire areas.

RESPONSE

Fire damper failures were considered in the multi compartment analyses. The ventilation communication paths between fire zones were included in the multi compartment analysis. Per NUREG/CR-6850 Section 11, failure of fire dampers was treated by allowing the hot gas to migrate to the fire zone communicated with resulting in failure of all targets in the multi compartment combination. Per NUREG/CR-6850 Section 11.5.4.1, smoke impact on equipment was assumed to have minimal impact. If the multi compartment interaction and fire damper failure resulted in a CDF less than 1E-7/yr, the multi compartment interaction was screened from the analysis. Otherwise, the multi compartment interaction was included in the quantification. In the specific case of fire dampers, all multi compartment interactions met were screened.

Separately, the potential for smoke migration into the main control room from a fire in the essential switchgear rooms given failure of the fire dampers was assessed in Section 5.2.4 of the Fire PRA Fire Scenario Report (report 0493080001.003). The assessment considered the potential for smoke visibility impacts on the operators using conservative assumptions. Based on the assessment, the postulated scenarios were considered low risk and screened from the quantitative analysis.

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RAI - PRA 42 DAEC RAI PRA Question 42 Clarify if anticipated transient without scram (ATWS) sequences were screened out or not in the FPRA. Provide justification for doing so if ATWS sequences were screened out.

RESPONSE

ATWS sequences were not screened out in the FPRA. The potential for fire induced failure to scram exists in fire areas CB1 and RB1. In CB1, both normal and backup scram circuits are located in control room panels 1C05, 1C015, and 1C017 and fire induced failure to scram was treated. In RB1, normal and backup scram components and cables are sufficiently separated that a fire was not postulated to damage both scram methods.

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RAI - PRA 44 DAEC RAI PRA 44 F&O 3-9 for SR FSS-D6 is related to a transient heat release rate approach, and the disposition of this F&O is noted to be deferred. This approach is also identified as an unreviewed analysis method in Attachment V of the LAR. In response to a staff request for clarification on this method, sensitivity analysis results were documented in the letter from Peter Wells, NextEra Energy Duane Arnold, LLC, to the U.S. Nuclear Regulatory Commission, "Clarification of Information Contained in License Amendment Request (TSCR-128): Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) (TSCR-128) ," dated October 14, 2011, (ML1128702452). With respect to the delta risk calculations referred to in response to the question, it was noted that there was no impact on delta-risk for many fire areas since the increased risk contribution influenced both the variant and compliant case. Explain this statement.

RESPONSE

The transient heat release rate sensitivity analysis estimated the risk increase using the larger NUREG/CR-6850 recommended transient heat release rate. Given the larger heat release rate a larger zone of influence was assessed for each postulated transient fire. Given the larger zone of influence, a risk increase was calculated for most transient fires. For each fire scenario, the risk increase applied for the variant and compliant case. However, the delta risk calculations in the fire risk evaluations were only impacted if the larger zone of influence for the fire scenario included additional VFDRs. In these cases, the variant case risk would increase more than the compliant case risk.

In conclusion, the delta risk calculations in the fire risk calculations would only be affected if:

" the larger zone of influence included additional VFDRs, or

" the larger zone of influence included a VFDR in question that had not been affected in that particular fire scenario because of a smaller zone of influence.

In the scenarios reviewed at DAEC as part of the sensitivity analysis, no additional impacts were noted. Therefore, the risk increase for the variant and compliant case would be approximately the same.

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RAI - PRA 45 DAEC RAI PRA 45 For Finding HR-Al-01A and HR-A2-02A in Attachment U of the LAR,

1. Clarify the identified deviation from the standard [RAI-45-1], and

2. Specify which standard was deviated from [RAI 45-2].

3. Clarify what is meant by "the approached used" noted in the dispositions [RAI-45-3].

RESPONSE

FINDING HR-A1-01A Finding HR-Al-01A is against SR HR-Al [RAI 45-2]. HR-Al states: For equipment modeled in the PRA, IDENTIFY, through a review of procedures and practices, those test, inspection, and maintenance activities that require realignment of equipment outside its normal operationalor standby status.

FINDING HR-A2-02A Finding HR-A2-02A is against SR HR-A2 [RAI 45-2]. HR-A2 states: IDENTIFY, through a review of procedures and practices, those calibrationactivities that if performed incorrectlycan have an adverse impact on the automatic initiationof standby safety equipment.

The following discussion is applicable to both findings:

The identified deviation from the standard was that a comprehensive review of those maintenance and surveillance procedures defined in the two supporting requirements was not performed ("review of procedures and practices'); instead pre-initiator Human Failure Events were identified for inclusion in the PRA model using a systematic identification process based on systems analysis approaches

[RAI-45-1]. The approach used system operation and configuration aspects as well as a limited review of procedures [RAI-45-3].

The primary focus of the evaluation was to understand and review system configurations and operational aspects of the alignment and calibration of equipment post maintenance. This assured that pre-initiators would not be overlooked and would meet the intent of the supporting requirements. The peer review team agreed that this approach met the associated high level requirement.

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RAI - PRA 46 PRA Question # 46 F&O 2-15 states that the FPRA HRA does not consider the operator access route and therefore the impact of timing for excontrol room operator actions in response to a fire event. Explain how this was addressed for operator actions and/or RAs in the FPRA.

RESPONSE

Operator access routes were considered in assessing the manipulation time, execution stress and assigning the execution performing shaping factors in the FPRA HRA. The FPRA credits ex control room (CR) actions that are located in the Essential Switchgear Rooms, HPCI Room, Turbine Building, and Pumphouse. Each of these locations may be accessed using multiple routes.

In the EPRI HRA Calculator, the quantitative impact for negative PSFs is applied through the use of execution stress factors. For credited ex-CR executed actions within the DAEC FPRA, the execution stress factors were increased from their base values as given in the FPIE HEP assessment unless the time available was extensive and the fire was considered to be out and thus not continuing to cause late scenario complicating disturbances.

For example, DHPCI-CNOPOPENDRHE-- had an original FPIE execution stress of high (x5) which was increased to "fire stress" (xl 0) because it has a relatively short time line and the Internal Events HFE calculation had already assigned negative PSFs because of a hot environment and emergency lighting. This is consistent with draft NUREG-1 921 guidance (Table B-1 8) which states that "A fire stress level should be used if more than 2 execution PSFs are negative."

There is a 10 minute increase assigned for ex-CR manipulation times for the FPRA to account for potential access delays introduced by the fire. NUREG/CR-6850 or draft NUREG-1921 do not specify an access delay but states that increases in manipulation time due to additional travel time for local actions should be considered. Based on the guidance and operator interviews, an increase in manipulation time of 10 minutes was used.

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RAI - PRA 47 PRA Question # 47 According to the Fire Model Development Report, some equipment in the FPRA is credited by exclusion in cases when cables could not be traced, and includes systems and signals which are containment-related. Discuss the extent (e.g., in terms of fire zones) that these containment-related systems and signals are credited by exclusion in the FPRA. For fire scenarios which involve potential impact on these containment-related systems and signals, describe what failure modes were considered when using credit for exclusion for containment isolation valves (CIVs), e.g., multiple spurious operation and containment isolation signals including low reactor pressure vessel water level or high drywell pressure.

RESPONSE

Cables were not traced for the containment isolation signal. The containment isolation signal was not credited by exclusion in terms of fire zones. Instead, the containment isolation signal was credited in terms of fire scenarios. The containment isolation signal was credited by exclusion in fire scenarios where only the ignition source was damaged.

Therefore, fire scenarios that result in damage to cables that could potentially impact the containment isolation signal did not credit the containment isolation signal by exclusion.

Cables were traced for containment isolation valves. As such, the functional failure and multiple spurious were considered in modeling containment isolation valves.

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RAI - Probabilistic Risk Assessment 48 DAEC RAI PRA 48 Monitoring suppression pool level is mentioned in the LAR section "Process Monitoring" for alternate shutdown capability instrumentation, but is not listed in the Final Safety Analysis Report (FSAR) section 7.4, "Systems Required for Safe Shutdown," as being required for safe shutdown. Is process monitoring of suppression pool level necessary for the alternate shutdown capability? If so, is suppression pool level monitoring modeled in the Fire PRA?

RESPONSE

The Alternate Shutdown Capability System (ASCS) is described in the Updated Final Safety Analysis Report (UFSAR) section 7.4.2. Panel locations, controls and indications for the ASCS are listed in UFSAR tables 7.4-1, 7.4-2 and 7.4-3. Process monitoring required for the ASCS was selected based on guidance similar to that provided in NRC Information Notice 84-09. Process monitoring of suppression (or torus) pool level is required for the ASCS.

Monitoring suppression pool level is not modeled in the Fire PRA. The Fire PRA models the instrumentation identified in the procedure for the applicable operator action.

AOP 915, Shutdown Outside Control Room, does not identify monitoring suppression pool level for the operator actions required for alternate shutdown capability. As such, the Fire PRA does not credit suppression pool level instrumentation for any of the operator actions.

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RAI - PRA 49 DAEC RAI PRA 49 The following FAQs were not included in Table H-i, "NEI 04-02 FAQs Utilized in the LAR Submittal": FAQ 08-0047, Spurious Operation Probability, and FAQ 08-0048, Fire Ignition Frequency. Discuss the guidance and methodology used in the LAR submittal in place of these FAQs.

RESPONSE

The guidance and methodology in NUREG/CR-6850 was used in place of FAQ 08-0047, Spurious Operation Probability, and FAQ 08-0048, Fire Ignition Frequency. The spurious operation probability was calculated using NUREG/CR-6850 Section 10. The fire ignition frequencies from NUREG/CR-6850 Section 6 were used. Use of NUREG/CR-6850 for the LAR submittal in place of the FAQs results in a conservative estimate of risk.

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RAI - Probabilistic Risk Assessment 52 DAEC RAI PRA 52 Confirm or clarify that the RAs identified in the LAR are new and are not previously approved RAs.

RESPONSE

NextEra Energy Duane Arnold confirms that the Recovery Actions identified in the the enclosure to the License Amendment Request (ML11221A280) (LAR) are new and are not previously approved Recovery Actions. All of the post-transition Recovery Actions identified in the LAR were among the pre-transition operator manual actions specified in procedures that implemented the alternative shutdown capability per Section III.L in Appendix R to 10 CFR 50. In accordance with guidance presented in FAQ 07-0030, the calculated change in CDF/LERF reported for Fire Area CB1 is a surrogate for the additional risk presented by use of these Recovery Actions to maintain a reasonable balance among defense-in-depth elements for the performance-based approach.

Although RG 1.205, Rev. 1, Regulatory Position 2.2.4.1 allows for treatment of "previously approved" Recovery Actions and risk offsets to determine the acceptability of the risk increase associated with transition, NextEra Energy Duane Arnold is not relying on this approach.

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RAI - Probabilistic Risk Assessment 53 DAEC RAI PRA 53 Some VFDRs in fire area CB1 are the same or similar to those in other fire areas, yet no RA is assigned to them. Some examples are VFDR-SSA-CB1-12 and VFDR-SSA-RB1-03 in fire area RB1, as well as VFDR-SSA-CB11 and VFDR-SSA-RB1-04 in fire area RB1. Explain why RAs are not assigned to such VFDRs in fire areas other than CB1.

RESPONSE

At DAEC, Fire Area CB1 is unique from the perspective of equipment and cables located in the fire area, fire protection equipment provided, amount and type of combustibles, and number and type of ignition sources. The area assessment for Fire Area CB1 includes fire scenarios that may result in control room abandonment and subsequent activation of the Alternate Shutdown Capability System (ASCS). Activation of the ASCS results in control for a single division of ECCS equipment and associated supporting functions. For this reason, DAEC considers transfer of control to the ASCS to be a challenge to defense-in-depth. Therefore, the DAEC fire risk evaluation for Fire Area CB1 determined that all transfers to Primary Control Stations and all identified Recovery Actions would be required to maintain a reasonable balance among defense-in-depth elements for the performance-based approach.

No fire scenarios involving similar VFDRs in other fire areas result in the need for control room abandonment and activation of the ASCS. With control maintained in the control room, there is a greater range and depth of equipment and systems available.

DAEC fire risk evaluations for those fire areas determined that the applicable risk, defense-in-depth, and safety margin criteria were satisfied without resorting to recovery actions.

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RAI - PRA 54 DAEC RAI PRA 54 Attachment G of the LAR, "Recovery Actions Transition," Table G-1, "Recovery Actions and Activities Occurring at the Primary Control Station(s)," identifies six RAs. For these RAs:

a. Explain why there are RAs identified with the 1A4 switchgear in the LAR, but not with the 1A3 switchgear.

b. The RA for "B EDG" states "If Bus 1A4 is de-energized, Then start diesel generator 1G21 locally .... " Explain why the diesel generator does not start automatically if bus 1A4 is de-energized.

RESPONSE

a. The referenced recovery actions are required to maintain a reasonable balance among defense-in-depth elements for the performance-based approach employed in Fire Area CB-1. The recovery actions support transfer of control to the Alternate Shutdown Capability System (ASCS) when control room abandonment is required. The design of the ASCS utilizes a single division of equipment and support equipment required to meet the performance goals of 10 CFR 50 Appendix R Section Ill.L; this also satisfies the Nuclear Safety Performance Criteria of NFPA 805. The AC electrical power support requirement is met by use of the Division 2 (1A4) switchgear. The Division 1 (1A3) switchgear does not supply power for any of the equipment controlled by the ASCS and is not recovered for cases requiring control room abandonment. Thus, no associated recovery actions are specified for the 1A3 switchgear.

b. The ASCS was designed to meet the requirements of 10 CFR 50 Appendix R Section III.L. One of the requirements of the ASCS is to accommodate postfire conditions where offsite power is available and where offsite power is not available for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . If offsite power is available and Bus 1A4 is energized, starting diesel generator 1G21 (automatically or locally) would not be required.

The recovery action accommodates fire scenarios where offsite power to Bus 1A4 is interrupted. Automatic start of diesel generator 1G21 may be disabled as a result of potential circuit damage in the fire area of concern. In addition, following transfer of control to the ASCS, the automatic start of diesel generator 1G21 on loss of offsite power will no longer function because that portion of the circuit is isolated to prevent fire damage from adversely affecting the control of 1G21.

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RAI - PRA 55 PRA Question # 55 In Attachment W of the LAR, the discussion under "Recovery Actions" notes that an HRA was performed for RAs required in the short term. Clarify what is "short term" and why an HRA is not needed for actions required after "short term." Clarify whether this refers to only fire scenarios involving main control room abandonment and how it relates to the RAs provided in Table G-1.

RESPONSE

The discussion in Attachment W of the enclosure to the License Amendment Request (ML11221A280) (LAR) applied to recovery actions identified in Table G-1 of the enclosure to the LAR. These are operator actions required to establish alternate shutdown capability for fire scenarios involving main control room abandonment. The statement does not apply to other fire scenarios because other fire areas did not have recovery actions.

Recovery actions to establish alternate shutdown capability were treated in the FPRA as being completely dependent. For example, if the action to locally start the diesel generator failed then alternate shutdown was considered unsuccessful. Based on industry guidance, actions required in the short term would be most likely impacted by fire events. While there is no way to predict the length of time it takes to terminate the full spectrum of fire events, NUREG/CR-6850 uses a time of 60 minutes as an estimate for completing the early response tasks and for getting additional personnel on location to address any remaining fire issues. After 60 minutes, the fire is assumed to be "out" and not continuing to cause delayed spurious activity or other complicating disturbances. In addition, draft NUREG-1921 states: "Long term actions are those that are not required during the early stage (e.g., first hour) of a fire event and are not expected to be performed until approximately one hour after the fire initiation and trip of the plant or until the fire is out. Thus, short term actions are those required within the first hour of a trip." This is considered to be reasonable and has been adopted in the FPRA to separate actions that may be impacted by fire mitigation activities and those that would not be impacted in a significant way.

Given the recovery actions to establish alternate shutdown were treated as being completely dependent, an HRA was not required for those actions not required in the short term given the short term actions are most likely impacted by fire events.

Page 1 of I Rev A.A. Page 1 of 1

RAI - PRA 56 DAEC RAI PRA 56 Clarify if the operator actions occurring at the primary control stations given in LAR Table G-1 are modeled in the FPRA.

RESPONSE

Operator actions occurring at the primary control stations given in Table G-1 of the enclosure to the License Amendment Request (ML11221A280) (LAR) are modeled in the FPRA. Operator actions occurring at the primary control stations are modeled as being completely dependent on the actions that do not occur at the primary control stations. That is, failure to complete the actions executed outside of the primary control stations, that are necessary for the actions at the primary control station to be successful (e.g., providing power or support to the necessary components controlled from the primary control station), results in the failure of the primary control station actions. Operator actions that do not occur at the primary control stations require multiple procedure steps at different locations. Operator actions at the primary control stations occur at a single panel, require simple steps using a hand switch, and are performed during the same time period as actions not at the primary control stations.

Page 1 of I Rev A. Page 1 of 1

RAI - PRA 57 DAEC RAI PRA 57 F&Os 4-12 and 4-15 contain specific peer review comments which do not appear to have been addressed or their evaluation documented. The dispositions of these F&Os mention that the Fire Model Development (FMD) report was updated to include a detailed LERF model review; however, very little technical justification was provided in the FMD report. The document PSA-L2-15, Appendix C, contains a discussion of the Level 2 modeling. The discussion in Appendix C does not appear to have been updated for fire-related impacts, but appears to be related to Internal Events PRA random failures. Address all the specific comments from the peer review for these F&Os and provide detailed technical justification for the conclusions in the disposition of these F&Os. Include a discussion on how Appendix C of PSA-L2-15 is impacted by fire scenarios.

RESPONSE

The DAEC Level 1 PRA model transfers directly to the Level 2 PRA model. As such, fire impacts on system components and support systems modeled in the Level 1 are transferred to the Level 2 model. Therefore, the scope of review for the Level 2 PRA model for potential fire impacts included the Level 2 PRA model features not included in the Level 1 PRA model. Appendix C of DAEC-PSA-L2-15 provides a discussion of the fault tree logic for each Level 2 event tree node and identifies the basic events in the model.

F&O 4-12 specifically discusses the potential for fire impacts on Level 2 modeling features associated with AC power recovery, containment isolation, operator depressurizing the RPV, core melt progression, combustible gas venting, containment integrity, containment flooding, and venting. The modeling of each of these features is further discussed.

AC power recovery (Section C.1) is not credited in the FPRA for fire induced failures.

The conditional probabilities associated with recovery of AC power in the Level 2 timeframe are set to 1.0.

Containment isolation (Section C.2) fire impacts are included in the FPRA. Cables for the containment isolation valves were selected. As such, fire impacts on cables for containment isolation valves will result in functional failure or spurious operation. Cables for the containment isolation signal were not selected. The containment isolation signal was assumed failed for scenarios with cable damage. Additional features modeled in the containment isolation logic consist of check valve failures, pipe breaks, and preexisting conditions which would not be affected by fire events.

RPV depressurization (Section C.3) in the Level 2 timeframe includes fire impacts on the SRVs from the Level 1 model. In addition, an operator action to depressurize in the Level 2 timeframe is included. However, this action would be required in the two hour timeframe and RPV level instrumentation has been verified available. Therefore, fire impacts on these actions are not postulated. Other features included in the model logic Rev A. Page 1 of 31

RAI - PRA 57 include potential impact from environmental conditions or system failure from containment breach or water hammer. These phenomenological conditions would not be impacted by fire events.

Core melt progression (Section A.4)arrested in vessel includes fire impacts on the injection systems from the Level 1 model. Several operator actions are modeled; however, these actions would be required in the 1-3 hour timeframe and RPV level instrumentation has been verified available. Therefore, fire impacts on these actions are not postulated. In addition, the fault tree logic does give credit to injection system recovery in the timeframe. Given potential fire impacts on cables, the recovery is not credited in the FPRA and the probability is set to 1.0.

Combustible gas venting (Section C.5) is modeled in the event tree node as not being required because containment is inerted or as being failed due to loss of AC power. The conditional probability that containment is inerted is not impacted by fire events. Fire impacts on AC power are included in the Level 1 logic transfers to Level 2.

Containment intact (Section C.6) modeling features are associated with containment behavior at the time molten core penetrates the RPV or during the time core melt is arrested in vessel. Fire impacts associated with the operator action that would occur in the 1-3 hour timeframe to restore injection during the time between the time when control rods begin to melt and fuel rods begin to melt are not postulated. The probability is assigned based on the small timeframe for the coincidental occurrence. The model includes numerous phenomenological conditions resulting in containment failure that would not be impacted by fire events. Vacuum breaker failures are also included; however, these are components that do not have cables associated with them and fire impacts are not postulated.

Containment flooding (Section C.9) modeling includes several operator actions. These operator actions would be required in the five hour timeframe and instrument cues are verified available. Therefore, fire impacts on these actions are not postulated. Additional features modeled include conditional probabilities associated with containment failure precluding successful flooding which are not impacted by fire events. Dependency on vent valve operation is included in the model logic. Fire impacts on vent valves are included in the Level 1 model and transferred to the Level 2 model.

Wetwell and Drywell venting (Section C.13 and C.17, respectively) fire impacts on the containment vent valves are included in the Level 1 model and transferred to the Level 2 model. The modeling includes additional phenomenological conditions that could prevent successful venting that would not be impacted by fire events.

F&O 4-15 specifically discusses fire impacts on the containment event tree node OP (operator depressurizing the RPV) including adverse reactor building conditions, instrumentation impacts, and ADS initiation. A conditional probability is modeled associated with the potential impact of containment failure or isolation failure on SRV logic and power supplies located in the reactor building. Fire impacts on containment failure and isolation failure are identified in the review of each Section of Appendix C.

Rev A. Page 2 of 31

RAI - PRA 57 Functional failures and spurious operations of containment isolation valves from fire impacts are included in the FPRA model. Fire impacts on SRV logic and power supplies due to fires in the reactor building result in failure of the SRVs in the Level 1 model and are transferred to the Level 2. The HRA for the Level 1 and Level 2 operator action to manually depressurize relies on RPV level as the instrument cue which has been verified available in each fire area. The Level 1 model logic gate ADS-INITIATE is transferred to the Level 2 model. Fire impacts on the operator actions to manually initiate ADS were included in the Level 1 HRA. Therefore, fire impacts on ADS initiation were included in the Level 2 model. Event DADS---NPHSRVSTKCE-models the conditional probability that SRVs may not stick open due to large number of cycles or high internal temperatures. The conditional probability applied is 0.55. In addition, fire impacts on cables could cause the SRVs to spurious open (assigned a 0.6 probability).

Therefore, the probability of SRVs not spurious opening would be 0.4 (1-0.6). As such, the failure probability assigned in the Level 2 model bounds the fire impacts and the model was not changed.

Each section of Appendix C of DAEC-PSA-L2 was reviewed and the potential for fire impacts on each Level 2 basic event was identified and the applicable changes made to the FPRA model. The following table provides a listing of each event and identifies the potential for fire impacts and the required model change.

The table will be included in the Fire Model Development report.

Page 3 of 31 Rev A. Page 3 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONTNFCV4305--RC-- AO-4305 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV3704--FC-- AO-3704 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV3704--RC-- AO-3704 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV3705--FC-- AO-3705 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV3705--RC-- AO-3705 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV3728--FC-- AO-3728 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV3728--RC-- AO-3728 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV3729--FC-- AO-3729 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV3729--RC-- AO-3729 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4300--FC-- AO-4300 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4300--RC-- AO-4300 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4301--FC-- AO-4301 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4301--RC-- AO-4301 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4302--FC-- AO-4302 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

Rev A. Page 4 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONTNFCV4302--RC-- AO-4302 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4303--FC-- AO-4303 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4303--RC-- AO-4303 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4304--RC-- AO-4304 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4306--FC-- AO-4306 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4306--RC-- AO-4306 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4307--FC-- AO-4307 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4307--RC-- AO-4307 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4308--FC-- AO-4308 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4308--RC-- AO-4308 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4311--FC-- AO-4311 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4311--RC-- AO-4311 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED DPCONTNFCV4312--FC-- AO-4312 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

DPCONTNFCV4312--RC-- AO-4312 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED I DPCONTNFCV4313--FC-- AO-4313 FAILS TO CLOSE C.2 Y Fire impacts included in the FPRA.

(FC)

Rev A. Page 5 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONTNFCV4313--RC-- AO-4313 FAILS TO REMAIN C.2 Y Fire impacts included in the FPRA.

CLOSED D-----

CFCH270011 FC-- CHECK VALVE 27-0011 FAILS C.2 N Check valve failure not assessed for to close ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DRHR--CFCH190149FC-- CHECK VALVE V1 9-0149 C.2 N Check valve failure not assessed for FAILS to close ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DRHR--CFCH20082-FC-- CHECK VALVE V20-082 FAILS C.2 N Check valve failure not assessed for to close ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DCSPRYCFCH210072FC-- CHECK VALVE V21-0072 C.2 N Check valve failure not assessed for FAILS to close ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DCSPRYCFCH210073FC-- CHECK VALVE V21-0073 C.2 N Check valve failure not assessed for FAILS to close ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DHPCI-CFCH230049FC-- CHECK VALVE V23-0049 C.2 N Check valve failure not assessed for FAILS to close ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DRCIC-CFCH250036FC-- CHECK VALVE V250036 FT C.2 N Check valve failure not assessed for CLOSE ON DEMAND fire impacts based on NUREG/CR-6850 guidance.

DPCONT-NICGNFAILTF-- Containment Isolation Signal C.2 Y Containment Isolation Signal Fails assumed failed in FPRA.

DCSPRYCSPPSBREAKRP-- CORE SPRAY SYSTEM PIPE C.2 N Conditional probability of pipe RUPTURE rupture. No fire impacts associated with pipe rupture.

DCSPRYCFMOSOPEN-CE-- CS ISOLATION VALVE C.2 N Conditional probability of preexisting INITIALLY OPEN condition not impacted by fire events.

Rev A. Page 6 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONTNFCV4329--RC-- CV 4329 FAILS TO REMAIN C.2 N Check valve failure not assessed for CLOSED fire impacts based on NUREG/CR-6850 guidance.

DPCONTNFCV4330--RC-- CV 4330 FAILS TO REMAIN C.2 N Check valve failure not assessed for CLOSED fire impacts based on NUREG/CR-6850 guidance.

DSBGT-CFMODWPG90CE-- DW PURGE LINE OPEN C.2 N Conditional probability of preexisting DURING NORMAL condition not impacted by fire OPERATION events.

DHPCI-CFMOHPIOPNCE-- HPCI ISOLATION VALVE C.2 N Conditional probability of preexisting INITIALLY OPEN condition not impacted by fire events.

D------- PPHPIBRERP-- HPCI SYSTEM PIPE RUPTURE C.2 N Conditional probability of pipe rupture. No fire impacts associated with pipe rupture.

DPCONT-NPHWFLR24CE-- Isolation Valves Auto Open to C.2 N Conditional probability that drainage Permit Drainage is required. Fire impacts on signal included with valve fire impacts.

D------- MOWEQP24CE-- ISOLATION VALVES AUTO C.2 N Conditional probability that drainage OPEN TO PERMIT DRAINAGE is required. Fire impacts on signal included with valve fire impacts.

DPCONT-NPH08-02-TF-- Large Pre-existing Failure C.2 N Conditional probability of preexisting condition not impacted by fire events.

DPCONTNNOPDWSPL9CE- Line Open During Normal C.2 N Conditional probability of preexisting Operation condition not impacted by fire events.

DSBGT-CFMOWWSP90CE- LINE OPEN DURING NORMAL C.2 N Conditional probability of preexisting OPERATION condition not impacted by fire events.

Page 7of31 Rev A. Page 7 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts D------- PPISDRAIRP-- MAIN STEAM LINE DRAIN C.2 N Conditional probability of pipe break.

PIPE BREAK OUTSIDE No fire impacts associated with pipe CONTAINMENT break.

DRHR--CFMO1902--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

MO1902 FAILS TO CLOSE ON DEMAND DRHR--CFMO1903--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M01903 FAILS TO CLOSE ON DEMAND DRHR--CFMO1905--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

MO1905 FAILS TO CLOSE ON DEMAND DRHR--CFMO1908--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

MO1908 FAILS TO CLOSE ON DEMAND DRHR--CFMO1909--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M01909 FAILS TO CLOSE ON DEMAND DRHR--CFMO1913--FC-- Motor Operated Valve M01913 C.2 Y Fire impacts included in the FPRA.

Fails TO CLOSE on Demand DRHR--CFMO1921--FC-- Motor Operated Valve MO1921 C.2 Y Fire impacts included in the FPRA.

Fails TO CLOSE on Demand DRHR--CFMO1932--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M01932 FAILS TO CLOSE ON DEMAND DRHR--CFMO1933--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

MO1933 FAILS TO CLOSE ON DEMAND DRHR--CFMO1934--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

MO1934 FAILS TO CLOSE ON DEMAND Rev A. Page 8 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DRHR--CFMO1935--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M01935 FAILS TO CLOSE ON DEMAND DRHR--CFMO1989--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

MO1989 FAILS TO CLOSE ON DEMAND DRHR--CFMO2000--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02000 FAILS TO CLOSE ON DEMAND DRHR--CFMO2001--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02001 FAILS TO CLOSE ON DEMAND DRHR--CFMO2003--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02003 FAILS TO CLOSE ON DEMAND DRHR--CFMO2005--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02005 FAILS TO CLOSE ON DEMAND DRHR--CFMO2006--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02006 FAILS TO CLOSE DEMAND DRHR--CFMO2007--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02007 FAILS TO CLOSE ON DEMAND DRHR--CFMO2009--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02009 FAILS TO CLOSE ON DEMAND DRHR--CFM02012--FC-- Motor Operated Valve M02012 C.2 Y Fire impacts included in the FPRA.

Fails TO CLOSE on Demand DRHR--CFMO2015--FC-- Motor Operated Valve M02015 C.2 Y Fire impacts included in the FPRA.

Fails TO CLOSE on Demand Rev A. Page 9 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DRHR--CFM02069--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02069 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2100-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02100 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2104-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02104 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2112-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02112 FAILS to close ON DEMAND DCSPRYCFMOO2117-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02117 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2120-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02120 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2124-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02124 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2132-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02132 FAILS to close ON DEMAND DCSPRYCFMOO2137-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02137 FAILS TO CLOSE ON DEMAND DCSPRYCFMOO2146-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02146 FAILS TO CLOSE ON DEMAND Page 10 of 31 Rev A.

Rev A. Page 10 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DCSPRYCFMOO2147-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02147 FAILS TO CLOSE ON DEMAND DHPCI-CFMO2238--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02238 FAILS TO CLOSE ON DEMAND DHPCI-CFMO2239--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02239 FAILS TO CLOSE ON DEMAND DHPCI-CFMO229OA-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02290A FAILS to close ON DEMAND DHPCI-CFMO2290B-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02290B FAILS to close ON DEMAND DHPCI-CFMO2312--FC-- Motor Operated Valve M02312 C.2 Y Fire impacts included in the FPRA.

Fails to Close on Demand DHPCI-CFMO2318--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02318 FAILS to CLOSE ON DEMAND DHPCI-CFMO2322--FC-- Motor Operated Valve M02322 C.2 Y Fire impacts included in the FPRA.

Fails to Close on Demand DRCIC-CFMOMO2400FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02400 FAILS TO CLOSE ON DEMAND DRCIC-CFMOMO2401 FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02401 FAILS TO CLOSE ON DEMAND DRCIC-CFMOMO2512FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02512 FAILS TO CLOSE ON DEMAND Rev A. Page 11 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DRCIC-CFMOMO2516FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02516 FAILS TO CLOSE ON DEMAND DRCIC-CFMOMO2517FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02517 FAILS TO CLOSE ON DEMAND DRWCU-CFMO2700--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02700 FAILS TO CLOSE ON DEMAND DRWCU-CFMO2701--FC-- Motor Operated Valve M02701 C.2 Y Fire impacts included in the FPRA.

Fails To Close on Demand DRWCU-CFM02740--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M02740 FAILS TO CLOSE ON DEMAND DPCONTNFMO4423--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M04423 FAILS TO CLOSE ON DEMAND DPCONTNFMO4424--FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M04424 FAILS TO CLOSE ON DEMAND DRBCCWCFMO4841A-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M04841A FAILS TO CLOSE ON DEMAND DRBCCWCFMO4841B-FC-- MOTOR OPERATED VALVE C.2 Y Fire impacts included in the FPRA.

M04841 B FAILS TO CLOSE ON DEMAND DRCIC-CFMOMO2510FC-- MTR-OPERATED VLV M02510 C.2 Y Fire impacts included in the FPRA.

FAILS TO CLOSE ON DEMAND DRBCCWCRPPCWPIPERP- RBCCW SYSTEM PIPE BREAK C.2 N Conditional probability of pipe break.

OUTSIDE CONTAINMENT No fire impacts associated with pipe

____________________________________________________break.

Rev A. Page 12 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DRCIC-CFMORCIOPNCE-- RCIC ISOLATION VALVE C.2 N Conditional probability of preexisting INITIALLY OPEN condition not impacted by fire events.

DRCIC-CSPPRCIBRERP-- RCIC SYSTEM PIPE RUPTURE C.2 N Conditional probability of pipe rupture. No fire impacts associated with pipe rupture.

DRECRCCRMPPSEAL-LK-- REACTOR RECIRC. PUMP C.2 N Conditional probability of heat SEAL HX LEAK exchanger leak. No fire impacts associated with heat exchanger leak.

DRHR--CFMOROPE--CE-- RHR ISOLATION VALVE C.2 N Conditional probability of preexisting INITIALLY OPEN condition not impacted by fire events.

DRHR--CN--RHRBRKRP-- RHR SYSTEM PIPE RUPTURE C.2 N Conditional probability of pipe rupture. No fire impacts associated with pipe rupture.

DFPOOLCRPPUPIPE-RP-- RWCU SYSTEM PIPE C.2 N Conditional probability of pipe RUPTURE rupture. No fire impacts associated with pipe rupture.

DRHR--CFMO1908--CE-- SDC ISOLATION VALVE C.2 N Conditional probability of preexisting M01908 INITIALLY OPEN condition not impacted by fire events.

DPCONT-NPHCIVS24TF-- Small Pre-existing Failure C.2 N Conditional probability of preexisting condition not impacted by fire events.

D------- MOLVOPENCE-- STEAM DRAIN VALVES C.2 N Conditional probability of preexisting INITIALLY OPEN condition not impacted by fire events.

D--------- 09-04-TF-- TORUS HATCHES NOT C.2 N Conditional probability of preexisting CLOSED AND SEALED condition not impacted by fire events.

Rev A. Page 13 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DSBGT-CFMOWWPG90CE- WW PURGE LINE OPEN C.2 N Conditional probability of preexisting DURING NORMAL condition not impacted by fire OPERATION events.

DADS--NNPHDWENVECE-- ADVERSE DW ENVIRONMENT C.3 N Conditional probability that adverse AFFECTS SRVs DURING environmental conditions fail SRVs.

SEVERE ACCIDENT (CLASS No fire impacts associated with the 1/111) phenomenon.

DLLSRVNNPHDWENV-CE-- ADVERSE DW ENVIRONMENT C.3 N Conditional probability that adverse AFFECTS SRVs DURING environmental conditions fail SRVs.

SEVERE ACCIDENT (CLASS No fire impacts associated with the II/IV) phenomenon.

DADS---NOPXENVOUCE-- Adverse Reactor Building C.3 N Conditional probability that adverse Conditions Cause Failure environmental conditions fail SRVs.

No fire impacts associated with the phenomenon.

DADS--NNPHALTDP-CE-- Alternate Depress Methods Not C.3 N Event has a 1.0 failure probability in Credited the PRA.

DADS--NNPHHITEMPCE-- High Primary System C.3 N Conditional probability that primary Temperature Does Not Cause system fails given high primary Failure system temperature. No fire impacts associated with the phenomenon.

DADS--ANOP-LVL2-HE-- Operator Fails to Depress C.3 N Operator action modeled in the 2 Before RPV Fails Given hour time frame. Instrument cues Operator Failed in Level 1 have been verified available. Based on NUREG/CR-6850 and draft NUREG-1 921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

Page 14 of 31 Rev A.

Rev A. Page 14 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DADS--NNPHSRVSTKCE-- SRVs Do Not Stick Open C.3 N Conditional probability that SRVs stick open due to number of cycles.

Spurious operation of two or more SRVs would have a similar impact on the model. However, the conditional probability assigned in the Level 2 model bounds the spurious operation probability.

DADS---NOPNFLADOCE-- Structural Breach in C.3 N Conditional probability that adverse Containment Causes ADS environmental conditions fail SRVs.

Failure No fire impacts associated with the phenomenon.

DADS--NNPHWTRHMRCE-- Water Hammer Does Not Cause C.3 N Conditional probability that water Failure of Mechanical System hammer results in system failure. No fire impacts associated with the phenomenon.

DSWYRDENOPRX###CE-- Recovery of AC Power C.3 Y Recovery of fire induced failure of offsite power is not credited in the FPRA given potential cable damage impacts.

DCRD--CN--CRDINJTF-- CRD INJECTION INADEQUATE C.4 N CRD is not credited in the PRA to prevent core melt progression.

DFPROTDN---INJECTF-- FIRE PROTECTION C.4 N Fire protection alignment is not INJECTION ALIGNMENT NOT credited in the PRA.

CREDITED IN L2 PRA DFEED-CN--H-NOFWTF-- MAIN FEEDWATER SYSTEM C.4 N FW is not credited in this timeframe.

UNAVAILABLE DHPCI-CSTPHPCIFLTF-- MANUAL SWITCH -HP FAILS C.4 N HPCI not credited in this timeframe ON DEMAND given previous failure or that the RPV is depressurized.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DSYSTM-NPHXRSTOOTF-- Melting Sufficient To Prevent C.4 N Conditional probability that despite Effective Heat Transfer recovery of coolant injection that molten debris may achieve a configuration that is no longer coolable. No fire impacts on conditional probability.

DSYSTM-NOPALTINXHE-- Operator Fails to Align Alternate C.4 N Operator action modeled in the 1-3 Injection Sources in Level 2 hour time frame. Instrument cues have been verified available. Based on NUREG/CR-6850 and draft NUREG-1 921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DSYSTM-NOP-RECVRRX-- Operator Fails to Recover C.4 Y Recovery of injection systems before Injection Before RPV Melt RPV melt is assigned a 0.9 probability. However, when considering fire damage to cables, system recovery may not be likely.

For fire, the recovery is set to 1.0.

DSYSTM-NOPTERMINHE-- Operator Intervenes and C.4 N Operator action modeled in the 1-3 Terminates Injection hour time frame. Instrument cues have been verified available. Based on NUREG/CR-6850 and draft NUREG-1921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DRCIC-CN---RCICFTF-- RCIC SYSTEM INADEQUATE C.4 N RCIC is not credited in the PRA to prevent core melt progression.

Page 16 of 31 Rev A. Page 16 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DSWYRDENOPRX###CE-- Recovery of AC Power C.4 Y Recovery of fire induced failure of offsite power is not credited in the FPRA given potential cable damage impacts.

DWELLWDN---INJECTF-- WELL WATER UNAVAILABLE C.4 N Well Water cross tie is not credited FOR INJECTION TO REACTOR in the PRA to prevent core melt progression.

DCACS-NPHGV1 FLCE-- Failure of Combustible Gas C.5 N Conditional probability that the Venting containment is inerted. No fire impact.

DCACS-NPHGV2FLCE-- Failure of Combustible Gas C.5 N Event has a 1.0 failure probability in Venting (No AC) the PRA.

DPCONT-NPHSTEAMCE-- Containment Not Steam Inerted C.6 N Conditional probability that sufficient steam is in containment to prevent combustion. No fire impacts associated with the phenomenon.

DPCONT-NPHMELTSCE-- Control Rods Melt Prior to Fuel C.6 N Conditional probability of the control Rods rods melting prior to the fuel rods.

No fire impacts associated with the phenomenon.

DPCONT-NPHDEIMPCE-- Debris Impingement on C.6 N Conditional probability of debris Containment Causes Failure impingement at the time of vessel failure. No fire impacts associated with the phenomenon.

DPCONT-NPHCONHTCE-- Direct Containment Heating C.6 N Conditional probability of direct Fails Containment containment heating at the time of vessel failure. No fire impacts associated with the phenomenon.

Page 17 of 31 Rev A. Page 17 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONT-NPHCONHICE-- Direct Containment Heating C.6 N Conditional probability of direct Fails Containment containment heating at the time of vessel failure. No fire impacts associated with the phenomenon.

DPCONT-NPHDWNCMCE-- Downcomer Vent Pipes Fail C.6 N Conditional probability of downcomer rupture. No fire impacts associated with the phenomenon.

DPCONT-NPHEXEXPCE-- Ex-Vessel Steam Explosion C.6 N Conditional probability of Ex-vessel steam explosion as the core slumps from the bottom of the vessel to the drywell. No fire impacts associated with the phenomenon.

DPCONTNNOPPHSLC-HE-- Failure to Inject SLC w/Boron for C.6 N The operator action is not credited in Low Water Level the PRA.

DPCONT-NPHINTEGCE-- Fuel Rod Integrity is Maintained C.6 N Conditional probability of fuel rod During the Reflood integrity is maintained during vessel reflood. No fire impacts associated with the phenomenon.

DPCONT-NPHHPBL1CE-- High Pressure Blowdown C.6 N Conditional probability of Overwhelms Vapor Suppression containment pressure spike at the time of vessel failure. No fire impacts associated with the phenomenon.

DPCONT-NPHH2GLBCE-- Hydrogen Deflagration Occurs C.6 N Conditional probability of hydrogen Globally deflagration when sufficient combustible gas concentrations exist in containment. No fire impacts associated with the phenomenon.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONT-NPHLOCA-CE-- Induced ISLOCA C.6 N Conditional probability of induced ISLOCA during core melt progression. Fire induced valve failures are treated in the Level 1 ISLOCA sequences. Conditional probability based on interfacing check valve failure probability.

DPCONT-NPHIVEXPCE-- In-Vessel Steam Explosion C.6 N Conditional probability of In-vessel steam explosion as the core slumps to the bottom of the vessel. No fire impacts associated with the phenomenon.

DPCONT-NPHMSSL-CE-- Missiles Generated Pierce C.6 N Conditional probability of structural Drywell debris generated during energetic events occurring in the RPV or containment. No fire impacts associated with the phenomenon.

DPCONT-NPHO2ADDCE-- Operation Deinerted Or 02 C.6 N Conditional probability that Introduced containment is deinerted. Fire does not impact this condition.

DPCONTNNOPCOINJ-HE-- Operator Restores Coolant C.6 N Operator action modeled in the 1-3 Injection After Control Rods are hour time frame. Instrument cues Melted have been verified available. Based on NUREG/CR-6850 and draft NUREG-1 921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

Page 19 of 31 Rev A. Page 19 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONT-NPHN2PIPCE-- Pre-Existing Failure due to C.6 N Conditional probability of failure of Liquid N2 Contact stainless steel piping in direct contact with liquid nitrogen. No fire impacts associated with the

_phenomenon.

DPCONT-NPHRINGCE-- Ring Header Failure C.6 N Conditional probability of ring header failure. No fire impacts associated with the phenomenon.

DTORVB-NVB240--FC-- Two Vacuum Breakers Fail to C.6 N Vacuum breakers are passive Close During Accident components without cables that do not postulate fire impacts per NUREG/CR-6850.

DTORVB-NVB24U--RO-- Two Vacuum Breakers Fail to C.6 N Vacuum breakers are passive Remain Closed components without cables that do not postulate fire impacts per NUREG/CR-6850.

DTORVB-NPHPOOL-CE-- Vapor Suppression Fails Due to C.6 N Conditional probability of vapor Loss of Pool suppression failure due to high pool level. Based on the discussion in the Level 2 notebook this event represents a pre initiator. Fire impacts from a MSO directing CST inventory to the pool could increase level, but would not increase level out of technical specification limits.

DPCONT-NPHDWHDFCE-- Drywell Head Closure Fails Due C.7 N Conditional probability that the to Overpressure drywell head fails to overpressure.

No fire impacts associated with the phenomenon.

DPCONT-NSSBARRI-FH-- DW Barriers Unable to Prevent C.7 N Event has a 1.0 failure probability in Debris Contact w/Shell the PRA.

Rev A. Page 20 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DFPROTDN---INJECTF-- FIRE PROTECTION C.7 N Fire protection alignment is not INJECTION ALIGNMENT NOT credited in the PRA.

CREDITED IN L2 PRA DPCONT-NSSOVERF-FH-- Melt Overflows Sump and C.7 N Event has a 1.0 failure probability in Attacks DW Shell the PRA.

DRHR--CNOPSPRYSIHE-- OP Fails to Initiate DW Sprays C.7 N Event has a 1.0 failure probability in for Debris Cooling (SI Node) the PRA.

DSYSTM-NOPALTINXHE-- Operator Fails to Align Alternate C.7 N Operator action modeled in the 1-3 Injection Sources in Level 2 hour time frame. Instrument cues have been verified available. Based on NUREG/CR-6850 and draft NUREG-1 921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DSYSTM-NOPTERMINHE-- Operator Intervenes and C.7 N Operator action modeled in the 1-3 Terminates Injection hour time frame. Instrument cues have been verified available. Based on NUREG/CR-6850 and draft NUREG-1921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DSWYRDENOPRX###CE-- Recovery of AC Power C.7 Y Recovery of fire induced failure of offsite power is not credited in the FPRA given potential cable damage impacts.

DPCONT-NSSSUMPR-FH-- Shell Failure Below Sump C.7 N Probability of shell failure below the Results in Release to Reactor pedestal. No fire impacts associated Building with the phenomenon.

Page 21 of 31 Rev A. Page 21 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONT-NSSWATER-FH-- Water Present but Debris not C.7 N Probability that debris is not coolable Coolable on DW Floor despite presence of water. No fire impacts associated with the phenomenon.

DWELLWDN---INJECTF-- WELL WATER UNAVAILABLE C.7 N Well Water cross tie is not credited FOR INJECTION TO REACTOR in the PRA to prevent core melt progression.

DPCONT-NPH-DP-CE-- HIGH DIFF PRESS PREVENTS C.9 N Conditional probability that high VALVE OPENING differential pressure prevents vent valves from opening. No fire impacts associated with the phenomenon.

DPCONT-NPHLGDWCE-- Large Drywell Failure C.9 N Conditional probability that a large drywell failure precludes flooding to TAF. No fire impacts associated with the phenomenon.

DPCONT-NOPHUCLVNHE-- Operator Fails to Close Wetwell C.9 N Operator action modeled in the 5 Vent hour time frame. Instrument cues have been verified available. Based on NUREG/CR-6850 and draft NUREG-1921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONTNNOPHUSUSFHE- OPERATOR SUSPENDS C.9 N Operator action modeled in the 5 FLOODING BASED ON hour time frame. Instrument cues ERRONOUS INDICATION have been verified available. Based on NUREG/CR-6850 and draft NUREG-1 921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DSYSTM-NOP-PCFLDHE-- Operators Fail to Implement C.9 N Operator action modeled in the 5 Primary Containment Flooding hour time frame. Instrument cues have been verified available. Based on NUREG/CR-6850 and draft NUREG-1 921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DPCONT-NPHSMCMCE-- Small Containment Failure Fails C.9 N Conditional probability that a small Injection or Precludes Flood to containment failure precludes TAF flooding to TAF. No fire impacts associated with the phenomenon.

DRHR---NPHENVIRCE-- Adverse RX Bldg Environment C.12 N Conditional probability that RHR fails Causes RHR Failure due to adverse environmental conditions. No fire impacts associated with the phenomenon.

DRHR---NPHCONTMCE-- Contingency Methods C.12 N Event has a 1.0 failure probability in Inadequate the PRA.

DCNDSRCNOPMSVOPNHE Op Fails to Open an MSIV C.12 N Event has a 1.0 failure probability in

-- _and/or Bypass Valve the PRA.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DRHR--CRMPEBRSOOCE-- PUMP BEARINGS CLOGGED C.12 N Conditional probability that suction BY MATERIAL RELEASED TO strainers are clogged with insulation POOL debris. No fire impacts associated with the phenomenon.

DRHR--CNPHRADLKCE-- Rad. Release Through RHR HX C.12 N Conditional probability that RHR is Causes Operator to Shutdown shutdown due to rad release. Fire RHR impacts would not likely alter TSCs decision from that during any other accident.

DRHR---NPHRHRRCCE-- RHR Not Recovered C.12 Y Recovery of RHR is assigned a 0.9 probability. However, when considering fire damage to cables, system recovery may not be likely.

For fire, the recovery is set to 1.0.

DRHR--CRMPVTATOOCE-- RHR PUMPS CAVITATE AT C.12 N Conditional probability that pumps SATURATED POOL cavitate at saturated pool conditions.

CONDITIONS No fire impacts associated with the phenomenon.

DSBGT--NPHNDINSCE-- Control Room Instrumentation C.13 N Conditional probability that the Fail due to Accident Conditions reactor building environment impacts indication. Fire impacts on indication are assessed on the operator action.

DSBGT--NPHWATERCE-- High Containment Water Level C.13 N Conditional probability that high Causes Vent Termination water level causes the operators to terminate venting. Fire impacts on indication are assessed on the operator action.

DSBGT--NPHHIDP-CE-- High Differential Pressure C.13 N Conditional probability that high Prevents Valve from Opening differential pressure prevents valves from opening. No fire impacts associated with the phenomenon.

Rev A. Page 24 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DSBGT--NOPL2VENTHE-- OP FAILS TO VENT C.13 N Operator action modeled in the 6 CONTAINMENT (EOP-2 Step hour time frame. Instrument cues PC/P-10) POST CORE have been verified available. Based DAMAGE on NUREG/CR-6850 and draft NUREG-1921 guidance, after one hour the fire is assumed out and not contributing to late scenario complications.

DSBGT--NPHRXENVCE-- Reactor Building Environment C.13 N Conditional probability RHR fails due Causes Failure to reactor building environmental conditions. No fire impacts associated with the phenomenon.

DPPCONTNN--DEBRITF-- DEBRIS INDUCED FAILURE C.14 N Probability that debris attack fails the OF RING HEADER torus. No fire impacts associated with the phenomenon.

DPCONTCNPPDCVENTTF-- DOWNCOMER VENT PIPES C.14 N Probability that high dynamic loads FAIL fail the torus. No fire impacts associated with the phenomenon.

DPCONTCNPPDCVEN2TF-- DOWNCOMER VENT PIPES C.14 N Probability that loss through open FAIL (CLASS II IV) wetwell vents results in inadequate pool inventory to cover the downcomers. No additional fire failure modes associated with the event.

DPCONTNN--POOL-2TF-- DYNAMIC LOADS ON THE C.14 N Probability that high dynamic loads POOL ARE EXCESSIVE fail the torus. No fire impacts associated with the phenomenon.

DPCONTNN--EXVNT1TF-- DYNAMIC LOADS ON THE C.14 N Probability that high dynamic loads POOL ARE EXCESSIVE fail the torus. No fire impacts (CLASS II IV) associated with the phenomenon.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONTNN--POOL12TF-- DYNAMIC LOADS ON THE C.14 N Probability that high dynamic loads POOL ARE EXCESSIVE fail the torus. No fire impacts (CLASS II IV) associated with the phenomenon.

DPCONTNN--EXVNT-TF-- EXCESS STEAM VENTING C.14 N Probability that high dynamic loads CAUSES POOL LEVEL fail the torus. No fire impacts DECREASE associated with the phenomenon.

DTORVBNFVBNGL--ORC-- ONE VACUUM BREAKER HAS C.14 N Probability that a vacuum breaker PRE-EXISTING FTRC FAILURE may have failed prior to the accident.

No fire impacts associated with the preexisting failure mode.

DTORVBNFVBSNGLE-FC-- ONE WW-DW VB FAILS FTC C.14 N Vacuum breakers are passive DURING ACCIDENT components without cables that do not postulate fire impacts per NUREG/CR-6850.

DTORVBNFVBSNGLE1FC-- ONE WW-DW VB FTC DURING C.14 N Vacuum breakers are passive ACCIDENT (CLS II IV) components without cables that do not postulate fire impacts per NUREG/CR-6850.

DADS--CFRVSRWB-FC-- SRV Discharge Vacuum C.14 N Vacuum breakers are passive Breakers Fail Open Following components without cables that do Blowdown not postulate fire impacts per NUREG/CR-6850.

DADS--CFRVSRWB1 FC-- SRV Discharge Vacuum C.14 N Vacuum breakers are passive Breakers Fail Open Following components without cables that do Blowdown (Class II, IV) not postulate fire impacts per NUREG/CR-6850.

DTORVB-NPHSPF--CE-- Suppression Pool Bypassed C.14 N Event has a 1.0 failure probability in

_ the PRA.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DTORVBNFVBEALSO0TF-- TEMPERATURE INDUCED C.14 N Probability that high temperatures FAILURE OF ALL VACUUM result in excessive leakage pass the BREAKER SEALS vacuum breaker seals bypass the torus. No fire impacts associated with the phenomenon.

DPCONT-NPHNCF--CE-- Large Containment Failure C.15 N Event has a 1.0 failure probability in the PRA.

DPCONT-NPHNCSISCE-- Large Containment Failure (CET C.15 N Split fraction for containment failure 1, RX=F, SI=S) size. No fire impacts associated with containment failure size failures.

DPCONT-NPHNCRXSCE-- Large Containment Failure (CET C.15 N Split fraction for containment failure 1, RX=S) size. No fire impacts associated with containment failure size failures.

DPCONT-NPHNC2--CE-- Large Containment Failure C.15 N Split fraction for containment failure (Class II) size. No fire impacts associated with containment failure size failures.

DSYSTM-NPHMUCVSCE-- FAILURE OF COOLANT C.16 N Conditional probability that INVENTORY MAKEUP (CV=S) containment venting induces injection failure. No fire impacts associated with the phenomenon.

DSYSTM-NPHMUNCFCE-- FAILURE OF COOLANT C.16 N Conditional probability that INVENTORY MAKEUP (NC=F) containment venting induces injection failure. No fire impacts associated with the phenomenon.

DSYSTM-NPHMUNCSCE-- FAILURE OF COOLANT C.16 N Conditional probability that INVENTORY MAKEUP (NC=S) containment venting induces injection failure. No fire impacts associated with the phenomenon.

Page 27 of 31 Rev A. Page 27 of 31

RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONT-NPHDIRXSCE-- Large Drywell Failure (Class II, C.17 N Conditional probability based on HIID, RX=S) containment failure size and location. No fire impacts associated with the phenomenon.

DPCONT-NPHCLSIVCE-- Large Drywell Failure (Class IV) C.17 N Conditional probability based on containment failure size and location. No fire impacts associated with the phenomenon.

DPCONT-NPHDISISCE-- Large Drywell Failure (RX=F, C.17 N Conditional probability based on SI=S) containment failure size and location. No fire impacts associated with the phenomenon.

DI-IIV see gate SBGT-DWVENT C.17 N Conditional probability based on containment failure size and location. Fire impacts associated with vent valves are included in the Level 1 fault tree logic transferred to Level 2.

DPCONT-NPHWWRXSCE-- Large Wetwell Failure (Class II, C.18 N Conditional probability based on IIID, RX=S) containment failure size and location. No fire impacts associated with the phenomenon.

DPCONT-NPHWWIIVCE-- Large Wetwell Failure (Class C.18 N Conditional probability based on IIV) containment failure size and location. No fire impacts associated with the phenomenon.

DPCONT-NPHWW-IVCE-- Large Wetwell Failure (Class IV) C.18 N Conditional probability based on containment failure size and location. No fire impacts associated with the phenomenon.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts DPCONT-NPHWWSISCE-- Large Wetwell Failure (RX=F, C.18 N Conditional probability based on SI=S) containment failure size and location. No fire impacts associated with the phenomenon.

PSCONT-NPHRB5--CE-- RX BLDG INEFFECTIVE C.19 N Conditional probability of reactor (FD=S) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

PSCONT-NPHRB1--CE-- RX BLDG INEFFECTIVE C.19 N Conditional probability of reactor (MU=F, DI=F, TD=F) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

PSCONT-NPHRB3--CE-- RX BLDG INEFFECTIVE C.19 N Conditional probability of reactor (NC=F, DI=F, WW=S) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts PSCONT-NPHRB2--CE-- RX BLDG INEFFECTIVE C.19 N Conditional probability of reactor (NC=S) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

PSCONT-NPHRB7--CE-- RX BLDG INEFFECTIVE (SI=F, C.19 N Conditional probability of reactor TD=S) (IAIE, IBL, IC, Ill) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

PSCONT-NPHRB6--CE-- RX BLDG INEFFECTIVE (SI=F, C.19 N Conditional probability of reactor TD=S) (IBE, ID) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

PSCONT-NPHRB8--CE-- RX BLDG INEFFECTIVE (SI=F, C.19 N Conditional probability of reactor TD=S) (11,IV) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

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RAI - PRA 57 Level 2 PRA Basic Event Basic Event Description Appendix C Fire Comments Section Impacts PSCONT-NPHRB4--CE-- RX BLDG INEFFECTIVE C.19 N Conditional probability of reactor (WW=F) building effectiveness based on containment location. No fire impacts associated with the phenomenon.

The assumptions include potential for penetration failures regardless of fire.

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RAI - PRA 58 DAEC RAI Category # PRA RAI 58 Describe the administrativecontrols and processes for maintainingthe FPRA quality for the NFPA 805 application, including updates based on changes to the Internal Events PRA. [RAI 58-1]

Describe your schedule with respect to NFPA 805 transition and post-transitionto align and to continue to update the alignment of the Internal Events PRA with the FPRA used to support the LAR. [RAI 58-2)

RESPONSE

RAI 58-1:

Duane Arnold procedure PSAG-2 Revision 4, PRA Model Maintenance and Update, meets all the MU supporting requirements in ASME/ANS RA-Sa-2009, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, and applies to both the Internal Events PRA and the Fire PRA.

PSAG-2 addresses the following program elements:

1. Monitoring of changes in design, operation, and maintenance that could affect the PRA. These changes include operating procedures, design configuration, initiating event frequencies, unavailabilities, and component failure rate data.

2. Monitoring of changes in PRA technology and industry experience that could change the results of the PRA.

3. Assessing changes in PRA inputs or other applicable new information and incorporating this information as appropriate in PRA maintenance activities (i.e.,

PRA update) or a PRA Upgrade.

4. Prioritizing changes that would impact risk-informed decisions with the objective to ensure that the most significant changes are incorporated as soon as practical.

5. Performing PRA changes consistent with the ASME/ANS RA-Sa-2009 supporting requirements.

6. Conducting peer reviews of PRA Upgrades (in accordance with Section 6 of the ASME PRA Standard) for those aspects of the PRA that have been upgraded.

7. Evaluating the cumulative impact of pending changes on risk applications.

8. Maintaining control of computer codes used to support PRA quantification.

Page 1 of 2 Rev A. Page 1 of 2

RAI - PRA 58 RAI 58-2:

The current requirement for updates of Internal Events PRA also requires updates of the Fire PRA. PSAG-2 applies to maintenance and update of the base Full Power Internal Events PRA model and the base Fire PRA model; it applies to the NFPA 805 transition as well as to the post-transition. The schedule of the updates to the Internal Events PRA and Fire PRA is a function of priority as summarized in item #4 above.

Procedure PSAG-2 does not cover the application of the Fire PRA model in the context of NFPA-805 requirements beyond those necessary to maintain and update the model.

Application of the Fire PRA model in support of NFPA-805 is a separate NextEra Energy fleet procedure that will be in place for transition.

Page 2 of 2 Rev A. Page 2 of 2

RAI - PRA 59 DAEC RAI PRA 59 The licensee procedure for Fire Risk Evaluations provides guidance for transition to NFPA 805 for evaluating variations from deterministic requirements. Provide a discussion of your procedure(s)/process(es) for plant change evaluations post-transition.

Include a discussion on how post-transition guidance for plant change evaluations addresses key uncertainties, assumptions, sensitivity analyses, and peer review F&Os (e.g., unaddressed F&Os).

RESPONSE

The RI-PB post transition change process, to be developed during the transition implementation period, is based upon the requirements of NFPA 805, and industry guidance in NEI 04-02, and RG 1.205. These requirements are summarized in Table 4-2, Change Evaluation Guidance Summary Table, of the enclosure to the License Amendment Request (ML11221A280) (LAR). The post-transition Plant Change Evaluation Process consists of the following 4 steps described in LAR Figure 4-9:

" Defining the Change I

" Performing the Preliminary Risk Screening.

Performing the Risk Evaluation

Evaluating the Acceptance Criteria Change Definition The Change Evaluation process begins by defining the change or altered condition to be examined and the baseline configuration as defined by the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).

The baseline is defined as that plant condition or configuration that is consistent with the Design Basis and Licensing Basis (NFPA 805 Licensing Basis post-transition).

" The changed or altered condition or configuration that is not consistent with the Licensing Basis is defined as the proposed alternative.

Preliminary Risk Review Once the definition of the change is established, a screening is then performed to identify and resolve minor changes to the fire protection program. This screening is consistent with fire protection regulatory review processes in place at nuclear plants under traditional licensing bases. This screening process is modeled after the NEI 02-03 process. This process will address most administrative changes (e.g., changes to the combustible control program, organizational changes, etc.).

Rev A. Page 1 of 3

RAI - PRA 59 The characteristics of an acceptable screening process that meets the 'assessment of the acceptability of risk' requirement of Section 2.4.4 of NFPA 805 are:

" The quality of the screen is sufficient to ensure that potentially greater than minimal risk increases receive detailed risk assessments appropriate to the level of risk.

The screening process must be documented and be available for inspection by the NRC.

The screening process does not pose undue evaluation or maintenance burden. If any of the above is not met, proceed to the Risk Evaluation step.

Risk Evaluation The screening is followed by engineering evaluations that may include fire modeling and risk assessment techniques. The results of these evaluations are then compared to the acceptance criteria. Changes that satisfy the acceptance criteria of NFPA 805 Section 2.4.4 and the license condition can be implemented within the framework provided by NFPA 805. Changes that do not satisfy the acceptance criteria cannot be implemented within this framework. The acceptance criteria require that the resultant change in CDF and LERF be consistent with the license condition. The acceptance criteria also include consideration of defense-in-depth and safety margin, which would typically be qualitative in nature. The risk evaluation involves the application of fire modeling analyses and risk assessment techniques to obtain a measure of the changes in risk associated with the proposed change. In certain circumstances, an initial evaluation in the development of the risk assessment could be a simplified analysis using bounding assumptions provided the use of such assumptions does not unnecessarily challenge the acceptance criteria discussed below. Implicit in this process is alignment with NEI 04-02, Section 5.3. In addition, FAQ 12-0061 is under development by the industry to incorporate lessons learned from the pilot post-transition change process. The following sections are applicable to this question:

5.3.3 Preliminary Risk Screening

5.3.4 Risk Evaluations - discusses assumption, procedures and acceptance criteria

5.3.5.4 Uncertainty considerations The fire protection change evaluation process, as with all risk-informed applications, will ensure that the Fire PRA used to perform the change in risk calculations, is adequate and appropriate to support the application.

Acceptability Determination The Change Evaluations are assessed for acceptability using the delta CDF (change in core damage frequency) and delta LERF (change in large early release frequency) criteria from the license condition. The proposed changes are also assessed to ensure Rev A. Page 2 of 3

RAI - PRA 59 they are consistent with the defense-in-depth philosophy and that sufficient safety margins were maintained.

References:

" The post-transition change evaluation is described in LAR section 4.7.2.

" Implementation item 15 in LAR Attachment S Page 3 of 3 Rev A. Page 3 of 3

RAI - PRA 60 DAEC RAI PRA 60 Summarize how a Fire Risk Evaluation PRA analysis is preformed for a VFDR, including how operator actions are, or are not, included. Are all retained VFDR's modeled in the PRA? If not, explain the guidelines used to decide not to model individual VFDRs in the PRA. Ifan individual VFDR does not receive a delta-risk quantification per your fire risk evaluation procedure, explain how is it treated in the Fire Risk Evaluation of the fire area.

RESPONSE

Section 5 of Report Number 0027-0042-000-004 Duane Arnold Energy Center Fire Risk Evaluations details the methodology for performing fire risk evaluations. Section 5.2.6 of the report describes the necessary preparatory evaluation. Each VFDR was reviewed to ensure that it was adequately reflected in the PRA model and to identify the PRA fault tree model basic events associated with the VFDR. This process was used to determine how to manipulate the PRA model to compare a variant case (the Fire PRA with the VFDR present) against a deterministically compliant case (typically the Fire PRA with the VFDR removed from the model to represent a deterministically compliant condition)

Section 5.3 of the report describes the PRA analysis. For the PRA analysis, the targets associated with a VFDR in the fire area were compared to the identified targets for the postulated fire scenarios in the fire area. If a fire scenario included a target associated with a VFDR then the fire scenario was identified as containing the variant condition.

For these fire scenarios, the variant condition was modified in the PRA to reflect a deterministically compliant case. This was accomplished by not failing the basic events from fire in the fire scenarios associated with the VFDR. A delta risk CDF and LERF calculation was performed between the fire scenarios with the variant case and the modified fire scenarios with the compliant case.

For some VFDRs, the targets for the VFDR were not included in the postulated fire scenarios in the applicable fire area. In these cases, the VFDR delta risk is zero and a formal delta risk calculation was not required. The acceptability of the results were documented in the applicable fire area fire risk evaluation for the VFDR. Per the fire risk evaluation process, maintenance of defense-in-depth and safety margin was reviewed for each fire area utilizing the fire risk evaluation process.

The only fire area at DAEC that is relying on post-transition recovery actions is Fire Area CBI. For the other fire areas, fire risk evaluations determined that recovery actions were not necessary. Recovery actions required to establish alternate shutdown capability were included in the PRA for fire area CB1, which includes the Main Control Room.

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v t e Letter Sequence Response to RAI
TAC:ME6818, (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance Administration Meeting Questions 3 October 2011 31 January 2012 24 February 2012 19 March 2012 26 October 2012 8 November 2012 5 December 2012 19 December 2012 23 December 2012 8 March 2013 22 March 2013 7 April 2013 26 April 2013 8 May 2013 27 June 2013 Answers 23 April 2012 23 May 2012 11 January 2013 12 February 2013 6 March 2013 1 May 2013 29 May 2013 2 July 2013 2 July 2013 Results Approval Other: ML11304A148, ML11334A150, ML12058A572, ML12284A021, ML13210A458
MONTHYEAR2012-02-27 [Table View]

NG-12-0177, Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generational Plants

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