View source for NG-12-0225, 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

{{Adams
| number = ML12146A094
| issue date = 05/23/2012
| title = 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
| author name = Wells P
| author affiliation = NextEra Energy Duane Arnold, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000331
| license number = DPR-049
| contact person = 
| case reference number = NG-12-0225
| document type = Letter type:NG
| page count = 115
| project = 
| stage = Response to RAI
}}

=Text=
{{#Wiki_filter:NEXTera ENER GY May 23, 2012 NG-12-0225 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 Liqht Water Reactor Generating Plants 

==References:==
: 1) License Amendment Request (TSCR-1 28): Transition to 10 CFR 50.48(c) -NFPA 805, Performance-Based Standard For Fire Protection For Liqht Water Reactor Generatinq Plants (2001 Edition), NG-1 1-0267, dated Auqust 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 Liqht Water Reactor Generatinq Plants (2001 Edition), NG-1 1-0384, dated October 14, 2011 3) 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 Generatinq Plants, NG-12-0177, dated April 23, 2012 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 provided responses to a portion of the requested information via Reference 3.Aoo06 KazL NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324 Document Control Desk NG-12-0225 Page 2 of 3 Per those same Staff discussions, NextEra Energy Duane Arnold committed to providing responses to the remaining requested information by May 23, 2012.Attachment 1 to this letter contains the requested information.
NextEra Energy Duane Arnold identified several mathematical symbols and equations in the Reference 3 submittal that appeared difficult to read. These symbols and equations appeared in responses to Fire Modeling (FM) RAIs 3 and 4. Attachment 2 of this letter contains revised pages 3 through 5 of RAI FM 3 and revised pages 3 through 4 of RAI FM 4. The symbols and equations on the revised pages are clearly legible. The revise pages supersede the corresponding pages submitted with Reference 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 following new commitments.
existing commitments and makes the RAI Response Number Description Fire Modeling 2 Complete cable identification, evaluation and actions in accordance with RAI FM-2.Safe Shutdown Analysis 1 Design and install the incipient detection modification (committed to in Table S-1 of the enclosure to the License Amendment Request (ML1 1221A280) (LAR)) as described in RAI SSA-1.Safe Shutdown Analysis 6 and 7 Implement a shutdown risk management process as described in RAIs SSA-6 and SSA-7.Radioactive Release 2 Enhance the pre-fire plans (committed to in Table S-2 of the enclosure to the License Amendment Request (ML1 1221A280) (LAR))as described in RAI RR-1.Probabilistic Risk Assessment 5 Revise Appendix C of the Fire Scenario Report to include barrier elements as discussed in RAI PRA-5.Probabilistic Risk Assessment 7 Update Appendix C of the Fire Scenario Report to reflect changes in MCA scenarios as discussed in RAI PRA-7.Probabilistic Risk Assessment 8 Revise Appendix C of the Fire Scenario Report to reflect changes in the MCA as discussed in RAI PRA-8.
Document Control Desk NG-12-0225 Page 3 of 3 If you have any questions or require additional information, please contact Steve Catron at 319-851-7234.
I declare under penalty of perjury that the foregoing is true and correct.Ex ed onMay 23, 2012 Peter ells Vice President, Duane Arnold Energy Center NextEra Energy Duane Arnold, LLC Attachments:
: 1) Response to Request for Additional Information, License Amendment Request to Adopt National Fire Protection Association Standard 805, Performance-Based Standard For 2) Fire Protection For Light Water Reactor Generating Plants Revised Pages to Response For Additional Information Fire Modeling RAI FM 3 and RAI FM 4 cc: NRC Regional Administrator NRC Resident Inspector NRC Project Manager M. Rasmusson (State of Iowa)
Attachment 1 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 Light Water Reactor Generating Plants 105 pages follow RAI -Monitoring 1 DAEC RAI Monitoring 1 NFPA 805, Section 2.6, "Monitoring," states that: "A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria." It also states that "Monitoring shall ensure that the assumptions in the engineering analysis remain valid." Specifically, NFPA 805, Section 2.6, states that: (2.6.1) "Acceptable levels of availability, reliability, and performance shall be established." (2.6.2) "Methods to monitor availability, reliability, and performance shall be established.
The methods shall consider the plant operating experience and industry operating experience." (2.6.3) "If the established levels of availability, reliability, or performance are not met, appropriate corrective actions to return to the established levels shall be implemented.
Monitoring shall be continued to ensure that the corrective actions are effective." Section 4.6 of the DAEC NFPA 805 Transition Report states that the DAEC NFPA 805 monitoring program will be implemented as part of the fire program transition to NFPA 805 (Attachment S, Table S-2, Implementation Items, Item 2 of the DAEC NFPA 805 Transition Report) after the safety evaluation is issued. Furthermore, the licensee has indicated that the monitoring program will be developed in accordance with, Frequently Asked Question (FAQ) 10-0059. The staff noted that the information provided in Section 4.6, "Monitoring Program," of the DAEP NFPA 805 Transition Report, is insufficient for the staff to complete its review of the monitoring program and as such is requesting that the following additional information be provided: a. A description of the process by which systems, structures, and components (SSCs) will be identified for inclusion in the NFPA 805 monitoring program including an explanation of how SSCs that are already included within the scope of the DAEC Maintenance Rule program will be addressed with respect to the NFPA 805 monitoring program.b. A description of the process that will be used to assign availability, reliability, and performance goals to SSCs within the scope of the DAEC NFPA 805 monitoring program including the approach to be applied to any SSCs for which availability, reliability, and performance goals are not readily quantified.
: c. A description of the procedures that will be employed to address SSCs that fail to meet assigned availability, reliability, or performance goals.d. A description of how the DAEC NFPA 805 monitoring program will address response to programmatic or training elements that fail to meet performance goals (examples include fire brigade response or performance standards and discrepancies in programmatic areas such as combustible control programs).
: e. A description of how the DAEC NFPA 805 monitoring program will address fundamental fire protection program elements.f. A description of how the guidance in EPRI Technical Report 1006756 will be integrated into the DAEC NFPA 805 monitoring program.g. A description of how periodic assessments of the monitoring program will be performed taking into account, where practical, industry wide operating Rev B Page 1 of 2 RAI -Monitoring 1 experience including whether this process will include both internal and external assessments and the frequency at which these assessments will be performed.
RESPONSE: DAEC will use the process as approved in FAQ 11-0059, revision 5. Revised LAR section 4.6.2 is attached.
Specific answers are provided below.a. The revised LAR Section LAR Section 4.6.2, Phase 2 Screening, describes the process by which systems, structures, and components (SSCs) will be identified for inclusion in the NFPA 805 monitoring program, including the approach to be applied to any fire protection SSCs that are already included within the scope of the Maintenance Rule program.b. The revised LAR Section 4.6.2, Phase 3 Risk Target Value Determination provides a description of the process that will be used to assign availability, reliability, and performance goals to HSS SSCs within the scope of the monitoring program. SSCs that do not meet the screening criteria in Phase 2 do not specifically require assignment of availability, reliability, and performance goals. Programmatic elements such as fire brigade performance, fire watches, combustible controls, etc., will be evaluated using the existing program health process.c. The revised LAR Section LAR Section 4.6.2, Phase 4 Monitoring Implementation, describes the process that will be employed to address SSCs that fail to meet the availability, reliability or performance goals.d. The revised LAR Section 4.6.2 Phase 4, Phase 4 Monitoring Implementation, provides a description of how the monitoring program will address response to programmatic elements that fail to meet performance goals. Training is implicitly included within the performance goals of programmatic elements.e. The revised LAR Section 4.6.2, Phase 1 Scoping and Phase 2 Screening, provide a description of how the monitoring program addresses fire protection systems and features and programmatic elements.f. As identified in License Amendment Request (ML11221A280) (LAR) Table B-I, Section 3.2.3(1) the frequency at which inspections, testing and maintenance of the fire protection systems and features is performed will be evaluated using the EPRI Technical Report 1006756. EPRI Technical Report 1006756 Section 11 contains the following guidance which ensures that reliability levels established are consistent with FPRA and Maintenance Rule, "In establishing reliability goals, each plant should determine if other programs, evaluations, or analyses have credited specific reliability values. For example, if the Fire PRA credits a specific level of reliability for a certain suppression system, the target reliability for surveillance optimization should not be below the credited value." g. The revised LAR Section 4.6.2 Phase 4 provides a description of how periodic assessments of the monitoring program will be performed including consideration of internal and external operating experience.
RevB Page 2 of 2 Rev B Page 2 of 2 DAEC REVISED SECTION 4.6 4.6 Monitoring Program 4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states: "A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria.Monitoring shall ensure that the assumptions in the engineering analysis remain valid." As part of the transition review, the adequacy of the inspection and testing program to address fire protection systems and equipment within plant inspection and the compensatory measures programs should be reviewed.
In addition, the adequacy of the plant corrective action program in determining the causes of equipment and programmatic failures and minimizing their recurrence should also be reviewed as part of the transition to a risk-informed, performance-based licensing basis.4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program This section describes the process that will be utilized to implement the post-transition NFPA 805 monitoring program. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805.See item for implementation in Attachment S. The monitoring process is comprised of four phases." Phase 1 -Scoping" Phase 2 -Screening Using Risk Criteria" Phase 3 -Risk Target Value Determination" Phase 4 -Monitoring Implementation Figure 4-8 provides detail on the Phase 1 and 2 processes.
The results of these phases will be documented in the DAEC NFPA 805 monitoring program evaluation developed during implementation.
Phase 1 -Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements will be reviewed during the implementation phase for inclusion in the NFPA 805 monitoring program: Structures, Systems, and Components required to comply with NFPA 805, specifically:
o Fire protection systems and features-Required by the Nuclear Safety Capability Assessment
-Modeled in the Fire PRA-Required by Chapter 3 of NFPA 805 DAEC REVISED SECTION 4.6 o Nuclear Safety Capability Assessment equipment 1-Nuclear safety capability assessment equipment-Fire PRA equipment-NPO equipment o Structures, systems and components relied upon to meet radioactive release criteria Fire Protection Programmatic Elements Phase 2 -Screening Using Risk Criteria The equipment from Phase 1 scoping will be screened to determine the appropriate level of NFPA 805 monitoring.
As a minimum, the SSCs identified in Phase 1 will be part of an inspection and test program and/or system/program health reporting.
If not in the current program, the SSCs will be added in order to assure that the criteria can be met reliably.The following screening process will be used to determine those SSCs that may require additional monitoring beyond normal inspection and test program and system/program health reporting activities and will be documented in the NFPA 805 Monitoring Program Engineering Evaluation.
: 1. Fire Protection Systems and Features Those fire protection systems and features identified in Phase 1 are candidates for additional monitoring in the NFPA 805 program commensurate with risk significance.
Risk significance is determined at the component, programmatic element, and/or functional level on an individual fire area basis. Compartments smaller than fire areas may be used provided the compartments are independent (i.e., share no fire protection SSCs). If compartments smaller than fire areas are used the basis will be documented in the NFPA 805 Monitoring Program Engineering Evaluation.
The Fire PRA is used to establish the risk significance based on the following screening criteria: Risk Achievement Worth (RAW) of the monitored parameter 
> 2.0 (AND) either Core Damage Frequency (CDF) x (RAW) > 1.OE-7 per year (OR)Large Early Release Frequency (LERF) x (RAW)> 1.OE-8 per year CDF, LERF, and RAW(monitored parameter) are calculated for each fire area. The 'monitored parameter' will be established at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration).
1 For the purposes of the NFPA 805 Monitoring, "NSCA equipment" is intended to include Nuclear Safety Capability Equipment, Fire PRA equipment, and NPO equipment.
DAEC REVISED SECTION 4.6 Fire protections systems and features that meet or exceed the criteria identified above are considered High Safety Significant (HSS) will be included in the monitoring program contained in the site Maintenance Rule Program described in NAP-415, Maintenance Rule Program Administration.
The remaining required fire protection systems and features will be monitored via the existing inspection and test program and in the existing system / program health program as described in ACP 1412.4, Impairments to Fire Protection Systems, ER-AA-201-2001 System and Program Health Reporting, and ER-AA-201-2002 System Performance Monitoring.
: 2. Nuclear Safety Capability Assessment Equipment Required NSCA equipment, except the NPO scope, identified in Phase 1 will be screened for safety significance using the Fire PRA and the Maintenance Rule guidelines differentiating HSS equipment from LSS equipment.
The screening will also ensure that the Maintenance Rule functions are consistent with the required functions of the NSCA equipment.
HSS NSCA equipment not currently monitored in Maintenance Rule will be included in the Maintenance Rule. All NSCA equipment that are not HSS are considered Low Safety Significant (LSS) and need not be included in the monitoring program (beyond normal inspection and test program and system/program health reporting activities).
For non-power operational modes, the qualitative use of fire prevention to manage fire risk during Higher Risk Evolutions does not lend itself to quantitative risk measurement.
Therefore, fire risk management effectiveness is monitored programmatically similar to combustible material controls and other fire prevention programs.
Additional monitoring beyond inspection and test programs and/or system/program health reporting is not considered necessary.
: 3. SSCs Relied upon for Radioactive Release Criteria The evaluations performed to meet the radioactive release performance criteria are qualitative in nature. The SSCs relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement.
Additionally, since 10 CFR Part 20 limits (which are lower than releases due to core damage and containment breach) for radiological effluents are not being exceeded, equipment relied upon to meet the radioactive release performance criteria is considered inherently low risk. Therefore, additional monitoring beyond inspection and test programs and system/program health reporting is not considered necessary.
: 4. Fire Protection Programmatic Elements Monitoring of programmatic elements is required in order to "assess the performance of the fire protection program in meeting the performance criteria".
These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements Programmatic aspects include: " Transient Combustible Control; Transient Exclusion Zones" Hot Work Control; Administrative Controls" Impairment and compensatory measures including program compliance" Fire Brigade Effectiveness DAEC REVISED SECTION 4.6 Monitoring of programmatic elements is more qualitative in nature since the programs do not lend themselves to the numerical methods of reliability and availability.
Therefore, monitoring is conducted using the existing system and program health programs.
Fire protection health reports, self-assessments, regulator and insurance company reports provide inputs to the monitoring program.Phase 3 -Risk Target Value Determination Phase 3 establishes the target values for reliability and availability for the fire protection systems and features that met or exceeded the screening criteria and the HSS NSCA equipment established in Phase 2.Target values for reliability and availability for the fire protection systems and features are established at the component level, program level, or functionally through the use of the pseudo system or 'performance monitoring group' concept. The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (-2-3 operating cycles). In addition, the EPRI Technical Report (TR) 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" (Reference 
: 28) will be used as input for establishing reliability targets, action levels, and monitoring frequency.
Since the HSS NSCA equipment have been identified using the Maintenance Rule guidelines, the associated equipment specific performance criteria will be established as in the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions.
When establishing the action level threshold for reliability and availability, the action level will be no lower than the fire PRA assumptions.
Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against established action levels. The Monitoring Program failure criteria and action level targets will be documented in the NFPA 805 Monitoring Program Engineering Evaluation.
Note that fire protection systems and features, NSCA equipment, SSCs required to meet the radioactive release criteria, and fire protection program elements that do not meet the screening criteria in Phase 2 will be included in the existing inspection and test programs and the system and program health programs.
Reliability and availability criteria will not be assigned.Phase 4 -Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established.
Monitoring consists of periodically gathering, trending, and evaluating information pertinent to the performance, and/or availability of the equipment and comparing the results with the established goals and performance criteria to verify that the goals are being met. Results of monitoring activities will be analyzed in timely manner to assure that appropriate action is taken. The corrective action process will be used to address performance of fire protection and nuclear safety SSCs that do not meet performance criteria.For fire protection systems and features and NSCA HSS equipment that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, corrective action in DAEC REVISED SECTION 4.6 accordance with PI-AA-205, Condition Evaluation and Corrective Action, will be initiated to identify the negative trend. A corrective action plan will then be developed to ensure the performance returns to the established level.When applicable, a sensitivity study can be performed to determine the margin below the action level that still provides acceptable fire PRA results to help prioritize corrective actions if the action level is reached.A periodic assessment will be performed (e.g., at a frequency of approximately every two to three operating cycles and be coordinated with the NRC Triennial Inspection Assessment), taking into account, where practical, industry wide operating experience.
This will be conducted as part of other established assessment activities.
Issues that will be addressed include: " Review systems with performance criteria.
Do performance criteria still effectively monitor the functions of the system? Do the criteria still monitor the effectiveness of the fire protection and NSCA systems?* Have the supporting analyses been revised such that the performance criteria are no longer applicable or new fire protection and NSCA SSCs, programmatic elements and/ or functions need to be in scope?" Based on the performance during the assessment period, are there any trends in system performance that should be addressed that are not being addressed?
DAEC REVISED SECTION 4.6 Phase 1 -Scoping Phase 2 -Screening Figure 4-8 -NFPA 805 Monitoring 
-Scoping and Screening RAI -Fire Modeling 1 DAEC RAI FM 1 NFPA 805, Section 2.4.3.3 requires that the PRA approach, methods, and data shall be acceptable to the Authority Having Jurisdiction (AHJ).Section 4.5.1.2, "Fire PRA," of the DAEC NFPA 805 Transition Report states that fire modeling was performed as part of the Fire PRA development (NFPA 805 Section 4.2.4.2).
Section 4.5.1.2, "Fire PRA," of the DAEC NFPA 805 Transition Report refers to Attachment J, "Fire Modeling V&V," for a discussion of the acceptability of the models that were used.Regarding the acceptability of the PRA approach, methods, and data: a. Of specific concern are fire location corner and wall proximity effects, which can affect entrainment and flame height, as well as Zone of Influence (ZOI) and target impacts. During the audit the staff discussed the issue of fire location and it was not clear how this concern was addressed.
The staff requests the licensee describe how the effects of fires located near corners and walls were accounted for in the fire modeling analyses; specifically for: 1. Fires that affect Main Control Room (MCR) abandonment.
: 2. Transient, small liquid fuel spill fires.3. Open electronic equipment (closed vented cabinet) fires throughout the plant.In addition, the staff requests the licensee describe the data collection method for specific ignition sources identified as being in close proximity to walls and/or corners.b. During the audit, the staff noted that fire modeling comprised the following:
-The Consolidated Fire Growth and Smoke Transport (CFAST) model was used to calculate abandonment times in the main control room (MCR).-The Generic Fire Modeling Treatments approach was used to determine the ZOI in all fire areas throughout plant.Explain if and how the modification to the critical heat flux for a target that is immersed in a thermal plume described in Section 2.4 of the Generic Fire Modeling Treatments document was used in the analyses to support the transition to NFPA 805. In addition, provide the title and describe the supplements to the Generic Fire Modeling Treatments that were developed and explain how these supplements were used in the analyses to support the transition to NFPA 805. Also provide a description of the specific CFAST input parameters and provide the CFAST input files and a summary of the results for the MCR abandonment study.Rev A. Page 1 of 14 Rev A.Page 1 of 14 RAI -Fire Modeling I c. During the audit, the staff observed that Section 8.7 of the FPRA Quantification Report, 493080001.04 (supporting documentation for the transition to NFPA 805), discusses essential switchgear room hot gas layer (HGL) refinements.
The first part of this discussion explains how the HGL tables from the Generic Fire Modeling Treatments document were applied to these specific fire areas.The second part of this discussion explains how new generic HGL results were developed in Generic Fire Modeling Treatments document, based on specific heat release rates (HRR), including fire growth.Provide clarification on the second part of this discussion and explain which documents were used for the HGL refinements and the process for applying the refined HGL tables.Provide the information that was obtained during the walk downs of the essential switchgear rooms (i.e. copies of the walk down sheets) together with any additional information needed to determine the ZOI in these rooms (e.g. geometry, type and location of the ignition sources and secondary combustibles, etc.)RESPONSE: a. The report entitled "Evaluation of Unit 1 Control Room Abandonment Times at the Duane Arnold Energy Center" documents the fire modeling performed for MCR abandonment.
The fire modeling evaluated transient and open electronic equipment (closed vented cabinet) fires located in an open configuration.
Based on a walkdown of the MCR for the fire model, no electronic equipment treated as ignition sources (closed vented cabinets) were adjacent to a wall or corner. As such, accounting for wall and corner effects was not required for the postulated open electronic cabinet fires.Appendix B of the MCR abandonment report was updated to include a sensitivity study for transients located in a wall or corner configuration.
Based on the results of the sensitivity study, a 9 8 th percentile transient fire in a wall or corner location would have small contribution to total plant risk. Transient fires along the wall could lead to abandonment given a loss of HVAC. Only Division 1 HVAC cables are potential targets for transient fires against the MCR walls. Therefore, Division 2 HVAC would be free of fire damage for postulated transients against walls and NUREG/CR-6850 abandonment conditions would not be met. Transient fires in a corner may result in abandonment with HVAC in normal or purge mode. Given a 9 8 th percentile transient fire in a corner, the CDF for an abandonment scenario is estimated less than 1 E-8/yr. Given the estimated CDF, transients against a wall or located in a corner in the MCR are negligible contributors to plant risk and would not change the conclusions in the LAR.The Generic Fire Modeling Treatments report was used to identify targets within the calculated critical separation distance for transients, small liquid fuel spill fires, and open electronic equipment (closed vented cabinets) throughout the plant. Section 3.3.7 of the report provides the following guidance for fuel packages positioned in a corner and a wall: 1. If the fuel package is within 0.6 m (2 ft) of a wall, then double the heat release rate and assume that the fire is centered at the fuel package edge adjacent to the wall.Rev A.Page 2 of 14 RAI -Fire Modeling 1 2. If the fuel package is within 0.6 m (2 ft) of a corner, then quadruple the heat release rate and assume that the fire is centered at the fuel package corner nearest the wall corner.During fire scenario walkdowns, data was collected for each fire ignition source throughout the plant using a fire scenario walkdown sheet. If a transient, small liquid fuel spill, or open electronic equipment (closed vented cabinet) ignition source was postulated adjacent to a wall or in a corner, then the location was identified as such on the walkdown sheet as a fire shaping factor (walkdown sheets included as Appendix F to the FPRA Fire Scenario Report, 0493080001.003).
: b. The modification to the critical heat flux for a target that is exposed to localized fire effects in an elevated ambient temperature environment is described in Section 2.4 of the Generic Fire Modeling Treatments report. Table 6-30 of the report provides a recommended treatment for targets in an elevated ambient temperature environment.
For room ambient temperatures up to 800C, the critical separation distances for IEEE-383 qualified cable were used. When the room ambient temperature was greater than 800C, based on Tables 6-31 through 6-39, the target critical separation distances for Class A combustible fuel packages were substituted.
When the room ambient temperature was greater than 1200C, based on Tables C2-1 through C2-9, the target critical separation distances for non IEEE-383 qualified cable were substituted.
Two supplements to the Generic Fire Modeling Treatments report were developed and are referenced in Attachment J of the enclosure to the License Amendment Request (ML1 1221A280) (LAR): 1. Supplemental Generic Fire Model Treatments:
Closed Electrical Panels (Rev. B)-The Generic Fire Modeling Treatments report provides treatment for open electronic panels. This supplement provides treatments applicable to closed electronic panels where the maximum leakage area between the panel interior and exterior is five percent.2. Supplemental Generic Fire Model Treatments:
Hot Gas Layer Tables (Rev. G) -The Generic Fire Modeling Treatments report provides hot gas layer tables for several generic heat release rates for steady state fires. This supplement provides additional hot gas layer tables for heat release rates that include fire growth and secondary combustibles.
Only the second supplement, Supplemental Generic Fire Model Treatments:
Hot Gas Layer Tables, was used in the analysis to support the transition to NFPA 805. This supplement was used only for the refinement of hot gas layer treatment in the Essential Switchgear Rooms as documented in Section 8.7 of FPRA Quantification Report, 493080001.004.
There are over 3,000 individual CFAST input files associated with the Generic Fire Modeling Treatments report and about 1,000 individual CFAST input files associated with the MCR abandonment study. Template input files are provided both in Appendix B Rev A.Page 3 of 14 RAI -Fire Modeling 1 of the Generic Fire Modeling Treatments report and in Appendix C of the MCR abandonment report. Output data are provided in Appendix A of the MCR abandonment report in the form of transient graphs of output parameters of interest against time.Output data for the CFAST simulations performed in support of the Generic Fire Modeling Treatments report are provided in tabular form (time to reach a particular temperature threshold) in Section 6.1 and in Appendix B of that report.c. For the sensitivity study, the Essential Switchgear Room hot gas layer (HGL)scenarios were refined based on the report Supplemental Generic Fire Model Treatments:
Hot Gas Layer Tables (Rev. G). Tables 9-1 through 9-3 of the report were used in the refined HGL evaluation.
The potential HGL is assessed based on the range of room volumes and ventilation configurations in the tables. Use of the supplement report for the sensitivity analysis did not require additional walkdowns.
However, additional walkdowns were performed in support of the clarification for the RAI response.
The walkdowns were used to verify location of ignition sources, ventilation configuration, and secondary combustibles, which are identified in Tables 1 and 2 of this RAI response.The calculated room volume of switchgear room 10E is 18,360 ft 3 and 1OF is 19,560 ft 3 (see Table C-1 of the Fire Scenario Report, 493080001.003).
From Table 0-2 of the report, the volumes considered in the HGL analysis for transient ignition sources were 11,016 ft 3 and 11,736 ft 3 for 10E and 1OF, respectively.
For other ignition sources, the volumes considered in the HGL analysis were 6,885 ft 3 and 7,335 ft 3 for 10E and 1OF, respectively.
Each switchgear room boundary includes fire dampers and normally closed doors.Additionally, each switchgear room has forced ventilation.
Given the variables in potential boundary openings, the potential for HGL is assessed for each opening configuration in the HGL result tables.Unlike the HGL analysis in Section 6 of the Generic Fire Modeling Treatments report, the Supplemental Generic Fire Model Treatments:
Hot Gas Layer Tables report includes fire growth and secondary combustibles.
Specifically, the report includes scenarios in which two 18-inch wide horizontal trays are located one foot above an electrical panel ignition source. However, there are instances in which the supplement tables are not applicable for an ignition source and the tables in Section 6 of the Generic Fire Modeling Treatments report were still used. These may include when an ignition source is in a wall or corner configuration and a larger heat release rate (HRR)is applied or when an ignition source includes more than two secondary combustibles.
Tables 1 and 2 of this RAI response identify the ignition source location, HRR applied, and the HGL results table used for each ignition source.Given the volume of the rooms, fire scenarios with transient ignition sources do not result in a HGL that can cause full room burnout. Tables 1 and 2 of this RAI response summarize the fixed ignition sources in the essential switchgear rooms and the key input parameters used in the HGL sensitivity study. From Tables 1 and 2, a HGL potentially becomes a concern greater than 30 minutes for some ignition sources when Rev A.Page 4 of 14 RAI -Fire Modeling 1 using bounding estimates from the HGL tables. When specific configurations were considered, the potential for HGL was considered even longer if the generic treatments were applied. HGL scenarios were quantified as part of the FPRA and resulted in less than one percent contribution to the total plant risk (see Table 5.4-1 of the FPRA Quantification Report). These scenarios were based on the result tables in Section 6 of the Generic Fire Modeling Treatments report. The sensitivity study was performed to determine the amount of conservatism in the postulated scenarios when compared to the result tables in the supplement.
Rev A. Page 5 of 14 Rev A.Page 5 of 14 RAI -Fire Modeling 1 Table 1 Summary of HGL Assessment for Fire Zone 10 E FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1XL80 69 Wall 1 tray 9-2 >60 min Table 9-2 for 237 (See note 1)kW used due to wall location 1 D25 702 Wall 1 tray GFMT 6-16 Table 6-16 for 1500 kW used for 2
* HRR due to wall location>30 min (See note 1)Conservative estimate that does not include fire growth and does not consider wall location adjustments for enclosure volume or boundary surface area (See GFMT Section 6.1.3.6).
If the volume is doubled then HGL is >60 min.1D20 2 1 1  Wall N/A- Closed non N/A ventilated panel Rev A. Page 6 of 14 Rev A.Page 6 of 14 RAI -Fire Modeling 1 Table 1 Summary of HGL Assessment for Fire Zone 10E FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1 B04 702 N/A Stack of 2 trays 9-3 >45 min Conservative estimate given the table results include a bounding tray configuration.
Consideration of tray actual (See note 1) height above panel and propagation to second tray would increase time to HGL.1B42 211 N/A N/A- Closed non N/A N ventilated panel 1 D22 702 N/A Stack of 2 trays 9-3 >45 min Conservative estimate given the table results include a bounding tray configuration.
Consideration of tray actual (See note 1) height above panel and propagation to second tray would increase time to HGL.1G61 69 N/A 1 tray 9-1 N 1 C352 702 N/A N/A -Closed non N/A N ventilated panel Rev A. Page 7 of 14 Rev A.Page 7 of 14 RAI -Fire Modeling 1 Table 1 Summary of HGL Assessment for Fire Zone 10E FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1X41 69 N/A Stack of 2 trays 9-1 >60 min (See note 1)702 1 D44 Wall Stack of 2 trays GFMT 6-16 Table 6-16 for 1500 kW used for 2
* HRR due to wall location>30 min (See note 1)Conservative estimate that does not include fire growth and does not consider wall location adjustments for enclosure volume or boundary surface area (See GFMT Section 6.1.3.6).
If the volume is doubled then HGL is >60 min.1 D40 211 Corner N/A -Closed non N/A ventilated panel 1 C422A 702 N/A N/A -Closed non N/A ventilated panel 1 D60 211 Wall N/A -Closed non N/A ventilated panel Rev A. Page 8 of 14 Rev A.Page 8 of 14 RAI -Fire Modeling I Table 1 Summary of HGL Assessment for Fire Zone 10E FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1C142 702 N/A N/A -Closed non N/A N ventilated panel 1A4 211 N/A 1 tray 9-2 >60 min (See note 1)1B15 211 N/A N/A- Closed non N/A N ventilated panel Table Notes 1. Time estimates are based on linear interpolation from the results provided in the tables.2. Table from Supplemental Generic Fire Model Treatments:
Hot Gas Layer Tables unless otherwise noted. If the ignition source is in a wall or corner configuration, then the HRR was increased and results from a different table from the supplement were used to conservatively estimate the HGL, if appropriate.
If the supplemental report did not include a results table to conservatively bound a specific configuration, then the Generic Fire Modeling Treatments (GFMT) report tables were used.Rev A. Page 9 of 14 Rev A.Page 9 of 14 RAI -Fire Modeling I Table 2 Summary of HGL Assessment for Fire Zone 1OF FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1D1O 211 Wall N/A -Closed N/A N non ventilated panel 1D12 702 Wall 1 tray GFMT 6-16 >30 min Conservative estimate that does not include fire growth and does Table 6-16 for 1500 not consider wall location kW used for 2
* adjustments for enclosure volume HRR due to wall (See note 1) or boundary surface area (See location GFMT Section 6.1.3.6).
If the volume is doubled then HGL is>60 min.1X31 69 N/A 3 trays 9-2 >60 min Table 9-2 for 237 kW used due to 3 trays (See note 1)1 B32 211 N/A N/A -Closed N/A N non ventilated panel Rev A.Page 10 of 14 RAI -Fire Modeling 1 Table 2 Summary of HGL Assessment for Fire Zone 1OF FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1D45/1 702 N/A Stack of 2 trays 9-3 >45 min Conservative estimate given the D4501 table results include a bounding tray configuration.
Consideration of tray actual height above panel (See note 1) and propagation to second tray would increase time to HGL.1Y4 69 N/A Stack of 2 trays 9-1 >60 min (JS401)(See note 1)1 C351 702 N/A N/A -Closed N/A N non ventilated panel 1D43 702 Wall 2 trays GFMT 6-16 >30 min Conservative estimate that does not include fire growth and does Table 6-16 for 1500 not consider wall location kW used for 2
* adjustments for enclosure volume HRR due to wall (See note 1) or boundary surface area (See location GFMT Section 6.1.3.6).
If the volume is doubled then HGL is>60 min.Rev A.Page 11 of 14 RAI -Fire Modeling 1 Table 2 Summary of HGL Assessment for Fire Zone 1OF FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1D120 702 Wall 1 tray GFMT 6-16 >30 min Conservative estimate that does not include fire growth and does Table 6-16 for 1500 not consider wall location kW used for 2
* adjustments for enclosure volume HRR due to wall (See note 1) or boundary surface area (See location GFMT Section 6.1.3.6).
If the volume is doubled then HGL is>60 min.1D15/1 702 N/A Stack of 2 trays 9-3 >45 min Conservative estimate given the D1501 table results include a bounding tray configuration.
Consideration of tray actual height above panel (See note 1) and propagation to second tray would increase time to HGL.1 G51 69 N/A Stack of 3 trays 9-2 >60 min Table 9-2 for 237 (See note 1)kW used due to 3 trays Rev A. Page 12 of 14 Rev A.Page 12 of 14 RAI -Fire Modeling I Table 2 Summary of HGL Assessment for Fire Zone 1OF FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern MUX2 702 Wall N/A -Closed N/A N non ventilated panel RMT2 702 Corner N/A -Closed N/A N UX non ventilated panel 1 D50 211 Wall N/A -Closed N/A N non ventilated panel 1X323 69 N/A 1 tray 9-1 >60 min 5 (See note 1)1A3 211 N/A 2 trays 9-2 >60 min (See note 1)Rev A. Page 13 of 14 Rev A.Page 13 of 14 RAI -Fire Modeling I Table 2 Summary of HGL Assessment for Fire Zone 1OF FIS HRR Location Secondary HGL Table Potential Comments Combustible HGL (kW) (Wall/Corner) (note 2) Concern 1 B03 702 N/A 3 trays GFMT 6-15 >40 min Conservative estimate that does not include fire growth.Consideration of tray actual height above panel would Table 6-15 for 1000 (See note 1) increase time to HGL.kW used due to 3trays Table Notes 1. Time estimates are based on linear interpolation from the results provided in the tables.2. Table from Supplemental Generic Fire Model Treatments:
Hot Gas Layer Tables unless otherwise noted. If the ignition source is in a wall or corner configuration, then the HRR was increased and results from a different table from the supplement were used to conservatively estimate the HGL, if appropriate.
If the supplemental report did not include a results table to conservatively bound a specific configuration, then the Generic Fire Modeling Treatments (GFMT) report tables were used.Rev A. Page 14 of 14 Rev A.Page 14 of 14 RAI -Fire Modeling 2 DAEC RAI FM 2 NFPA 805 Section 2.5 requires damage thresholds be established to support the performance-based approach.
Thermal impact(s) must be considered in determining the potential for thermal damage of structures, systems, or components.
Appropriate temperature and critical heat flux criteria must be used in the analysis.Section 4.5.1.2, "Fire PRA," of the Transition Report states that fire modeling was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2).
FPRA Fire Scenario Report Revision 2, Section 1.3 Assumptions and Limitations 
#8 states "DAEC has Institute of Electrical and Electronic Engineers (IEEE)-383 cables. Damage criteria for thermoset cables are assumed." Section 2.3 Damage Criteria states "Based on NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, April 2005, cable damage thresholds are generally assumed to be the limiting vulnerability.
Cable damage threshold limits are dependent on the type of cables at DAEC. The DAEC UFSAR (Section 8.3) states that DAEC has IEEE-383 equivalent cables. Therefore, the damage criteria associated with IEEE-383 cables was used." However, Table B-1 of the LAR Section 3.3.5.3 electrical cable construction is described not as compliant with IEEE-383, but based on an earlier Insulated Power Cable Engineers Association (IPCEA) Standard S-19-81, and previously approved by NRC Safety Evaluation Report dated June 1, 1978 (Section 4.8).Additionally, Section 5.1.4.5 Self Ignited Cable/Junction Box Fires states: "DAEC has IEEE-383 cables; therefore, self-ignited cable fires are not postulated per NUREG/CR-6850, Appendix R. Junction box fires are not considered given the lack of an ignition source." During the audit, the staff noted that fire modeling in support of the transition to NFPA 805 involved the use of the Generic Fire Modeling Treatments approach to determine the ZOI in all fire areas throughout plant. The Generic Fire Modeling Treatments approach constitutes an implicit use of fire modeling.The staff also noted that the ZOI in all fire areas with cable targets was determined on the basis of the tables in the Generic Fire Modeling treatments document for, "IEEE-383 Qualified Cable Target". The tables for "non-IEEE-383 Qualified Cable Target," were not used in the ZOI determination.
Section 2.0 of the Generic Fire Modeling Treatments document provides a discussion of damage criteria for different types of targets. Section 2.1 of the Generic Fire Modeling Treatments document states: "Damage to IEEE-383 qualified cables is quantified as either an imposed incident heat flux of 11.4 kW/m2 (1 Btu/s-ft2) or an immersion temperature of 329 0 C (625 0 F) per Nuclear Regulatory Guidance [NRC, 2005, NUREG 6850, 2005]." Section 2.2 of the Generic Fire Modeling Treatments document states: "Damage to non-IEEE-383 qualified cables is quantified as either an imposed incident heat flux of 5.7 kW/m2 (0.5 Btu/s-ft2) or an immersion temperature of 204&deg;C (400'F) per Nuclear Regulatory Guidance [NRC, 2005, NUREG 6850, 2005]." Rev A.Page 1 of 8 RAI -Fire Modeling 2 The above statements from Generic Fire Modeling Treatments document imply that in the Generic Fire Modeling Treatments document, IEEE-383 qualified cables are assumed to be equivalent in terms of damage thresholds to "thermoset" cables as defined in Table 8-2 of NUREG/CR-6850.
In addition, non-IEEE-383 qualified cables are assumed to be equivalent to "thermoplastic" cables as defined in Table 8-2 of NUREG/CR 6850. These assumptions may or may not be correct. An IEEE-383 qualified cable may or may not meet the criteria for a "thermoset cable" as defined in NUREG/CR-6850.
It is also possible that a non-lEEE-383 qualified cable actually meets the NUREG/CR-6850 criteria for a "thermoset" cable.The staff requests the licensee provide substantiation for the exclusive use of the ZOI tables for "IEEE-383 Qualified Cable Target" in the Generic Fire Modeling Treatments.
Further, the staff does not consider these two flame test standards (IPCEA S-19-81 and IEEE-383) alone, as qualifying criteria for the as installed cable critical damage temperature and self-ignition to be used in the Fire PRA. The staff requests the licensee provide the following information:
: a. Characterize the installed thermoset and thermoplastic cabling in the power block specifically with regard to the critical damage threshold temperatures and critical heat flux as described in NUREG/CR-6850.
: b. If thermoplastic cabling is present, discuss the additional targets created/identified using the lower critical temperature and/or heat flux criteria of NUREG/CR-6850.
: c. If thermoplastic cabling is present, discuss impact on ZOI size due to increased HRR and fire propagation.
: d. If thermoplastic cabling is present, discuss self-ignited cables and their impact to additional targets created.e. If more targets are identified what would the impact be to core damage frequency (CDF) and large early release frequency (LERF), as well as ACDF and ALERF for those fire areas affected.RESPONSE: a. DAEC has approximately 14,000 cables installed in the plant and has retrieved information on the cable material and fire test qualification of approximately 13,300 cables. Two (2) cables were identified that are not qualified to IEEE-383 or equivalent as identified by FAQ 06-0022. The two unqualified cables are routed entirely in conduit with no other cables present. DAEC has identified 459 cables that have a thermoplastic insulation or jacket material.
DAEC will complete data gathering and Fire PRA analysis of the remaining 700 cables as noted below.Rev A. Page 2 of 8 Rev A.Page 2 of 8 RAI -Fire Modeling 2 b. Table 1 provides the routing points of identified thermoplastic cables. Plant walkdowns were performed to identify the routing points and to determine if the presence of thermoplastic cables results in the routing point being an additional target not previously evaluated using thermoset cable damage criteria.
From Table 1, no new targets were included given the presence of thermoplastic cables.c. NUREG/CR-6850 Appendix E and Appendix R identify recommended HRRs for cables based on the cables being qualified or unqualified.
The impact of identified thermoplastic cables does not change the HRR or fire propagation.
: d. NUREG/CR-6850 Appendix R second paragraph of Section R.1 states, "Self ignited cable fires should be postulated in rooms with unqualified cables only or a mix of qualified and unqualified cables." DAEC did not identify any unqualified cables in fire zones modeled in the FPRA. The two identified cables are in fire zone 21G and 21N which were screened from the FPRA.e. Based on the identified unqualified or thermoplastic cables, no additional targets were identified.
As such, there is no impact to CDF and LERF, as well as ACDF and ALERF for those fire areas with identified thermoplastic cables.DAEC is continuing data collection on the remaining 700 cables where either the qualification to IEEE-383 or equivalent or insulation and jacket material are not yet known. DAEC will evaluate any additional thermoplastic or unqualified cables for impact on the Fire PRA and resultant CDF/LERF and ACDF/ALERF.
Based on results to date, DAEC does not expect to identify more than a small fraction of thermoplastic or unqualified cables nor any impact on the Fire PRA from cables that may be determined to be thermoplastic or unqualified.
Rev A. Page 3 of 8 Rev A.Page 3 of 8 RAI -Fire Modeling 2 Table 1 Thermoplastic Cable Target Summary Routing Point Fire Zone Additional target 1 C939 01AN No 1J1653 01AN No 2C829 01AN No 2J1650 01AN No 1C371 02A No 1 C933 02A No 1C939 02A No 1J1654 02A No 2C057 02A No 2C275 02A No 2C829 02A No 2J1655 02A No 2LOA01 02A No 2LOA02 02A No 2LOA03 02A No 2LOA04 02A No 2LOA05 02A No 2LOA06 02A No 2LOB01 02A No 2NOB02 02B No 2NOB03 02B No 2NOB04 02B No 2NOB05 02B No 2NOB06 02B No 2NOB07 02B No 2NOB08 02B No 2NOB09 02B No 2NOB10 02B No 2NOB11 02B No 1C933 03A No 1D127 03A No. Target cables included in raceway 1L9B01.1L9A01 03A No 1L9A40 03A No 1L9A41 03A No 1L9A42 03A No 1L9A43 03A No 1L9BO1 03A Raceway target included in fire scenario.1L9B02 03A Raceway target included in fire scenario.1C371 03D No. Target cables included in raceway 1L9A38.11L9A38 03D Raceway target included in fire scenario.11L9A39 03D Raceway target included in fire scenario.11L9A40 03D Raceway target included in fire scenario.Rev A.Page 4 of 8 RAI -Fire Modeling 2 Table 1 Thermoplastic Cable Target Summary Routing Point Fire Zone Additional target D1AO1 07B No D1A02 07B No D1A03 07B No D1A04 07B No J0102 07B No J0203 07B No K107 07B No K113 07B No K216 07B No K236 07B No L120 07B No L121 07B No L146 07B No L147 07B No L151 07B No L152 07B No L259 07B No L260 07B No C2A 07E No C2B 07E No K107 07E No K113 07E No K216 07E No K236 07E No L158 07E No L159 07E No L265 07E No L266 07E No L120 08B No L121 08B No Cable Spreading Room is evaluated with a bounding 2C275 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2P407 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2P617 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2P855 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2UOA09 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2UOA10 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2UOA11 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2U8X 11A treatment without consideration of cable type.Rev A.Page 5 of 8 RAI -Fire Modeling 2 Table 1 Thermoplastic Cable Target Summary Routing Point Fire Zone Additional target Cable Spreading Room is evaluated with a bounding 2VOF 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOA04 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOA05 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOA22 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOA23 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOD05 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOG04 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOG05 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOG06 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOG07 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOG08 11A treatment without consideration of cable type.Cable Spreading Room is evaluated with a bounding 2WOK01 11A treatment without consideration of cable type.1D127 12A No 1P198 12A No 1S5A18 12A No 1 $5A19 12A No 1S5A20 12A No 1U5R 12A No 1U7A 12A No 1W9A18 12A No 1W9A19 12A No 1W9A20 12A No 1W9A21 12A No 1W9A22 12A No 1W9A23 12A No 1W9A24 12A No 1W9A25 12A No 1W9A26 12A No 1W9A27 12A No 1W9A28 12A No 1W9A29 12A No 1W9A30 12A No 1W9A31 12A No 1W9A32 12A No 1W9A33 12A No Rev A.Page 6 of 8 RAI -Fire Modeling 2 Table 1 Thermoplastic Cable Target Summary Routing Point Fire Zone Additional target 1W9A34 12A No 1W9A35 12A No 1W9A36 12A No 1W9B01 12A No 1W9B02 12A No 1W9B03 12A No 1W9B04 12A No 1W9B05 12A No 1W9B06 12A No 1W9CO1 12A No 1W9C02 12A No 1W9C03 12A No 1W9C04 12A No 1W9C05 12A No 1 W9C06 12A No 1 W9C07 12A No 1W9C08 12A No 1W9D04 12A No 1W9D05 12A No 1W9D06 12A No 1W9D08 12A No 1W9G 12A No 1W9M 12A No 1W9N 12A No 2P407 12A No 2P617 12A No 1J0386 12B No 1J0543 12B Raceway target included in fire scenario.1N930 12B Raceway target included in fire scenario.1N931 12B Raceway target included in fire scenario.1P198 12B Raceway target included in fire scenario.1P398 12B No 2J0615 12B Raceway target included in fire scenario.2J0617 12B Raceway target included in fire scenario.2J0618 12B Raceway target included in fire scenario.2N802 12B No 2N886 12B Raceway target included in fire scenario.2N887 12B Raceway target included in fire scenario.2P855 12B Raceway target included in fire scenario.Fire zone is evaluated with a bounding treatment G5HO1 CT1 without consideration of cable type.Fire zone is evaluated with a bounding treatment G5H02 CT1 without consideration of cable type.G5L01 CT2 Fire zone is evaluated with a bounding treatment Rev A.Page 7 of 8 RAI -Fire Modeling 2 Table 1 Thermoplastic Cable Target Summary Routing Point Fire Zone Additional target without consideration of cable type.Fire zone is evaluated with a bounding treatment G5L02 CT2 without consideration of cable type.Fire zone is evaluated with a bounding treatment SYCH OAG without consideration of cable type.Fire zone is evaluated with a bounding treatment TW-A OAG without consideration of cable type.Fire zone is evaluated with a bounding treatment MH104 OUG without consideration of cable type.Fire zone is evaluated with a bounding treatment MH105 OUG without consideration of cable type.Fire zone is evaluated with a bounding treatment MHI16 OUG without consideration of cable type.Fire zone is evaluated with a bounding treatment MH107 OUG without consideration of cable type.Rev A. Page 8of8 Rev A.Page 8 of 8 RAI -Safe Shutdown Analysis 1 DAEC RAI SSA 1 Incipient Detection 
-In LAR Attachment S, Table S-1, an incipient detection system is identified as a committed modification to 12 Control Room Panels.a. Because of the various vendor types of incipient detection systems, provide a description of the incipient detection system being installed/considered.
If the system has not yet been designed or installed, provide the specified design features for the proposed system along with a comparison of these specified design features to their role in satisfying or supporting the risk reduction features being credited in FAQ 08-0046. Include in this description the installation testing criteria to be met prior to operation.
: b. Describe the physical separation of the cabinets in which incipient detection is being installed.
Describe the process for estimating the conditional probability of damage to a set of target items as defined in Appendix L, Main Control Board Fires of NUREG/CR-6850.
Justify any deviations from the methods described in NUREG/CR-6850.
: c. Describe how each cabinet will be addressable by the detection system.d. Provide the codes of record for the design/installation.
: e. Based on the operator recognizing the impacted cabinet(s) fire location sufficiently early, describe what operator actions are necessary to limit fire impact and allow safe shutdown of the plant from the control room? Describe how will the operator be made aware of what must be done to remain in the control room for plant shutdown.RESPONSE: a. Duane Arnold has not designed and developed the modification at this time.Incipient detection will be designed and installed in cabinets 1C03, 1C04, 1C05, 1C06, 1C08, 1C15, 1C17, 1C26, 1C31, 1C32, 1C33, and 1C44. Two diverse detection technologies; cloud chamber and laser detection are under consideration.
The design will be based on FAQ 08-0046 and will meet the FAQ guidance such as; sensitivity, equipment voltage restrictions and fast versus slow acting devices in regard to fire growth. The system will be tested in accordance with the manufacturers and code requirements including sensitivity.
A preliminary inventory of the cabinets indicates that there is no equipment that will exceed the 250 VDC and 480VAC restriction and there is a minimal amount of equipment that will be classified as fast acting. The Fire PRA design credit includes a conservative estimate which is described in RAI PRA-35 and is based on NUREG 6850 guidance.
The risk reduction credit is Rev A. Page 1 of 2 Rev A.Page I of 2 RAI -Safe Shutdown Analysis 1 described in the Updated NFPA 805 LAR Model Updated Quantification Report (049080001.004).
: b. The cabinets where incipient detection will be installed are individual cabinets each having a substantial metal outer wall. Some of the cabinets include metal inter-cabinet division panels as well. The cables for the Main Control Board Cabinets and other cabinets enter the cabinet into penetrations at the top and bottom of the cabinets (i.e., cable are not routed between cabinets but may be routed through the inter-cabinet divisions).
Each of these cabinets currently contains ionization smoke detectors.
Therefore, fire spread across cabinets was considered low risk (see NUREG/CR-6850 Section 11.5.2.8) and only fire impacts for the cabinet internals was quantified.
For these scenarios, NUREG/CR-6850 Appendix L was not used given the DAEC control room cabinet configuration (i.e., cable routing and cabinet separation).
Use of Appendix L considering DAEC configuration may further reduce the conditional probability of fire damage. However, the current treatment is conservative and provides the necessary risk insights for cabinet fires in the control room.c. The detection system configuration is under investigation and will consist of a common air piping system with a common alarm unit or an alarm unit that is individually assigned to each cabinet. If a common alarm module is selected the means to diagnose an alarm to identify the specific cabinet in question will be included in the design and performed by responders using local supplemental incipient detection equipment guided by operating procedures.
: d. The system will be designed and installed in accordance with NFPA 72 and 76.e. Alarm Response Procedures will be developed to guide the Operator response to both alert and alarm events. The procedures will provide guidance on a cabinet by cabinet basis as to what actions are recommended in regard to diagnosing the cause of an alert/alarm, providing recommended compensatory measures and identification of support resources.
The Alarm Response Procedures will be designed to work in conjunction with existing operating procedures, abnormal operating and emergency response procedures.
Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -Safe Shutdown Analysis 2 DAEC RAI SSA 2 Provide a gap analysis on the differences between the alignments using NEI 00-01, Revision 2, as the basis for transitioning, compared to NEI 00-01, Revision 1.RESPONSE: DAEC has performed a comparison between NEI 00-01, Revision 1, and NEI 00-01, Revision 2. This gap analysis has been documented as Attachment 3 in an update to FPLDAO1 3-PR-002, Table B-2 -NFPA 805 Chapter 2 Nuclear Safety Transition Methodology Review (the supporting document for Attachment B to the DAEC LAR).Based on the gap analysis, there are no significant differences between alignment with NEI 00-01, Revision 1 and NEI 00-01, Revision 2, for the DAEC.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -Safe Shutdown Analysis 6 DAEC RAI SSA 6 Provide the following pertaining to non-power operations (NPO) discussions provided in Section 4.3 and Attachment D of the LAR: a. Identify and describe the changes to outage management procedures, risk management tools, and any other document resulting from incorporation of KSF identified as part of NFPA 805 transition.
Include changes to any administrative procedures such as "Control of Combustibles".
: b. Provide a list of the additional components 
[for which cable selection was performed]
and a list of those at-power components that have a different functional requirement for NPO. Describe the difference between the at-power safe shutdown function and the NPO function.c. Provide a list of KSF pinch points by fire area that were identified in the NPO fire area reviews using FAQ 07-0040 guidance including a summary level identification of unavailable paths in each fire area. Describe how these locations will be identified to the plant staff for implementation.
: d. Provide a description of any actions, including pre-fire staging actions, being credited to minimize the impact of fire-induced spurious actuations on power operated valves (e.g., air operated valves (AOVs) and motor operated valves (MOVs)) during NPO (e.g., pre-fire rack-out, "pinning" valves, or isolation of air supply).e. Describe the types of compensatory actions that will be used during [normal outage evolutions when certain NPO credited equipment will have to be removed from service].f. Identify those recovery actions and instrumentation relied upon in NPO by physical analysis unit and describe how recovery action feasibility is evaluated.
Include in the description whether these have been or will be factored into operator procedures supporting these actions.RESPONSE: a. DAEC is planning a top-down hierarchical approach to revising NextEra Energy fleet level outage shutdown risk management procedures and the associated DAEC-specific procedures for managing shutdown risk. These documents will provide departments and organizations that plan outage related work and the DAEC Risk Assessment Team with shutdown and risk management guidance to include:* basis for NFPA 805 non power operational requirements;
* criteria for specifying HREs;* identification of KSF pinch points associated with direct fire damage or removal of credited equipment from service;* appropriate contingency measures for consideration; and* proposed options to reduce fire risk in those locations where fire can result in loss of defense in depth for one or more KSFs during HREs. These would include: o Prohibition or limitation of hot work in fire areas during periods of increased vulnerability; Rev A. Page 1 of 12 Rev A.Page 1 of 12 RAI -Safe Shutdown Analysis 6 o Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability; and o Provision of additional fire patrols at periodic intervals during increased vulnerability.
The following procedures currently implement shutdown risk and the essential work planning and implementing processes.
These and other pertinent procedures will be reviewed and modified as noted to implement these changes and requirements:
* NP-909, Shutdown Risk* OM-AA-101-1000 (DAEC), Shutdown Risk Management (DAEC Specific Information)
* MA-AA-1 00, Conduct of Maintenance
* MA-AA-203, Work Order Planning Process* MA-AA-204, Preventive Maintenance and Surveillance Process" WM-AA-200, Work Management Process Overview* ACP 1412.2, Control of Combustibles
* ACP 1412.3, Control of Ignition Sources* ACP 1412.4, Impairments to Fire Protection Systems* MA-AA-100-1008, Station Housekeeping and Material Control* WM-AA-1 000, Work Activity Risk Management
* Site Fire Plan b. The NPO Modes Review identified systems used for accomplishment of required KSFs and grouped those components making up success paths into Function Codes. Because they were not credited in the at-power analysis, cable selection was performed for the following 115 electrically-supervised components:
COMPONENT COMPONENT FUNCTION 152-101 AUXILIARY TRANSFORMER 1X2 FEEDER TO 1A1 152-103 REACTOR FEED PUMP 1P-1A 152-104 REACTOR RECIRCULATION MG SET 1G-201A 152-105 CIRC WATER PUMP 1P-4A 152-106 CONDENSATE PUMP 1P-8A 152-107 TB 480VAC LOAD CENTER 1 B1 (VIA 1X11)152-108 COOLING TOWER 480VAC LOAD CENTER 1B7(VIA 1X71)152-109 TB 480VAC LOAD CENTER 1B5 (VIA 1X51)152-110 4160/480VAC SWITCHYARD LOAD CENTER TRANSFORMER Rev A.Page 2 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION 152-201 AUXILIARY TRANSFORMER lX2 FEEDER TO 1A2 152-203 REACTOR FEED PUMP 1P-1B 152-204 REACTOR RECIRCULATION MG SET 1G-201B 152-205 CIRC WATER PUMP 1 P-4B 152-206 CONDENSATE PUMP 1P-8B 152-207 480 VAC LOAD CENTER 1B2 VIA TRANSFORMER 1X21 152-208 480 VAC LOAD CENTER 1B8 VIA TRANSFORMER 1X81 152-209 480 VAC LOAD CENTER 1B6 VIA TRANSFORMER 1X61 152-210 GENERAL SERVICE WATER PUMP 1P-89C 152-211 WELL WATER PUMP 1P-58D, POWER PANEL 1C-374 1B01 TURBINE BUILDING 480VAC LOAD CENTER 1805 TURBINE BUILDING 480VAC LOAD CENTER 1B06 TURBINE BUILDING 480 VAC NONESSENTIAL LOAD CENTER 1B13 PUMP HOUSE 480 VAC MOTOR CONTROL CENTER 1B14 RB 812' LEVEL NORTH END MOTOR CONTROL CENTER 1815 480V MCC 1B15 18333 TURBINE BUILDING 480 VAC MOTOR CONTROL CENTER 1 B35 RB 786' LEVEL 480 VAC MOTOR CONTROL CENTER 1 B43 RB 757' LEVEL 480 VAC MOTOR CONTROL CENTER 11345 480V MCC 1 B45 1B52 TURBINE BUILDING 480 VAC MOTOR CONTROL CENTER 18362 TURBINE BUILDING 480 VAC MOTOR CONTROL CENTER Rev A.Page 3 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION 1BR91 INST AIR BUILDING 480VAC MOTOR CONTROL CENTER 1BR92 INST AIR BLDG 480 VAC MOTOR CONTROL CENTER 1D45 120 VOLT UNINTERRUPTIBLE AC POWER SUPPLY 1K001 BACKUP INSTRUMENT AIR COMPRESSOR analysis includes cables for: S SV4753 -100# AIR SUPPLY TO CV-4753 (1K-1 COOLING WATER SUPPLY)1K90A INST AIR COMPRESSOR analysis includes cables for:* SV3080A -1 K-90A COOLING WATER INLET ISOLATION 1 K90B INST AIR COMPRESSOR analysis includes cables for:* SV3080B -1K-90B COOLING WATER INLET ISOLATION 1K90C INST AIR COMPRESSOR analysis includes cables for: e SV3080C -1K-90C COOLING WATER INLET ISOLATION 1P011 CONDENSATE SERVICE WATER JOCKEY PUMP 1PO12A CONDENSATE SERVICE WATER PUMP analysis includes cables for: " SV5228A -CONTROL AIR SUPPLY ISOLATION for CV-5228A (COND SERVICE WATER PUMP 1P-12A DISCH ISOLATION)" SV5228B -CONTROL AIR SUPPLY ISOLATION for CV-5228B (COND SERVICE WATER PUMP 1P-12A DISCH ISOLATION) 1P012B CONDENSATE SERVICE WATER PUMP analysis includes cables for:* SV5229A -CONTROL AIR SUPPLY ISOLATION for CV-5229A (COND SERVICE WATER PUMP 1P-12B DISCH ISOLATION)" SV5229B -CONTROL AIR SUPPLY ISOLATION for CV-5229B (COND SERVICE WATER PUMP 1P-12B DISCH ISOLATION)
Rev A. Page 4 of 12 Rev A.Page 4 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION 1P058A WELL PUMP analysis includes cables for: " SV4417A -NORMAL CONTROL PILOT SOLENOID for CV-4417 (WELL PUMP 1P-58A DISCHARGE CHECK VALVE)" SV4417B -EMERGENCY CONTROL PILOT SOLENOID for CV-4417 (WELL PUMP 1P-58A DISCHARGE CHECK VALVE)1P058B WELL PUMP analysis includes cables for: " SV4422A -NORMAL CONTROL PILOT SOLENOID for CV-4422 (WELL PUMP 1 P-58B DISCHARGE CHECK VALVE)" SV4422B -EMERGENCY CONTROL PILOT SOLENOID for CV-4422 (WELL PUMP 1P-58B DISCHARGE CHECK VALVE)1 P058C WELL PUMP analysis includes cables for:* SV4483A -NORMAL CONTROL PILOT SOLENOID for CV-4483 (WELL PUMP 1 P-58C DISCHARGE CHECK VALVE)" SV4483B -EMERGENCY CONTROL PILOT SOLENOID for CV-4483 (WELL PUMP 1P-58C DISCHARGE CHECK VALVE)1P058D WELL PUMP 1P081A RB CLOSED COOLING WATER PUMP 1P081B RB CLOSED COOLING WATER PUMP 1 P081C RB CLOSED COOLING WATER PUMP 1P089A GENERAL SERVICE WATER PUMP 1P089B GENERAL SERVICE WATER PUMP 1 P089C GENERAL SERVICE WATER PUMP 1P209A CONTROL ROD DRIVE HYDRAULIC PUMP 1P209B CONTROL ROD DRIVE HYDRAULIC PUMP 1P214A FUEL POOL COOLING PUMP 1P214B FUEL POOL COOLING PUMP Rev A. Page 5 of 12 Rev A.Page 5 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION 1 P278A INST AIR COMP COOLING WATER RECIRC PUMP analysis includes cables for: " 1VEF090A -HEAT EXCHANGER 1V-HX-90 COOLING FAN* 1VEF091A -HEAT EXCHANGER 1V-HX-91 COOLING FAN 1P278B INST AIR COMP COOLING WATER RECIRC PUMP analysis includes cables for: " 1VEF090B -HEAT EXCHANGER 1V-HX-90 COOLING FAN" 1VEF091B -HEAT EXCHANGER 1V-HX-91 COOLING FAN 1S024 GSW AUTOMATIC BACKWASH STRAINER 1T206A DEMINERALIZER,FILTER,FPCC,FUEL POOL analysis includes cables for: " SV3504A -CONTROL AIR SUPPLY ISOLATION for CV-3504A (1T-206A MAIN DRAIN TO WASTE SLUDGE TANK)* SV3510A -CONTROL AIR SUPPLY ISOLATION for CV-3510A (FUEL POOL F/D 1T-206A DISCHARGE ISOLATION)
* SV3515A -CONTROL AIR SUPPLY ISOLATION for CV-3515A (FUEL POOL F/D 1T-206A BACKWASH AND FILL VALVE)" SV3518 -CONTROL AIR SUPPLY ISOLATION for CV-3518 (FUEL POOL F/D 1T-206A INFLUENT CONTROL VALVE)* SV3526A -CONTROL AIR SUPPLY ISOLATION for CV-3526A (FUEL POOL F/D 1T-206A PRECOAT RETURN VALVE)* SV3530A -CONTROL AIR SUPPLY ISOLATION for CV-3530A (FUEL POOL F/D 1T-206A PRECOAT SUPPLY VALVE)Rev A. Page 6 of 12 Rev A.Page 6 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION 1 T206B DEMINERALIZER,FILTER,FPCC,FUEL POOL analysis includes cables for: " SV3504B -CONTROL AIR SUPPLY ISOLATION for CV-3504B (1T-206B MAIN DRAIN TO WASTE SLUDGE TANK)* SV3510B -CONTROL AIR SUPPLY ISOLATION for CV-3510B (FUEL POOL F/D 1T-206B DISCHARGE ISOLATION)" SV3515B -CONTROL AIR SUPPLY ISOLATION for CV-3515B (FUEL POOL F/D 1T-206B BACKWASH AND FILL VALVE)" SV3517 -CONTROL AIR SUPPLY ISOLATION for CV-3517 (FUEL POOL F/D 1T-206B INFLUENT CONTROL VALVE)" SV3526B -CONTROL AIR SUPPLY ISOLATION for CV-3526B (FUEL POOL F/D 1T-206B PRECOAT RETURN VALVE)" SV3530B -CONTROL AIR SUPPLY ISOLATION for CV-3530B (FUEL POOL F/D 1T-206B PRECOAT SUPPLY VALVE)1VAC013A CRD PUMP ROOM COOLING UNIT 1VAC013B CRD PUMP ROOM COOLING UNIT lX011 480VAC LOAD CENTER 11B1 SUPPLY TRANSFORMER 1X051 480VAC LOAD CENTER 1B5 SUPPLY TRANSFORMER 1X061 LOAD CENTER 1B6 FEEDER TRANSFORMER FROM 1A2 1XR004 WELL HOUSE B 4160/480 VAC TRANSFORMER 1XR9A LOAD CENTER 1BR91 SUPPLY TRANSFORMER 1XR9B LOAD CENTER 1 BR92 SUPPLY TRANSFORMER 1Y002 INSTRUMENT AC PANEL 1Y21 SUPPLY TRANSFORMER 1Y004 UNINTERRUPTIBLE AC 1Y23 REGULATING TRANSFORMER 1Y022 1Y02 TO 1Y23 AUTOMATIC TRANSFER SWITCH 1Y023 120V UNINTERRUPTIBLE AC DISTRIBUTION PANEL Rev A. Page 7 of 12 Rev A.Page 7 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION 52-101 480VAC LOAD CENTER 1B1 FEEDER (FROM 1A1 07)52-103 PUMP HOUSE 480VAC MOTOR CONTROL CENTER 1B13 52-104 RB 480VAC MOTOR CONTROL CENTER 1B14 52-105 CB 480VAC MOTOR CONTROL CENTER 1B15 52-302 CB 480VAC MOTOR CONTROL CENTER 1 B33 52-304 CB 480VAC MOTOR CONTROL CENTER 1 B35 BREAKER 52-402 RB 480VAC MOTOR CONTROL CENTER 1 B43 BREAKER 52-404 TB 480VAC MOTOR CONTROL CENTER 1B45 52-501 480VAC LOAD CENTER 1B5 FEEDER (FROM 1A1 09)52-502 TB 480VAC MOTOR CONTROL CENTER 1B52 52-601 480 VAC LOAD CENTER 1B6 FEEDER (FROM 1A209)52-602 TURBINE BUILDING 480 VAC MCC 1B62 72-406 INVERTER 1D45 DC SUPPLY BKR GB 36KV DAEC LLRPSF XFMR 1XR1 SUPPLY BREAKER BKR GC 36KV DAEC LLRPSF XFMR 1XR2 SUPPLY BREAKER BKR LQ 161KV WEST BUS -DAEC LLRPSF XFMRS BREAKER E/P1814 CRD FLOW CONTROL STATION E/P CONVERTER FIC4414A A WELL HOLD ON LOSS OF SIGNAL CONTROLLER analysis includes cables for: e M04414A -WELL PUMP 1P-58A DISCHARGE ISOLATION FIC4414B B WELL HOLD ON LOSS OF SIGNAL CONTROLLER analysis includes cables for:* M04414B -WELL PUMP 1P-58B DISCHARGE ISOLATION FIC4414C C WELL HOLD ON LOSS OF SIGNAL CONTROLLER analysis includes cables for:* M04414C -WELL PUMP 1P-58C DISCHARGE ISOLATION Rev A. Page 8 of 12 Rev A.Page 8 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION FIC4414D D WELL HOLD ON LOSS OF SIGNAL CONTROLLER M01830 CRD DRIVE WATER PRESSURE CONTROL VALVE M02039A CB CHILLER 1V-CH-1A WELL WATER SUPPLY ISOLATION M02039B CB CHILLER 1V-CH-1B WELL WATER SUPPLY ISOLATION M02039C CB CHILLERS WELL WATER BYPASS ISOLATION M02077 CHILLER 1V-CH-1A DISCH TO WELL WTR ISOLATION M02078 CHILLER 1V-CH-1 B DISCH TO WELL WTR ISOLATION M04627 RX RECIRC PUMP 1P-201A DISCHARGE ISOLATION M04628 RX RECIRC PUMP 1P-201B DISCHARGE ISOLATION M04772 TURBINE BLDG NORTH END GSW SUPPLY HDR ISOLATION M04775 TURBINE BLDG SOUTH END GSW SUPPLY HDR ISOLATION RHR LOGIC RHR LOGIC (SCHEME 1S106)lS106 RHR LOGIC RHR LOGIC (SCHEME 2S206)2S206 SV1497 CONTROL AIR SUPPLY ISOLATION for CV-1497 (CRD HYDRAULIC SYSTEM SUCTION FROM COND REJECT)SV3034 CONTROL AIR SUPPLY ISOLATION for CV-3034 (BALANCE OF PLANT INST AIR HEADER ISOLATION)
SV3035 CONTROL AIR SUPPLY ISOLATION for CV-3035 (TURBINE BLDG INSTRUMENT AIR HEADER ISOLATION)
SV3039 CONTROL AIR SUPPLY ISOLATION for CV-3039 (RX BLDG INSTRUMENT AIR SUPPLY HDR ISOLATION)
SV7103A CONTROL AIR SUPPLY ISOLATION for CV-7103A (1V-AC-13A WELL WATER RETURN ISOLATION)
SV7103B CONTROL AIR SUPPLY ISOLATION for CV-7103B (1V-AC-13B WELL WATER RETURN ISOLATION)
SV7104A CONTROL AIR SUPPLY ISOLATION for CV-7104A (1V-AC-13A WELL WATER SUPPLY ISOLATION)
SV7104B CONTROL AIR SUPPLY ISOLATION for CV-7104B (1V-AC-13B WELL WATER SUPPLY ISOLATION)
Rev A.Page 9 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION XR1 TRANSFORMER,36KV TO 4.16KV,LLRWPF XR2 TRANSFORMER,36KV TO 4.16KV,LLRWPF The majority of equipment required to maintain the NPO KSFs is the same as that required to safely shutdown the plant while at power. The following 14 electrically-supervised safe shutdown components have a different functional requirement during NPO modes; however, since all cables for these components were selected in the current SSA, no additional cable selection was required: COMPONENT COMPONENT FUNCTION NPO FUNCTION AT-POWER FUNCTION 152-102 STARTUP TRANSFORMER CLOSED AVAILABLE 1X3 FEEDER TO 1Al (power supply) (trip capability) 152-202 STARTUP TRANSFORMER CLOSED AVAILABLE 1X3 FEEDER TO 1A2 (power supply) (trip capability)
M01908 RHR SHUTDOWN COOLING OPEN CLOSED SUCTION ISOLATION (SDC mode) (RPV isolation)
M01909 RHR SHUTDOWN COOLING OPEN CLOSED OUTBOARD SUCTION (SDC mode) (RPV isolation)
ISOL M01912 RHR PP 1P-229B S/D CLNG OPEN CLOSED& FUEL POOL CLNG (SDC mode) (LPCI, SPC SUCTION modes)M01913 RHR PUMP 1P-229B TORUS CLOSED OPEN SUCTION ISOLATION (SDC mode) (LPCI, SPC modes)M01920 RHR PP 1P-229D S/D CLNG CLOSED OPEN& FUEL POOL CLNG (SDC mode) (LPCI, SPC SUCTION modes)M01921 RHR PUMP 1P-229D TORUS OPEN CLOSED SUCTION ISOLATION (SDC mode) (LPCI, SPC modes)M01935 RHR PUMPS 1P-229B/D CLOSED AVAILABLE MINIMUM FLOW BYPASS (SDC mode) (LPCI, SPC modes)M02009 RHR PUMPS 1P-229A/C CLOSED AVAILABLE MINIMUM FLOW BYPASS (SDC mode) (LPCI, SPC modes)M02011 RHR PP 1 P-229A S/D CLNG OPEN CLOSED& FUEL POOL CLNG (SDC mode) (LPCI, SPC SUCTION modes)Rev A.Page 10 of 12 RAI -Safe Shutdown Analysis 6 COMPONENT COMPONENT FUNCTION NPO AT-POWER FUNCTION FUNCTION M02012 RHR PUMP 1P-229A TORUS CLOSED OPEN SUCTION ISOLATION (SDC mode) (LPCI, SPC modes)M02015 RHR PUMP 1P-229C TORUS CLOSED OPEN SUCTION ISOLATION (SDC mode) (LPCI, SPC modes)M02016 1P-229C SHUTDOWN OPEN CLOSED COOLING & FUEL POOL (SDC mode) (LPCI, SPC COOLING SUC modes)c. NPO fire scenarios assumed room/area burnout for the same fire areas evaluated in the At-power Analysis.
Thus, entire fire areas are identified as KSF pinch points when the NPO fire area review indicates failure of all methods for achieving one or more KSFs. In the seven areas listed below, the assumed NPO fire scenario could cause a loss of all success paths for one or more KSFs: Area KSF Pinch Point / unavailable KSF Path(s)CB1 Decay Heat Removal -RHR Suction Line, A&B RHR SDC Mode, Manual-ADS/SRV, Electrical Power Availability, and Support Systems.Inventory Control -A&B Core Spray, A&B RHR LPCI Mode, CRD Hydraulics, Condensate Service, Electrical Power Availability, and Support Systems Electrical Power Availability 
-LPCI Swing Bus, 1A3&1A4 Essential AC Pwr, Offsite AC Power, 1A1&1A2 Non-Essential AC Pwr, and Uninterruptible 120 VAC Pwr.Support Systems -RHR/ESW Discharge, A&B RHRSW, A&B River Water, A&B ESW, RX bldg CCW, GSW, WW, A&B EDG Fuel Oil, A&B EDG Rm HVAC, A&B RHR/CS Rm HVAC, and CRD Pump Room Cooling.CB2 Decay Heat Removal -RHR Suction Line, Electrical Power Availability,"B" RHR SDC Mode.Electrical Power Availability -LPCI Swing Bus CB3 Decay Heat Removal -RHR Suction Line DRY Decay Heat Removal -RHR Suction Line, A&B RHR SDC Mode, and Manual -ADS/SRV Rev A. Page 11 of 12 Rev A.Page 11 of 12 RAI -Safe Shutdown Analysis 6 Area KSF Pinch Point / unavailable KSF Path(s)RB1 Decay Heat Removal -RHR Suction Line, A&B RHR SDC Mode, Electrical Power Availability, and Support Systems.Inventory Control -A&B Core Spray, A&B RHR LPCI Mode, CRD Hydraulics, Condensate Service, Electrical Power Availability, and Support Systems Electrical Power Availability 
-LPCI Swing Bus, 1A3&1A4 Essential AC Pwr, and 1AI&1A2 Non-Essential AC Pwr Support Systems -A&B RHRSW, A&B RHR.CS Rm HVAC, and CRD Pump Room Cooling RB3 Decay Heat Removal -RHR Suction Line, A&B RHR SDC Mode and Support Systems.Inventory Control -A&B Core Spray, A&B RHR LPCI Mode, Condensate Service, and Support Systems Support Systems -A&B ESW, RX bldg CCW.TB1 Decay Heat Removal -A&B RHR SDC Mode Each of these areas will be identified to the plant staff through administrative procedures governing fire protection defense in depth features, shutdown risk management, and work control.d. No particular configuration changes/equipment realignments have been specified to prevent failure of any KSF due to fire during NPO Modes.e. The additional KSF pinch points introduced by removal of credited equipment from service will be identified to the plant staff through administrative procedures governing fire protection defense in depth features, shutdown risk management, and work control. In the unlikely event that such equipment is deliberately removed from service coincident with a planned or emergent HRE, the DAEC Risk Assessment Team will consider appropriate contingency measures to reduce fire risk at the additional locations.
As with pinch points associated with direct fire damage (see sub-item a. above), proposed options to reduce fire risk will include:* Prohibition or limitation of hot work in fire areas during periods of increased vulnerability;
* Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability; and* Provision of additional fire patrols at periodic intervals during increased vulnerability.
: f. No particular operator actions have been specified for restoration of any KSF.Rev A.Page 12 of 12 RAI -Safe Shutdown Analysis 7 DAEC RAI SSA 7 Provide a description of what [the] changes [needed to implement the results of the Non-Power Operational Modes Analysis]
entail and where will they be incorporated.
RESPONSE: As described in response to sub-item a. to DAEC RAI SSA 6, DAEC is planning a top-down hierarchical approach to revising NextEra Energy fleet level outage shutdown risk management procedures and the associated DAEC-specific procedures for managing shutdown risk. These documents will provide departments and organizations that plan outage related work and the DAEC Risk Assessment Team with shutdown and risk management guidance to include:* basis for NFPA 805 non power operational requirements;
* criteria for specifying HREs;* identification of KSF pinch points associated with direct fire damage or removal of credited equipment from service;* appropriate contingency measures for consideration; and" proposed options to reduce fire risk in those locations where fire can result in loss of defense in depth for one or more KSFs during HREs. These would include: o Prohibition or limitation of hot work in fire areas during periods of increased vulnerability; o Prohibition or limitation of combustible materials in fire areas during periods of increased vulnerability; and o Provision of additional fire patrols at periodic intervals during increased vulnerability.
The following procedures currently implement shutdown risk and the essential work planning and implementing processes.
These and other pertinent procedures will be reviewed and modified as noted to implement these changes and requirements:
* NP-909, Shutdown Risk* OM-AA-101-1000 (DAEC), Shutdown Risk Management (DAEC Specific Information)" MA-AA-1 00, Conduct of Maintenance
* MA-AA-203, Work Order Planning Process* MA-AA-204, Preventive Maintenance and Surveillance Process* WM-AA-200, Work Management Process Overview* ACP 1412.2, Combustible Control of Combustibles
* ACP 1412.3, Control of Ignition Sources* ACP 1412.4, Impairments to Fire Protection Systems* MA-AA-100-1008, Station Housekeeping and Material Control* WM-AA-1 000, Work Activity Risk Management
* Site Fire Plan Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -Radioactive Release 1 DAEC RAI RR 1 For those compartments in LAR Attachment E -Radioactive Release Transition, that are identified as areas where gaseous radioactive effluents (caused by fire fighting activities -excluding the fire itself) would not be contained (areas without provision for radiation monitor detection capability with automatic closure to isolate the gaseous effluent), or where liquid effluents would be generated, provide a bounding analysis, qualitative analysis, quantitative analysis, or other analysis that demonstrates that the amount of radioactive effluent from the fire fighting activities will meet the gaseous effluent dose rate limits and the liquid effluent concentration limits specified in the plant's Technical Specifications (TS).a. If a qualitative analysis is being performed, provide information on: i. Type of fire most likely to occur in that fire area (e.g., electrical, transient combustibles, fuel)ii. Type and amount of radioactive contamination in the fire area (e.g., particulate, gas, iodine)iii. Type of fire suppression (e.g., water, foam, halon , C02)iv. Duration of anticipated fire fighting activities
: v. Anticipated amount of water to be generated vi. Capability of sumps and tanks to contain the estimated amount of water to be generated vii. Potential use of smoke educators and the impact of the exhaust as a new release path to the environment
: b. For a bounding analyses or a quantitative analysis, estimate the effluent concentrations discharged to the unrestricted area and demonstrate the doses rate limits of the TS are met.RESPONSE: A qualitative analysis was not performed.
A bounding analysis was performed for the YARD-RCA compartment as it is the only compartment where the gaseous and liquid radioactive effluent would be generated and not be contained (either through ventilation controls or liquid containment.)
Other compartments may have paths leading to the exterior but have engineering controls (ventilation, drainage, etc.) and administrative guidance through pre-fire plans, fire brigade training, and standard operating procedures to identify paths and stress actions and methods to prevent potentially contaminated materials from escaping a building.Rev A.Page 1 of 2 RAI -Radioactive Release 1 DAEC performed a bounding analysis for the YARD-RCA compartment (Radiological Engineering Calculation No. 12-002C.)
The bounding analysis was a worst-case calculation using the source term of a dry active waste (DAW) shipment trailer contents consumed by a fire. The analysis evaluated the effects of fire fighting activities (no containment of gaseous effluent (smoke) and liquid effluent (hose stream flow) on the limits set forth in the DAEC Offsite Dose Assessment Manual (ODAM) which incorporates the limits identified in the DAEC Technical Specifications (Section 5.5.4).The analysis determined that no ODAM (Technical Specification) limits would be exceeded.Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -Radioactive Release 2 DAEC RAI RR 2 Liquid Effluents 
-For those areas where containment is not specifically engineered (e.g., concrete floors, walls, sumps, tanks), describe other methods in the fire pre-plans that are used to provide containment (e.g., spill control kits, temporary dikes, storm drain covers, settling ponds etc.).RESPONSE: The DAEC Report "DAEC NFPA 805 Radioactive Release Review" identified actions necessary for compliance with the requirements of NFPA 805 and the guidance in NEI 04-02 and RG 1.205 with respect to the radioactive release requirements.
One of these actions will be to provide and stage drain covers, diversion equipment, or other acceptable means to prevent water runoff from certain areas from entering the storm drain system in the event of a fire.Not all storm drains will require administrative fire brigade responder actions for containment.
The storm drains in some areas of the plant (in particular near the Radwaste Storage Tank and Low Level Radwaste Processing Facility) route to a retention pond which acts as the containment mechanism to prevent release to the Cedar River.In addition, there is an action to develop a standard operating procedure to prevent radioactive release in event of a fire. The document will stress actions to prevent escape of potentially contaminated materials from a building or area boundary.Additional guidance will be provided for fires in yard areas and locations with limited or no engineering controls.
Lastly, the Area Fire Plans and Fire Brigade Training will be enhanced to provide additional guidance for fires in yard areas and locations with limited or no engineering controls for actions/methods for preventing potential radiological release.The fire pre-plans (known as Area Fire Plans at DAEC) will be enhanced as part of the NFPA 805 implementation process as identified in DAEC LAR Table S-2 Implementation Item #14 as clarified in DAEC RAI FP 6.Rev A.Page 1 of I RAI -Radioactive Release 3 DAEC RAI RR 3 For those compartments in LAR Attachment E -Radioactive Release Transition, that make a conclusion that the radioactive release will not exceed the limits of NFPA 805, provide justification as to how the release will also not exceed the instantaneous release rate limits in the plant's TS (reference FAQ-09-0056).
RESPONSE: The DAEC radioactive release review documented in LAR Attachment E made the conclusion statement "Using installed engineering controls combined with pre-fire plans, training and procedures provides reasonable assurance that fire suppression activities will not cause a radioactive release that exceeds the requirements of NFPA 805, 2001 edition." for each compartment.
This conclusion is accurate in that engineering and administrative controls for compartments with containment are not anticipated to have an unmonitored release, therefore inherently meeting the radioactive release performance criteria of NFPA 805.Per FAQ 09-0056 Closure Memo (ML102920405), compliance with the radioactive release goals, objectives, and performance criteria can be demonstrated by review of engineering controls to ensure containment of gaseous and liquid effluents.
Otherwise, provide an analysis that demonstrates the limitations for instantaneous release of radioactive effluents specific in each unit's Technical Specifications are met. The NRC staff position is that the limitations in a licensee's Technical Specifications are structured to maintain the 10 CFR Part 20 limits and meet the NFPA 805 Radioactive Release Performance Criteria.The only DAEC compartment that does not rely on engineering controls for containment of gaseous and liquid effluents is the YARD-RCA.
Therefore, the compartments with engineering controls for containment of gaseous and liquid effluents are not required to have an analysis to demonstrate the limitations for release are met.DAEC performed a bounding analysis for the YARD-RCA compartment (Radiological Engineering Calculation No. 12-002C).
The bounding analysis was a worst-case calculation using the source term of a dry active waste (DAW) shipment trailer contents consumed by a fire. The analysis evaluated the effects of fire fighting activities on the limits set forth in the DAEC Offsite Dose Assessment Manual (ODAM) which incorporates the limits identified in the DAEC Technical Specifications (Section 5.5.4).The analysis determined that no ODAM (Technical Specification) limits are exceeded.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 1 DAEC RAI PRA 1 Numerous Facts and Observations (F&Os) (5-1, 5-2, 5-3, 5-4, 5- 5, 5-6 and 5-15)discuss the lack of documentation of the review of cutsets and other outputs of the Fire Probabilistic Risk Analysis (FPRA) model to provide assurance that the FPRA logic model is accurate and producing the intended results. The dispositions for these F&Os state that Sections 5.2 and 5.3 of the Fire Quantification Report have been updated to document these reviews. During the audit, the licensee informed the staff that two non-logical cutsets had been discovered in the top core damage frequency (CDF) cutsets and that their elimination (along with other similar cutsets) would result in a significant reduction in both total core damage frequency (CDF) and large early release frequency (LERF). It was stated that these erroneous cutsets occurred as the unintended consequence of a logic transfer in the FPRA. Address the following:
: a. Provide further information on this issue and its cause.b. Review the license amendment request (LAR) submittal and provide updated LAR information impacted or changed by the corrected FPRA model. Discuss what had changed in the model to result in the revised LAR information.
: c. Address the broader implication of this discovery in terms of quality of the review of FPRA results and steps taken, such as new reviews of important and non-important cutsets, sequences and other results, to provide assurance that other similar errors do not exist in the current model.d. Confirm that RAI responses which utilize the FPRA model used the corrected model.RESPONSE: a. The ASME/ANS standard and guidance in NUREG/CR-6850 recognize the challenges in selecting a specific initiator for each postulated fire scenario.
The DAEC approach relies on using the general transient event tree to propagate fire induced failures through appropriate accident sequences.
Some end states from the general transient event tree are transferred to other trees to further refine impacts of mitigating system successes and failures.
An unexpected interaction between event tree transfer logic and DC power model logic resulted in generation of erroneous cutsets.Specifically, model logic that transfers induced loss of offsite power (LOOP) sequences to the station blackout (SBO) event tree and model logic for DC power availability during SBO were constructed differently.
In the FPRA quantification, this difference allowed a fire induced LOOP event to be transferred to the SBO model logic at the same time DC power availability was evaluated using the general transient event tree. Since assumptions for DC power availability are different for SBO than for general transients, invalid cutsets were generated.
Rev A. Page 1 ofI3 Rev A.Page I of 13 RAI -PRA 1 The FPRA model was corrected by adding mutually exclusive logic to prevent the differing assumptions for the DC power system from being considered at the same time.The error was confirmed to not be present in the MPIE PRA model and so the correction was not applied to it.b. The quantification of the updated FPRA results in changes to the FPRA results in Attachment W of the LAR. The attached tables supersede LAR Table W-1 through W-4.c. The reviews performed for the FPRA ensure the risks associated with postulated fires are properly reflected in the FPRA model. These reviews are performed for fire areas, fire zones, fire scenarios, accident classes, accident sequences, cutsets, and basic events. Each review determines whether the model adequately reflects the postulated fire damage to equipment and cables and whether it adequately reflects random failure of equipment available to mitigate the fire event. For the NFPA 805 application, additional reviews were performed to ensure each VFDR was properly reflected in the model. Additionally, the results were reviewed to identify areas of high risk and potential resolutions to reduce the risk.The FPRA was requantified with the model change described above and was reviewed to ensure the results properly reflect single fire sequences and that the erroneous cutsets were removed. Additionally, review of the new results did not identify any additional anomalies.
Specifically regarding the erroneous cutsets and review, the results for the essential switchgear rooms reflected the expected fire damage consistent with the equipment relied upon in the NSCA. The risk significance of the fire scenarios is consistent with DAEC plant design and operation.
Additionally, the high risk associated with the fire scenarios is not related to a VFDR; therefore, the model correction does not change the results of risk input to the LAR.d. The RAIs that require quantification results of the FPRA were completed after requantification with the corrected model (i.e., PRA RAIs 1, 5, 7, 8, 12, 20, 35, 50, and 51).Rev A. Page 2 of 13 Rev A.Page 2 of 13 RAI -PRA I Table 1 Summary of Risk Significant CDF Fire Scenarios (CDF Contribution 
> 1.0%)(LAR Table W-1 Replacement)
Scenario Description Contribution 1)I Risk Insights FIF(1) CCDP(1) CDF()I OF F40 1A3 -4160V HEAF -Cub. 301-302 10.2% Fire scenario 1OF F40 is a high energy arching fault fire at the 1A3 5.88E-05 7.55E-02 4.44E-06 essential switchgear that is postulated to damage the adjacent cubicles and the cable tray above. The scenario results in a loss of offsite power given damage to the lX3 and 1X4 protective relaying cables. 1A4 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1G21. Offsite power recovery is not credited given the fire damage.10E F56 1A4 -HEAF -Cub. 401-402 9.8% Fire scenario 10E F56 is a high energy arching fault fire at the 1A4 5.88E-05 7.26E-02 4.27E-06 essential switchgear that is postulated to damage the adjacent cubicles and the cable tray above. The scenario results in a loss of offsite power given damage to the IX3 and 1X4 protective relaying cables. 1A3 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1G31. Offsite power recovery is not credited given the fire damage.1OF F14 1 B3 -480V LC Fire -Target Damage 5.7% Fire scenario 1OF F14 is a fire at the 1B3 480V Load Center which is 2.14E-04 3.45E-02 2.50E-06 postulated to damage cables in the 98th percentile heat release rate zone of influence.
The scenario results in complete loss of Division 1 equipment coupled with maintenance or random failure of Division 2 equipment.
I0E FO0 1 D44 -BC Fire -Target Damage -Full ZOI 5.5% Fire scenario 10E FO0 is a fire at the 1D44 battery charger which is 1.90E-04 7.40E-02 2.39E-06 postulated to damage cables in the 98th percentile heat release rate zone of influence.
The scenario results in a loss of offsite power given damage to the 1X3 and 1X4 protective relaying cables. 1A3 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1G31. Offsite power recovery is not credited given the fire damage. Additionally, HPCI and RCIC are unavailable given fire damage to cables.1OE F76 1 D44 -BC Fire -Target Damage -First Set 2.7% Fire scenario 10E F76 is a fire at the 1D44 battery charger which is 1.90E-04 7.70E-02 1.17E-06 postulated to damage cables in the first overhead cable tray. The scenario results in a loss of offsite power given damage to the IX3 and 1X4 protective relaying cables. 1A3 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1 G3 1. Offsite power recovery is not credited given the fire damage.Rev A.Page 3 of 13 RAI -PRA I Table 1 Summary of Risk Significant CDF Fire Scenarios (CDF Contribution 
> 1.0%)(LAR Table W-1 Replacement)
Scenario Description Contribution' ) Risk Insights FIF 1) CCDP(1) CDF(')10E F46 1 B4 -480V LC Fire -Target Damage 2.0% Fire scenario 10E F46 is a fire at the I B4 load center which is postulated 2.14E-04 1.21E-02 8.82E-07 to damage targets in the 9 8gh percentile heat release rate zone of influence.
Fire induced damage results in loss of Division 2 equipment.
Dominate sequences include loss of room cooling scenarios.
Additionally, the MSIVs have target cables in the scenario and operator failure to manually close the MSIVs from the MCR is postulated to result in a large steam LOCA.10F F07 1A3 Cub. 307 -4160V SWGR Fire -Target 1.9% Fire scenario 10F F07 is a panel fire at the IA3 essential switchgear 7.09E-05 8.57E-02 8.50E-07 Damage cubicle 307 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario results in a loss ofoffsite power given damage to the 1X3 and 1X4 protective relaying cables. 1A4 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1 G2 1.Offsite power recovery is not credited given the fire damage.10E F05 1XL80 -XFMR Fire -Target Damage 1.8% Fire scenario 10E F05 is a fire at the 1XL80 transformer that damages the 2.19E-04 7.64E-03 7.86E-07 cables in the above trays. The scenario results in complete loss of Division 2 equipment coupled with maintenance or random failure of Division I equipment.
1OF F12 1A3 Cub. 312 -4160V SWGR Fire -Target 1.8% Fire scenario 10F F12 is apanel fire at the IA3 essential switchgear 7.09E-05 7.88E-02 7.82E-07 Damage cubicle 312 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario IOF F07.1OF FO IA3 Cub. 301 -4160V SWGR Fire -Target 1.7% Fire scenario 1OF FOI is a panel fire at the 1A3 essential switchgear 7.09E-05 7.61E-02 7.55E-07 Damage cubicle 301 that is postulated to damage the adjacent cubicles and the cable tray above.IOF F02 1A3 Cub. 302 -4160V SWGR Fire -Target 1.7% Fire scenario IOF F02 is a panel fire at the 1A3 essential switchgear 7.09E-05 7.61E-02 7.55E-07 Damage cubicle 302 that is postulated to damage the adjacent cubicles and the cable tray above.IOF F03 1A3 Cub. 303 -4160V SWGR Fire -Target 1.7% Fire scenario ]OF F03 is a panel fire at the IA3 essential switchgear 7.09E-05 7.61E-02 7.55E-07 Damage cubicle 303 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario 1OF F07.IOF F04 1A3 Cub. 304 -4160V SWGR Fire -Target 1.7% Fire scenario IOF F04 is a panel fire at the 1A3 essential switchgear 7.09E-05 7.55E-02 7.50E-07 Damage cubicle 304 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario 1OF F07.1OF F05 1 A3 Cub. 305 -4160V SWGR Fire -Target 1.7% Fire scenario I OF F05 is a panel fire at the 1 A3 essential switchgear 7.09E-05 7.55E-02 7.50E-07 Damage cubicle 305 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario IOF F07.Rev A.Page 4 of 13 RAI -PRA I Table 1 Summary of Risk Significant CDF Fire Scenarios (CDF Contribution 
> 1.0%)(LAR Table W-1 Replacement)
Scenario Description Contribution(1)
I Risk Insights FIF( ) CCDP(1) CDF(')1OF F06 IA3 Cub. 306 -4160V SWGR Fire -Target 1.7% Fire scenario IOF F06 is a panel fire at the IA3 essential switchgear 7.09E-05 7.55E-02 7.50E-07 Damage cubicle 306 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario 1 OF F07.IOF F08 1A3 Cub. 308 -4160V SWGR Fire -Target 1.7% Fire scenario IOF F08 is a panel fire at the 1A3 essential switchgear 7.09E-05 7.55E-02 7.50E-07 Damage cubicle 308 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario 1OF F07.1OF F09 1A3 Cub. 309 -4160V SWGR Fire -Target 1.7% Fire scenario I OF F09 is a panel fire at the 1A3 essential switchgear 7.09E-05 7.55E-02 7.50E-07 Damage cubicle 309 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario IOF F07.1OF F38 IA3 -4160V SWGR Fire -No Target 1.7% Fire scenario 1OF F38 is a panel fire at the 1A3 essential switchgear 7.12E-05 1.18E-02 7.22E-07 Damage -Cub. 302 cubicle 302 that damages the cables in the cubicle only. The scenario results in complete loss of Division 1 equipment coupled with maintenance or random failure of Division 2 equipment.
IOE F51 1A4 Cub. 401 -4KV MVSG Fire -Target 1.6% Fire scenario 10E F51 is a panel fire at the IA4 essential switchgear 7.09E-05 7.21E-02 7.16E-07 Damage cubicle 401 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario results in a loss of offsite power given damage to the 1X3 and 1X4 protective relaying cables. 1A3 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1G3 1.Offsite power recovery is not credited given the fire damage.10E F77 1A4 Cub. 402 -4KV MVSG Fire -Target 1.6% Fire scenario 1OE F77 is a panel fire at the 1A4 essential switchgear 7.09E-05 7.21E-02 7.16E-07 Damage cubicle 402 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario 1 OE F5 1.10EF75 IA4 Cub. 405-412- 4KV MVSG Fire -1.4% Fire scenario 10E F75 is a fire at the 1 A4 essential switchgear that does 5.70E-04 7.89E-03 6.29E-07 Target Damage not result in a fire induced LOOP. Fire induced damage results in loss of Division 2 equipment.
Dominate sequences include loss of room cooling scenarios.
The scenario results generally in transient sequences with random failures of Division I equipment resulting in core damage.1OF F16 1X31 -1B03 XFMR Fire -Target Damage 1.4% Fire scenario IOF F16 is a fire at the 1B3 480V Load Center IX31 2.19E-04 8.40E-03 6.24E-07 transformer which is postulated to damage cables in the 98th percentile heat release rate zone of influence.
The scenario results in complete loss of Division 1 equipment coupled with maintenance or random failure of Division 2 equipment.
Rev A. Page 5 of 13 Rev A.Page 5 of 13 RAI -PRA 1 Table 1 Summary of Risk Significant CDF Fire Scenarios (CDF Contribution 
> 1.0%)(LAR Table W-1 Replacement)
Scenario Description Contribution(1) Risk Insights FIF(1) CCDP(1) CDF()1OF F19 1 D43 -BC Fire -Target Damage 1.4% Fire scenario IOF F19 is a fire at the 1 D43 Battery Charger that damages 1.90E-04 7.58E-03 6.05E-07 the cables in the above trays. The scenario results in complete loss of Division I equipment coupled with maintenance or random failure of Division 2 equipment.
1OF F34 1A3 -4160V SWGR Fire -No Target 1.3% Fire scenario IOF F34 is a fire at the 1A3 essential switchgear that does 7.12E-04 9.43E-04 5.77E-07 Damage -Cub. 303 -312. not result in a fire induced LOOP. Fire induced damage results in the loss of the switchgear and Division I equipment. 
.The scenario results generally in transient sequences with random failures of Division I equipment resulting in core damage.I IA AOl Bounding Fire -CDF 1.3% Fire scenario 1 IA AOl is a bounding fire in the Cable Spreading Room. 5.70E-06 L.OOE+00 5.70E-07 Detailed analysis is not performed given the lack of fixed ignition sources and the large amount of cables. A transient fire is considered low frequency given the DAEC maintenance procedures, hot work procedures, and transient combustible control procedures.
A bounding fire considers complete loss of Division 2 equipment.
10F F57 1A3 Cub. 310 -4160V SWGR Fire -Target 1.1% Fire scenario 1OF F57 is a panel fire at the 1A3 essential switchgear 7.09E-05 7.55E-02 4.82E-07 Damage -First Set cubicle 310 that is postulated to damage the adjacent cubicles and the first cable tray above. The scenario impacts are similar to scenario IOF FO0.10F F58 1A3 Cub. 311 -4160V SWGR Fire -Target 1.1% Fire scenario 1OF F58 is a panel fire at the IA3 essential switchgear 7.09E-05 7.55E-02 4.82E-07 Damage -First Set cubicle 311 that is postulated to damage the adjacent cubicles and the first cable tray above. The scenario impacts are similar to scenario IOF FOI.1OF FI 1 1A3 Cub. 311 -4160V SWGR Fire -Target 1.1% Fire scenario 1OF Fl1 is a panel fire at the IA3 essential switchgear 7.09E-05 1.35E-01 4.78E-07 Damage -Full ZOI cubicle 311 that is postulated to damage the adjacent cubicles and both cable trays above. The scenario impacts are similar to scenario 1OF FOI with the additional fire induced failure of HPCI and RCIC.1OF FIO 1A3 Cub. 310 -4160V SWGR Fire -Target 1.1% Fire scenario lOF FlO is a panel fire at the 1A3 essential switchgear 7.09E-05 1.34E-01 4.74E-07 Damage -Full ZOI cubicle 310 that is postulated to damage the adjacent cubicles and both cable trays above. The scenario impacts are similar to scenario I OF FO 1 with the additional fire induced failure of HPCI and RCIC.Rev A. Page 6 of 13 Rev A.Page 6 of 13 RAI -PRA 1 Table 2 Summary of Risk Significant LERF Fire Scenarios (LERF Contribution 
> 1.0%)(LAR Table W-2 Replacement)
Contribution Scenario Description (1)Risk Insights FIF() CLERP ) LERF JOE FOI 1D44 -BC Fire -Target Damage -Full ZOI 14.0% Fire scenario I1E FO1 is a fire at the 1D44 battery charger which is 1.90E-04 6.91E-02 2.23E-06 postulated to damage cables in the 98th percentile heat release rate zone of influence.
The scenario results in a loss of offsite power given damage to the 1X3 and IX4 protective relaying cables. 1A3 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator 1 G31. Offsite power recovery is not credited given the fire damage. Additionally, HPCI and RCIC are unavailable given fire damage to cables. Alternate depressurization methods are not credited post core damage resulting in a failure to arrest the core melt in vessel and subsequent large early release.1OF F14 I B3 -480V LC Fire -Target Damage 11.8% Fire scenario 1OF F14 is a fire at the 1B3 480V Load Center which is 2.14E-04 2.58E-02 1.88E-06 postulated to damage cables in the 98th percentile heat release rate zone of influence.
The scenario results in complete loss of Division I equipment coupled with maintenance or random failure of Division 2 equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.10FF12 1A3 Cub. 312 -4160V SWGR Fire -Target 4.4% Fire scenario 10F F]2 is a panel fire at the 1A3 essential switchgear 7.09E-05 7.04E-02 6.99E-07 Damage cubicle 312 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario IOF F07.1CE F51 1A4 Cub. 401 -4KV MVSG Fire -Target 4.2% Fire scenario ICE F51 is a panel fire at the 1A4 essential switchgear 7.09E-05 6.73E-02 6.68E-07 Damage cubicle 401 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario results in a loss ofoffsite power given damage to the IX3 and IX4 protective relaying cables. IA3 Essential switchgear is unavailable due to the loss of offsite power coupled with maintenance or random failures of the Standby Diesel Generator I G31.Offsite power recovery is not credited given the fire damage. Alternate depressurization methods are not credited post core damage resulting in a failure to arrest the core melt in vessel and subsequent large early release.ICE F77 IA4 Cub. 402 -4KV MVSG Fire -Target 4.2% Fire scenario ICE F77 is a panel fire at the 1 A4 essential switchgear 7.09E-05 6.73E-02 6.68E-07 Damage cubicle 402 that is postulated to damage the adjacent cubicles and the cable tray above. The scenario impacts are similar to scenario I CE F51 Rev A. Page 7 of 13 Rev A.Page 7 of 13 RAI -PRA I Table 2 Summary of Risk Significant LERF Fire Scenarios (LERF Contribution 
> 1.0%)(LAR Table W-2 Replacement)
Contribution Scenario Description tc s h Risk Insights FIF(1) CLERP(1) LERF(1)10E F46 1 B4 -480V LC Fire -Target Damage 3.7% Fire scenario 10E F46 is a panel fire at the 1 B4 480V load center that 2.14E-04 8.11E-03 5.89E-07 damages the cables in the above trays. The scenario results in complete loss of Division 2 equipment coupled with maintenance or random failure of Division 1 equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.tOE F05 1XL80 -XFMR Fire -Target Damage 2.7% Fire scenario 1OE F05 is a fire at the IXL80 transformer that damages the 2.19E-04 4.13E-03 4.24E-07 cables in the above trays. The scenario results in complete loss of Division 2 equipment coupled with maintenance or random failure of Division I equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.1OE F75 1A4 Cub. 405-412- 4KV MVSG Fire -2.3% Fire scenario IE F75 is a panel fire at the 1A4 cubicles 405-412 that 5.70E-04 4.56E-03 3.64E-07 Target Damage damages the cables in the above trays. The scenario results in complete loss of Division 2 equipment coupled with maintenance or random failure of Division I equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.1OF FI 9 1D43 -BC Fire -Target Damage 2.1% Fire scenario IOF F19 is a fire at the 1D43 Battery Charger that damages 1.90E-04 4.25E-03 3.39E-07 the cables in the above trays. The scenario results in complete loss of Division 1 equipment coupled with maintenance or random failure of Division 2 equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.IE F45 1X41 -XFMR Fire -Target Damage 2.1% Fire scenario I0E F45 is a transformer fire at the IB4 480V load center 8.75E-04 6.40E-03 3.36E-07 1X41 transformer that damages the cables in the above trays. The scenario results in complete loss of Division 2 equipment coupled with maintenance or random failure of Division I equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.Rev A. Page 8 of 13 Rev A.Page 8 of 13 RAI -PRA 1 Table 2 Summary of Risk Significant LERF Fire Scenarios (LERF Contribution 
> 1.0%)(LAR Table W-2 Replacement)
Contribution Scenario Description Risk Insights FIF(') CLERP(1) LERF )1OF F16 1X31 -IB03 XFMR Fire -Target Damage 2.0% Fire scenario 1OF F16 is a fire at the 1B3 480V Load Center IX31 2.19E-04 4.29E-03 3.19E-07 transformer which is postulated to damage cables in the 98th percentile heat release rate zone of influence.
The scenario results in complete loss of Division 1 equipment coupled with maintenance or random failure of Division 2 equipment.
Post core damage, core melt is not arrested in vessel, combustible gas venting fails, drywell shell fails, and the Reactor Building in ineffective resulting in a large early release.IOE F03 1 D22 -BC Fire -Target Damage 1.6% Fire scenario 10E F03 is a 1 D22 battery charger fire that results in fire 1.90E-04 4.48E-03 2.56E-07 damage to targets in the 98'h percentile heat release rate zone of influence.
The scenario results in transient sequences without high pressure injection.
HPCI and RCIC are assumed failed due to DC power supply failures.Core damage at high pressure without injection results in drywell shell failure.1OF FI I A3 Cub. 311 -4160V SWGR Fire -Target 1.4% Fire scenario 1OF FlI is a panel fire at the 1A3 essential switchgear 7.09E-05 6.25E-02 2.22E-07 Damage -Full ZOI cubicle 311 that is postulated to damage the adjacent cubicles and both cable trays above. The scenario impacts are similar to scenario 1OF FOI with the additional fire induced failure of HPCI and RCIC.1OF FlO 1A3 Cub. 310 -4160V SWGR Fire -Target 1.4% Fire scenario 1OF F10 is a panel fire at the 1A3 essential switchgear 7.09E-05 6.17E-02 2.19E-07 Damage -Full ZOI cubicle 310 that is postulated to damage the adjacent cubicles and both cable trays above. The scenario impacts are similar to scenario IOF FO1 with the additional fire induced failure of HPCI and RCIC.1OF F36 I B32 -480V MCC Fire 1.4% Fire scenario IOF F36 is a fire at the 1 B32 MCC results in fire damage to 5.70E-04 3.82E-04 2.17E-07 the MCC and loss of Division I equipment.
The scenario results in transient sequences without high pressure injection.
HPCI and RCIC are assumed failed due to DC power supply failures.
Core damage at high pressure without injection results in drywell shell failure.10F F38 1A3 -4160V SWGR Fire -No Target 1.3% Fire scenario 10F F38 is a fire in the IA3 essential switchgear cubicle 302. 7.12E-05 3.45E-03 2.11E-07 Damage -Cub. 302 The fire results in a loss of the switchgear and Division 1 equipment.
The scenario results in a loss oftorus cooling due to random failures in Division 2 equipment.
Containment is successfully vented; however injection post containment venting is not successful.
IOE F52 11B42 480V MCC Fire 1.2% Fire scenario I0E F52 is a 1B42 MCC fire that results in fire damage to the 4.98E-04 3.97E-04 1.98E-07 MCC and loss of Division 2 equipment.
The scenario results in transient sequences without high pressure injection.
HPCI and RCIC are assumed failed due to DC power supply failures.
Core damage at high pressure without injection results in drywell shell failure.Rev A.Page 9 of 13 RAI -PRA 1 Table 2 Summary of Risk Significant LERF Fire Scenarios (LERF Contribution 
> 1.0%)(LAR Table W-2 Replacement)
Contribution Scenario Description C tb iRisk Insights FIF(1) CLERP(') LERF ()IOE F50 Bus Duct HEAF 1.2% Fire scenario 1OE F50 is a bus duct HEAF that results in fire damage to 5.24E-05 3.78E-03 1.98E-07 targets in the zone of influence.
The scenario results in transient sequences without high pressure injection.
HPCI and RCIC are assumed failed due to DC power supply failures.
Core damage at high pressure without injection results in drywell shell failure.I OF F34 1 A3 -4160V SWGR Fire -No Target 1.1% Fire scenario 1 OF F34 is a fire in the 1A3 essential switchgear cubicles.
7.12E-04 2.95E-04 1.81E-07 Damage -Cub. 303 -312. The fire results in a loss of the switchgear and Division I equipment.
The scenario results in a loss of low pressure injection sequences due to random failures of Division 2 equipment.
Core damage at low pressure without injection results in drywell shell failure.10E F04 IG-61 -MG Fire -Target Damage 1.1% Fire scenario 1OE F04 is a fire at the 1G61 MG set that results in fire 3.88E-04 3.29E-03 1.79E-07 damage to targets in the 98th percentile zone of influence.
The scenario results in transient sequences without high pressure injection.
HPCI and RCIC are assumed failed due to DC power supply failures.
Core damage at high pressure without injection results in drywell shell failure.IOD AOl Bounding Fire 1.1% Fire scenario 1OD AOl is a fire in the ID1 battery room. The fire scenario 5.02E-04 3.52E-04 1.76E-07 results in loss of high pressure injection sequences.
HPCI and RCIC are assumed lost due to DC power failure. Core damage at high pressure without injection results in drywell shell failure.IOF F47 1D120 -BC Fire -No Target Damage 1.1% Fire scenario 1OF F47 is a fire at the 1D120 battery charger. The scenario 1.90E-04 1.1 1E-03 1.75E-07 results in loss of high pressure injection sequences.
Random failures in DC power supplies results in assumed loss of HPCI and RCIC for high pressure injection.
Core damage at high pressure without injection results in drywell shell failure.I IA A02 Bounding Fire -LERF 1.1% Fire scenario I IA A02 is a bounding fire in the Cable Spreading Room. 5.70E-06 1.OOE+00 1.71E-07 Detailed analysis is not performed given the lack of fixed ignition sources and the large amount of cables. A transient fire is considered low frequency given the DAEC maintenance procedures, hot work procedures, and transient combustible control procedures.
A bounding fire considers complete loss of Division 2 equipment.
I OF F25 1Y4 (JS401) -XFMR Fire -Target Damage 1.1% Fire scenario I OF F25 is a fire in the 1Y4 transformer resulting in fire 2.19E-04 4.29E-03 1.69E-07 damage to targets in the 98"' percentile heat release rate zone of influence.
The fire results in a loss of Division 1 equipment.
The scenario results in loss of low pressure injection sequences.
Random failures in DC power supplies results in assumed loss Division 2 low pressure injection.
Core damage at low pressure without injection results in drywell shell failure.Rev A.Page 10 of 13 RAI -PRA I Table 3 Duane Arnold Energy Center / Fire Area Risk Summary (LAR Table W-3 Replacement)
Fire Area Description NFPA 805 Fire Area CDF/LERF VFDR RAs Fire Risk Eval Additional Risk of Area Basis (/yr)"' (Yes/No) A CDF/LERF (/yr)(1) RAs (/yr)(1)BA Buffer Area 4.2.3 2.42E-07/4.63E-08 No No N/A N/A CB 1 Control Building -Cable 4.2.4.2 1.37E-06/4.30E-07 Yes Yes 1.49E-08/1.49E-08 1.49E-08/1.49E-08 Spreading Room, Control Room, and Control Building HVAC Room CB2 Control Building -Div. 2 4.2.4.2 1.57E-05/7.68E-06 Yes No N/A N/A Essential Switchgear Room and 1D2 Battery Room CB3 Control Building -Div. 1 4.2.4.2 2.28E-05/6.45E-06 Yes No 1.43E-08/1.28E-08 N/A Essential Switchgear Room and 1 D1 Battery Room CB4 Control Building -Battery Room 4.2.4.2 9.22E-08/1.73E-08 Yes No 2.40E-lO/7.23E-12 N/A Corridor and 250 VDC Battery Room DRY Drywell Containment 4.2.3.4 N/A No No N/A N/A EXI Exterior Plant Areas 4.2.3.2 1.81 E-07/1.32E-07 No No N/A N/A i1s Intake Structure 
-Div. I Pump 4.2.3.2 3.49E-08/7.63E-09 No No N/A N/A Area IS2 Intake Structure 
-Div. 2 Pump 4.2.3.2 6.30E-08/1.37E-08 No No N/A N/A Area PHI Pump House -Div. 2 Pump Area 4.2.3.3(b) 4.07E-08/9.80E-09 No No N/A N/A PH2 Pump House -Div. I Pump Area 4.2.3.2 1.34E-07/2.88E-08 No No N/A N/A RBI Reactor Building -Northwest, 4.2.4.2 5.12E-07/2.83E-07 Yes No 8.49E-09/1.32E-09 N/A Southeast, Southwest Comer Rooms and 757' Elevation RB3 Reactor Building -786' Elevation 4.2.4.2 4.93E-07/2.94E-07 Yes No 7.28E-08/2.65E-08 N/A and above RB4 Reactor Building -Northeast 4.2.3.2 5.33E-09/8.76E-10 No No N/A N/A Comer Room Rev A.Page 11 of 13 RAI -PRA 1 Table 3 Duane Arnold Energy Center / Fire Area Risk Summary (LAR Table W-3 Replacement)
Fire Area Description NFPA 805 Fire Area CDF/LERF VFDR RAs Fire Risk Eval Additional Risk of Area Basis (/yr)(1) (Yes/No) A CDF/LERF (/yr)(1) RAs (/yr)(1)TB1 Turbine Building 4.2.4.2 1.94E-06/4.99E-07 Yes No 3.80E-10/1.IOE-10 N/A Total 4.36E-05/1.59E-05 1.11E-07/5.55E-08 1.49E-08/1.49E-08 Rev A. Page 12 of 13 Rev A.Page 12 of 13 RAI -PRA 1 Table 4 Summary of Risk Decrease Associated with Modifications (LAR Table W-4 Replacement)
Fire Area Modification Fire Area Risk Decrease Fire Area Area Description CDF/LERF (/yr)(1) (Yes/No) CDF/LERF Without Associated with Modification 
(/yr)(1) Modification 
(/yr)(1)BA Buffer Area 2.42E-07/4.63E-08 No 2.42E-07/4.63E-08 N/A CBI Control Building -Cable Spreading Room, Control 1.37E-06/4.30E-07 Yes 7.51E-06/1.93E-06 6.14E-06/1.50E-06 Room, and Control Building HVAC Room CB2 Control Building -Div. 2 Essential Switchgear Room 1.57E-05/7.68E-06 No 1.57E-05/7.68E-06 N/A and 1 D2 Battery Room CB3 Control Building -Div. 1 Essential Switchgear Room 2.28E-05/6.45E-06 No 2.28E-05/6.45E-06 N/A and IDI Battery Room CB4 Control Building -Battery Room Corridor and 250 9.22E-08/1.73E-08 No 9.22E-08/1.73E-08 N/A VDC Battery Room DRY Drywell Containment N/A No N/A N/A EXI Exterior Plant Areas 1.81E-07/1.32E-07 No 1.81E-07/1.32E-07 N/A ISI Intake Structure 
-Div. 1 Pump Area 3.49E-08/7.63E-09 No 3.49E-08/7.63E-09 N/A IS2 Intake Structure 
-Div. 2 Pump Area 6.30E-08/1.37E-08 No 6.30E-08/1.37E-08 N/A PHI Pump House -Div. 2 Pump Area 4.07E-08/9.80E-09 No 4.07E-08/9.80E-09 N/A PH2 Pump House -Div. 1 Pump Area 1.34E-07/2.88E-08 No 1.34E-07/2.88E-08 N/A RB1 Reactor Building -Northwest, Southeast, Southwest 5.12E-07/2.83E-07 No 5.12E-07/2.83E-07 N/A Comer Rooms and 757' Elevation RB3 Reactor Building -786' Elevation and above 4.93E-07/2.94E-07 No 4.93E-07/2.94E-07 N/A RB4 Reactor Building -Northeast Comer Room 5.33E-09/8.76E-10 No 5.33E-09/8.76E-10(2)
N/A TB 1 Turbine Building 1.94E-06/4.99E-07 Yes 2.74E-06/6.62E-07 8.04E-07/1.63E-07 Total 4.36E-05/1.59E-05 5.06E-05/I.76E-05 6.94E-06/1.66E-06 Rev A. Page 13 of 13 Rev A.Page 13 of 13 RAI -PRA 5 DAEC RAI PRA 5 F&O 3-7. For multi-element rated barriers, the probability of failure used in the multi-compartment analysis (MCA) was for the most bounding element in the barrier. Provide revised total/delta risk estimates that include all elements (see Table 11-3 of NUREG/CR-6850) providing pathways from one compartment to another or a justification that including the failure probability of all elements will not significantly impact the results.RESPONSE: The MCA was revised in response to RAI PRA 05. In the revised MCA, the probability of barrier failure was taken as the sum of all barrier elements.
For example, the probability of barrier failure was updated to include the probability of penetration failures, wall failures, damper failures, and door failures.
The revised MCA did not result in changes to the total and delta risk calculations.
For the MCA, multi-compartment fire scenarios still screened when considering each barrier element.Appendix C of the FPRA Fire Scenario Report will be updated to include each barrier element in the MCA.The MCA fire scenarios are screened from the analysis and are not risk significant due to the substantial barriers separating the fire zones at DAEC. Additionally, the divisional separation of equipment and cables at DAEC contributes to the negligible impact that multi-compartment scenarios have on the overall fire risk.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 6 DAEC RAI PRA 6 F&O 3-10. For PAU 02E and 02B, Table C-4 of the MCA justifies screening of PAU 02E and 02B by stating that "App N of NUREG/CR-6850 indicates only 1 (-1%) event caused structural damage beyond blowing doors open." Explain how this statement can be derived from Appendix N and, based on this derivation, provide further justification for screening this scenario.RESPONSE: Appendix N of NUREG/CR-6850 guidance was used in the assessment of hydrogen recombiner fires. PAU 02E is the hydrogen recombiner building.
The hydrogen recombiners are separated from PAU 02B by a 12 inch concrete wall with nonrated doors at each end of 02E. The FPRA MCA postulated a hydrogen fire that potentially damages targets in PAU 02B. The fire ignition frequency for PAU 02E is calculated to be 2.94E-2/yr, most of which is due to hydrogen recombiner fires. NUREG/CR-6850 guidance estimates that 10% of the hydrogen fires will damage components beyond the system. For these postulated fire scenarios, the frequency was adjusted by 10% to reflect the guidance in NUREG/CR-6850.
PAU 02E has no risk significant equipment or cables and the MCA interaction was assigned a screening CCDP of 0.1. Given the above assumptions the MCA interaction did not screen.NUREG/CR-6850 guidance indicates that further probability reduction factors may be applied based on the location of targets with respect to the location of postulated hydrogen fires; however, NUREG/CR-6850 does not provide guidance regarding further probability reduction factors and multi compartment interactions associated with hydrogen recombiner fires. These factors would include the probability of failure of the concrete wall and doors resulting in damage to targets in PAU 02B. NUREG/CR-6850 Section N.2.2 provides discussion of four hydrogen recombiner fire events. The qualitative statement in Table C-4 of the Fire Scenario Report regarding the one percent inappropriately describes insights from Section N.2.2 of NUREG/CR-6850.
A more appropriate assessment of the MCA interaction would be to apply a CCDP reflective of the potential failures given the hydrogen fire and fire barrier failure. DAEC evaluation FPE-B97-019 provides an evaluation of the fire barrier between 02E and 02B. The evaluation concludes that a fire originating in 02E is considered to propagate to 02B through the doors. When considering a hydrogen recombiner fire event, the fire is most likely to propagate to 02B through the east recombiner vault door and HVAC room door. The target set outside the HVAC room door in 02B does not contain FPRA cables. Therefore, the CCDP for 02E should be applied for the postulated fire scenario (i.e., 02E CCDP = 2E-6).Applying the CCDP reflective of the potential failures from a hydrogen recombiner fire results in a MCA interaction estimated CDF of 6E-9/yr (i.e., 2.94E-2
* 0.1
* 2E-6).The estimated CDF bounds the potential for the 12 inch concrete wall failure. The failure probability of the wall would be estimated using the NUREG/CR-6850 probability of 1 E-3.Rev A.Page 1 of 2 RAI -PRA 6 Targets in 02B adjacent to the wall include MCC 1 D41. The CCDP for the target set would be consistent with the MCC 1 D41 fire scenario (1 E-4). Given these probabilities, the MCA interaction considering concrete wall failure is a CDF of 3E-10/yr.With the use of the DAEC barrier evaluation and application of CCDPs reflective of potential damage given a hydrogen recombiner fire, the 02E to 02B MCA interactions result in a CDF that screen from the quantitative analysis in the FPRA.The MCA was revised in response to RAI PRA 05. As part of the MCA evaluation, the interaction between 02E and 02B was updated to reflect the corrections in the response to this RAI. Additionally, Appendix C of the Fire Scenario Report was updated.Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -PRA 7 DAEC RAI PRA 7 The MCA for diesel generator (DG) room fires (e.g., Exposing PAU 08H to Exposed PAU 08D) in Table C-4 of the Fire Scenario Report used a Severity Factor (SF) of 0.01 based on a review of the Fire Events Database (FEDB) which indicates that no DG fires damaged equipment beyond the DG room (0.5/49.5 
= 0.01). This SF was applied to a calculated CDF that already credited a Type 1 barrier (3 hour door) having a NUREG/CR-6850 recommended failure probability of 7.4E-03 (from Table C-3 of the Fire Scenario Report). Application of this SF in these scenarios appears to be double-counting credit already given for the barrier failure probability.
Provide revised total/delta risk results in which the SF of 0.01 is removed for those scenarios where this double-counting is applied. Describe the revised analysis and assumptions for these scenarios and, if applicable, provide justification for the use of change in risk results based on a SF/NSP less than one and/or subsequent qualitative screening.
RESPONSE: The MCA was revised in response to RAI PRA 05. In the revised MCA, the SF of 0.01 for the DG rooms (PAUs 08F and 08H) was removed. The MCA for the DG rooms instead included credit for the automatic suppression system which was identified as a credited suppression system in the License Amendment Request (ML1 1221A280)(LAR) (see Report Number 0027-0042-000-004, Duane Arnold Energy Center Fire Risk Evaluations).
In addition, the conservative estimate of conditional core damage probability was updated to be consistent with that of the turbine building including the ESW B pump control circuit modification.
Removal of the 0.01 SF and inclusion of a more appropriate conditional core damage probability resulted in the MCA fire scenarios still being screened from the quantification.
Appendix C of the FPRA Fire Scenario Report will be updated to reflect the changes in these MCA scenarios.
Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 8 DAEC RAI PRA 8 The results of the MCA were not included in the quantified CDF and LERF reported in the LAR. Provide revised total/delta CDF/LERF results by incorporating the MCA results consistent with the quantification as required by Supporting Requirement (SR) FSS-G6 Capability Category (CC)-II/III.
RESPONSE: The MCA was revised in response to RAI PRA 05, 06, and 07. Initially, five MCA fire scenarios did not screen from the MCA. In response to RAI PRA 06, MCA interaction between 02E and 02B was clarified and still screened from quantification.
In response to RAI PRA 07, MCA interaction for the DG rooms (MCA scenarios 08F-08D and 08H-08D) was clarified and screened from quantification.
The other two MCA fire scenarios (07C-07E and 07C-08C) were marginal in the MCA analysis for the LAR. However, the revised MCA applied the appropriate bounding turbine building conditional core damage probability with the ESW B pump control circuit modification.
Given this, these two MCA fire scenarios also screened from quantification.
Therefore, each postulated MCA fire scenario was screened from quantification.
Appendix C of the FPRA Fire Scenario Report, 0493080001.003, will be updated to reflect the revised MCA.Rev A. Page 1 of I Rev A.Page I of 1 RAI -PRA 12 PRA Question # 12 It was recently stated at the industry fire forum that the Phenomena Identification and Ranking Table (PIRT) Panel being conducted for the circuit failure tests from the DESIREE-FIRE tests may be eliminating the credit for Control Power Transformer (CPT) (about a factor 2 reduction) currently allowed by Tables 10-1 and 10-3 of NUREG/CR-6850 when estimating alternating current (AC) circuit failure probabilities.
Provide a sensitivity analysis that removes this CPT credit from the PRA and provide new results that show the impact of this potential change for CDF, LERF, A CDF, and A LERF.RESPONSE: The DAEC FPRA applied circuit failure mode conditional probabilities for components identified as risk significant throughout the numerous iterations of the FPRA. As such, use of values from the approved guidance in NUREG/CR-6850, Volume 2, Chapter 10, for circuits powered from control power transformers (CPTs) was sparingly applied in the FPRA. The CPT value is typically applied for motor operated valves. Table 4.0-1 of the FPRA Fire Scenario Report documents the circuit failure probabilities applied in the FPRA. Section 8.3 of the FPRA Quantification Report discusses application of circuit failure probabilities to risk significant components.
From Table 4.0-1 of the FPRA Fire Scenario Report, the circuit failure probability associated with a CPT was applied in fire scenario 02B-F01 (fire area RB1) and 03A-D12 (fire area RB3). For fire area RB1, the circuit failure probability was changed to reflect the values in NUREG/CR-6850 associated with circuits without a CPT. This change resulted in no noticeable change in CDF/LERF or A CDF/LERF.
For fire area RB3, removal of the CPT credit resulted in minor increases in the results. Fire area RB3 CDF increased from 4.93E-7/yr to 4.96E-7/yr and A CDF increased from 7.28E-8 to 7.61 E-8. Fire area RB3 LERF did not have a noticeable change and A LERF increased from 2.65E-8 to 2.70E-8. The negligible increase in RB3 fire risk given removal of credit for CPT does not change the risk input conclusions for the LAR.Rev A. Page 1 of I Rev A.Page I of 1 RAI -PRA 14 DAEC RAI PRA 14 F&O 2-20 and 5-29 appear to identify the following two deviations from NUREG/CR-6850 that have been applied in the cable spreading room (CSR). Provide a sensitivity analysis for the CSR that applies the guidance in NUREG/CR-6850 with no deviations.
: a. The analysis of the CSR applied a hot work pre-initiator factor of 0.01, which is an Unreviewed Analysis Method (UAM) or deviation from NUREG/CR-6850.
: b. The LAR Supplement apportions transient fire frequency in the CSR using a "Very Low" transient fire influencing factor for maintenance which is a deviation from NUREG/CR-6850.
RESPONSE: The CSR fire ignition frequency (FIF) is the focus of the use of a UAM involving a hot work pre-initiator factor for values reported in the License Amendment Request (ML1 1221A280) (LAR) and separately, the use of a "Very Low" influence factor category in the characterization of CSR fires and associated uncertainty evaluations in the CSR Fire Risk Assessment, FPRA report 0493080001.005.
Based on NUREG/CR-6850, FIFs and transient influence factor categories, the CSR FIF is 5.7E-4/yr (see Table 1 of the CSR Fire Risk Assessment).
This value, which is largely influenced by hot work fires, does not consider administrative controls in place at the DAEC regarding activities within the CSR. When the FIF is applied to the CSR event trees in Figures 1 through 3 of the CSR Fire Risk Assessment, the point estimate CDF from Figure 1 is calculated to be 3.3E-7/yr.
The upper bound is calculated to be 8.OE-6/yr and the lower bound is calculated to be 5.9E-9/yr.
Revised point estimate using NUREG/CR-6850 FIF and Figure 1 of the CSR assessment:
F -5.7E-4*0.1 
*0.5*0.87*0.01 
*1 *0.5*0.05=6.2E-9/yr G -5.7E-4*0.1 
*0.5*0.87*0.01 
*1 *0.5*1 =1.2E-7/yr K -5.7E-4*0.1 
*0.5*0.13*0.1 
*1 *0.5*0.1 =1.9E-8/yr L -5.7E-4*0.1 
*0.5*0.13*0.1 
*1 *0.5*1 =1.9E-7/yr Estimated point estimate sum total = 3.3E-7/yr Revised upper bound estimate using NUREG/CR-6850 FIF and Figure 2 of the CSR assessment:
F -5.7E-4*0.4*0.9*0.8*0.02*1 
*0.5*0.1 =1.6E-7/yr G -5.7E-4*0.4*0.9*0.8*0.02*1 
*0.5*1 =1.6E-6/yr K -5.7E-4*0.4*0.9*0.2*0.2*1*0.5*0.5=2.1 E-6/yr L -5.7E-4*0.4*0.9*0.2*0.2*1 
*0.5*1 =4.1 E-6/yr Estimated upper bound sum total = 8.OE-6/yr Rev A. Page 1 of 3 Rev A.Page 1 of 3 RAI -PRA 14 Revised lower bound estimate using NUREG/CR-6850 FIF and Figure 3 of the CSR assessment:
F -5.7E-4*0.01 
*0.1 *0.9*0.01 
*1 *0.5*0.05=1.3E-10/yr G -5.7E-4*0.01*0.1*0.9*0.01*1*0.5*1=2.6E-9/yr K -5.7E-4*0.01*0.1*0.1*0.1 
*1*0.5*0.1 
=2.9E-10/yr L -5j7E-4*0.01*0.1*0.1*0.1*1*0.5*1=2.9E-9/yr Estimated lower bound sum total = 5.9E-9/yr Application of revised upper bound estimates of ODF and LERF for the CSR to the sensitivity analysis in Section 3.2 of the CSR Fire Risk assessment does not change its conclusions.
The point estimate CDF of 3.3E-7/yr using the NUREG/OR-6850 FIF of 5.70E-4/yr is still less than the 5.7E-07/yr point estimate reported in the LAR, where the hot work pre-initiator factor was applied. Applying the CSR estimated upper bound ODF, the fire CDF would increase from approximately 5.7E-5/yr to 6.5E-5/yr.
The CSR upper bound LERF is estimated to be 30% of the CSR upper bound CDF, or 2.4E-6/yr.
Therefore, the fire LERF would increase from 8.9E-6/yr to 1.1 E-5/yr. The discussion in Section 3.2 of the CSR Fire Risk Assessment, FPRA report 0493080001.005 regarding the sensitivity analysis impact on the LAR remains valid.Additionally, since the time that the LAR was submitted, consensus regarding the UAM was reached by the Fire PRA Methods Review Panel. The consensus resulted in updated frequencies for hot work fires. For NUREG/OR-6850 Bin 5 and 6, the DAEO FPRA used a FIF of 1.48E-3/yr and 6.24E-3/yr, respectively.
The updated FIFs for Bin 5 and Bin 6 are 2.69E-4/yr and 3.53E-3/yr, respectively.
Using these updated frequencies, the DAEC OSR FIF would decrease from 5.70E-4/yr to 1.72E-4.FIF Bin Description Weighting DAEC FIF OSR FIF Factor 05 Cable fires caused by 2.85E-1 2.69E-4 7.67E-5 welding and cutting 06 Transient fires caused 1.96E-2 3.53E-3 6.92E-5 by welding and cutting 07 Transient 7.25E-3 3.57E-3 2.59E-5 Total 1.72E-4 Use of the updated FIFs results in an upper bound ODF of 5.9E-5/yr and an upper bound LERF of 9.6E-6/yr.
F -1,72E-4*0.4*0.9*0.8*0.02*1 
*0.5*0.1 =5.OE-8/yr G -1.72E-4*0.4*0.9*0.8*0.02*1 
*0.5*1 =5.OE-7/yr K -1.72E-4*0.4*0.9*0.2*0.2*1 
*0.5*0.5=6.2E-7/yr L -1.72E-4*0.4*0.9*0.2*0.2*1 
*0.5*1 =1.2E-6/yr Estimated upper bound sum total = 2.4E-6/yr Rev A.Page 2 of 3 RAI- PRA 14 Upper Bound CDF = 5.7E-5 + 2.4E-6 = 5.9E-5 Upper Bound LERF = 8.9E-6 + 0.3
* 2.4E-6 = 9.6E-6 Furthermore, on March 21, 2012 there was an NRC public meeting on transients and hot work. From the meeting, it was learned that an influence factor of zero was only intended for the occupancy influence factor. Therefore, procedural restrictions could be used as a basis for assigning a zero or a value less than one for the storage and maintenance influence factors. This conclusion further supports the use of a "Very Low" transient influence factor for maintenance activities in the CSR.Rev A. Page 3 of 3 Rev A.Page 3 of 3 RAI -PRA 15 DAEC RAI PRA 15 F&O 4-21. Section 5.2.1 of the FSR explains that the main control room (MCR) non-abandonment analysis credited an internal fire barrier or single metal wall (separating redundant divisions) for electrical cabinets 1(C06, 1C08, and 10C31 (essentially treating each of these cabinets as two different panels). The analysis assumed that a fire in one panel would not result in damage to components in the adjacent panel until 10 minutes after the peak heat release rate (HRR). This assumption is non-conservative since the 10 minutes is the guidance in Appendix S of NUREG/CR-6850 for damage to sensitive electronics when such equipment is protected from a cabinet fire with double walls with an air gap. Provide an estimate of the change in risk results either assuming no credit for the single metal wall, or justify any time delay that was used based on crediting the single metal wall.RESPONSE: Appendix S of NUREG/CR-6850 does not provide guidance regarding delay to fire damage in adjacent cabinets when a single wall is present. However, insights from Appendix S do confirm that delay to temperatures reaching damage criteria in adjacent cabinets is expected.
Additionally, the delay time is expected to depend on whether the cable is qualified or not.Section 11.5.2.8, Step 8.b, of NUREG/CR-6850 provides guidance related to use of Appendix S with modifications for control panels. For cabinets separated by a single wall and closed back, the guidance is to use the approach in Appendix L to establish the likelihood of fires in the exposing cabinet that could damage the wall. Then using Appendix S, a second non suppression probability is recommended based on a 15 minute fire duration.When applying NUREG/CR-6850 Appendix L in the case of a panel with an internal barrier, a zero distance may be most appropriate when determining the likelihood of fire damage. The zero distance probability estimated from Appendix L Figure L-1 is approximately 6E-03 for qualified cable. From FAQ 08-0050, the manual non suppression probability for control room fires at 15 minutes is 0.007. Therefore per the guidance of NUREG/CR-6850, the probability that an exposing control room cabinet fire results in fire damage to an adjacent cabinet when separated by a single wall is 4E-5 (6E-3
* 0.007).In the case of panels 10C06, 10C08, and 10C31, a single wall is used to separate qualified cables for redundant divisions within the panel.Given the above discussions, the panels could be treated using NUREG/CR-6850 to determine the likelihood of fire damage at zero distance without consideration of the single wall (i.e., 6E-3). However, this treatment is overly conservative given that additional protective features exist. Treating the panels consistent with Section 11.5.2.8 of NUREG/CR-6850 may provide too much credit for an internal single wall (i.e., 4E-5).For the FPRA, the probability of damage was estimated at 1 E-3.Rev A.Page 1 of 2 RAI -PRA 15 The discussion in the Fire Scenario Report inappropriately defines a delay time to damage using NUREG/CR-6850 Appendix S. Section 11.5.2.8 provides more appropriate guidance in modeling these cabinets.
Based on the guidance, a 1E-3 probability of fire damage given a single wall is a reasonable estimate.The Fire Scenario Report has been updated to include the discussion in this RAI response.For the FPRA, the fire scenarios crediting the single wall are 12A F05, 12A F07, and 12A F29 which had a total CDF of 1.3E-8/yr and a LERF of 7.8E-9/yr (from Table L-1 and Table M-1 of the FPRA Quantification Report, 0493080001.004, Rev. 3, respectively).
Applying the conservative probability of 6E-3 for fire damage at zero distance in place of the reasonable estimate of 1 E-3 would result in a factor of six increase of CDF to 7.8E-8/yr (1.3E-8/yr
* 6) and an increase of LERF to 4.7E-8/yr (7.8E-9/yr
* 6). Therefore, the change in overall plant risk would be negligible.
Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI PRA 18 DAEC RAI PRA 18 F&O HR-C1-01A.
It was concluded that the impact of evaluating pre-initiators at the system level instead of the train level was judged to have little or no impact on the results of the LAR. However, fire often consequentially fails one train, increasing the sensitivity of the results on the remaining train's availability.
Provide an assessment of the impact of using train level unavailability where ever system level values are currently used on the total/delta risk estimates developed for the LAR. It is noted that draft NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines", dated November 2009, states that existing internal events PRA pre-initiators do not need to be re-analyzed for the fire PRA. However, this presumes there are no unresolved F&Os on the issue. Appropriate resolution of this internal event PRA F&O is relevant to the NFPA 805 application since use of the system level HFE values could yield non-conservative results.RESPONSE: DAEC reviewed the pre-initiators to identify which were applied at the system level but not applied at the train level. Only the Emergency Diesel Generator Fuel oil level switch calibrations were applied as a pre-initiator as a system level pre-initiator but not applied as a train level pre-initiator.
A sensitivity analysis was completed with a best estimate Emergency Diesel Generator Fuel oil level switch train level failure rate. The sensitivity analysis demonstrated implementing the train level pre-initiator failure rates did not impact the results of the application including the total/delta risk estimates developed for the LAR.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -Probabilistic Risk Assessment 19 DAEC RAI PRA 19 According to discussions during the audit, there are procedures for a fire in the CSR that leads to Control Room (CR) evacuation upon equipment damage leading to loss of control from the CR. However, it was stated that CR evacuation from loss of control was not modeled in the fire PRA. There may be other fires outside of the CR that lead to CR evacuation because of loss of control or loss of habitability.
: a. Is CR evacuation from loss of control modeled in the FPRA. If not, why not?b. Clarify if fire scenarios initiated outside the MCR that may impact habitability in the CR were considered.
If not, why not?c. Identify any deviation from the guidance in NUREG/CR-6850 CR evacuation following both loss of control room function and loss of habitability.
RESPONSE: a. For all fire areas, DAEC evaluates loss of control of components as a result of postulated fire-induced damage for fire scenarios and impact on the total CDF/LERF for equipment not related to VFDRs and on delta CDF/LERF for components related to a VFDR. For the control room, the DAEC FPRA evaluated two scenarios for each ignition source (i.e., panel), a fire at a given panel that does not result in control room abandonment due to habitability and a fire that does result in control room abandonment due to habitability.
The Fire PRA evaluated the conditional probability of damage to a set of targets in the source panel for the non-abandonment scenarios representing loss of control of target sets. These scenarios were reviewed and a postulated fire that damages redundant division cable separated by a fire barrier in panels 1C06, 1C08, or 1C31 result in loss of sufficient control for functions resulting in high consequence.
For these panels, the postulated fire damage was also postulated to result in failure of ASC; therefore, control room abandonment for loss of control was not modeled for these panels. For the other panels, the postulated fire did not result in a loss of sufficient set of controls and the potential benefit of ASC functions were not credited.
This represents potential conservatism in the FPRA.For fires outside the control room, loss of control from target damage was assessed for all fire scenarios.
Fire scenarios with a high consequence due to loss of control (i.e, CCDP greater than 0.05) were identified and discussed in Table 5.4-1 of the FPRA Quantification Report. For the fire scenarios with high consequence outside the control room, the high consequence is the result of fire induced loss of offsite power in which ASC would not provide benefit. In the specific case of the CSR, a fire damaging division 1 cables would not impact the redundant division cables for the impacted function.
Therefore, the loss of sufficient control for functions was not postulated.
Additionally, it is noted that the CSR was analyzed using a 0.1 CCDP representing availability of a single division without offsite power; which would be equivalent to the ASC probability used for the CSR. This modeling, while not explicitly addressing the Rev A. Page 1 of 2 Rev A.Page 1 of 2 RAI -Probabilistic Risk Assessment 19 abandonment of the control room due to loss of control, provides a bounding assessment of the fire risk associated with the CSR.b. Fire scenarios initiated outside the Main Control Room that may impact habitability were evaluated in the multi-compartment analysis.
In addition, fire areas CB2 and CB3 were evaluated with the potential to exhaust smoke to the Main Control Room. See Fire Scenario Report Section 5.2.5.c. DAEC has not deviated from the guidance in NUREG/CR-6850 regarding evaluation of Control Room evacuation.
The FPRA considered fire induced loss of control for all fire areas, Control Room abandonment due to habitability conditions, and the potential for issues with Control Room habitability for fires outside the Control Room.Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -PRA 20 DAEC RAI PRA 20 No transient combustible was postulated in the corner of the Division 2 CSR which contains Division 1 and Division 2 equipment cabling. It is recognized that access to that portion of the room is difficult, however the consequences of a fire in this location may be severe. Provide a discussion of the analysis, assumptions, and risk (ignition frequency and conditional core damage probability (CCDP) and conditional large early release probability (CLERP) of a fire in this location).
RESPONSE: The DAEC FPRA applied a 0.1 CCDP and a 0.03 CLERP for all fires in the CSR. The CSR fire ignition frequency included an unreviewed analysis method (UAM) hot work pre initiator factor (refer to RAI PRA 14 response for discussion of CSR fire ignition frequency and sensitivity analysis).
Division 1 cables for MCC 1B34 are routed in conduits against the south wall of the CSR. Redundant cables for MCC 1 B44 are not routed within the zone of influence of a transient fire postulated to damage 1 B34 cables.In response to this RAI, a transient fire was assumed at the Division 1 conduits damaging MCC 11B34 cables. The quantification resulted in a CCDP of 0.01 and a CLERP of 0.005. These probabilities are consistent with fire scenarios in other plant locations in which a single division is available to mitigate the postulated fire damage (i.e., loss of a single division with offsite power available).
Given the quantified CCDP and CLERP were less than values assumed for the CSR bounding fire scenario, the CSR fire scenario is bounding and the FPRA results input to the License Amendment Request (ML11221A280) (LAR) are not changed.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 23 DAEC RAI PRA 23 According to Table V-1 of the LAR, F&O FSS-C8 on raceway fire wraps is listed as "NA" by the peer review and DAEC. Provide justification that this supporting requirement is not applicable to the fire PRA. Identify if any variance from deterministic requirement (VFDRs) in the LAR involved performance-based evaluations of wrapped or embedded cables. If applicable, describe how wrapped or embedded cables were modeled in the Fire PRA including assumptions and insights on how the PRA modeling of these cables contributes to the VFDR delta-risk evaluations.
RESPONSE: SR FSS-C8 states, "If raceway fire wraps are credited, (a) ESTABLISH a technical basis for their fire-resistance rating, and (b) CONFIRM that the fire wrap will not be subject to either mechanical damage or direct flame impingement from a high-hazard ignition source unless the wrap has been subject to qualification or other proof of performance testing under these conditions." The FPRA assessed SR FSS-C8 with respect to fire wrap and not embedded cables. Raceway fire wrap was not credited in preventing or delaying cable damage in the FPRA; therefore, the SR was listed as not applicable.
From Table B-3 in the LAR, three hour electrical raceway fire barrier systems (ERFBS)are credited in fire area RB1 for RPV isolation and manual operation of SRVs and in fire area TB1 for Div. 2 RHRSW pumps, Div. 2 River Water pumps, and Div. 2 ESW pump and Div. 2 AC Power.In the FPRA, fire damage to wrapped cables (i.e., Darmatt ERFBS in RB1) was postulated consistent to that of exposed cables due to the difficulty in confirming that the wrap will not be subject to mechanical damage that would prevent damage to the protected cables. Given the FPRA did not credit the wrap to prevent damage, the delta risk evaluations reflect a larger delta risk than if the no damage was postulated to the wrapped cables.Cable damage was not postulated to embedded cables (i.e., concrete chase in TB1).Per Table B-3, the concrete chase is a three hour rated. While the concrete chase may potentially be subject to mechanical damage, the damage is unlikely to diminish the fire rating of the chase to the extent that the protected cables would be damaged by postulated fires in the FPRA. The cables protected by the concrete chase are not the subject of a VFDR in TB1. Therefore, the FPRA modeling associated with the concrete chase does not contribute to the VFDR delta-risk evaluations.
Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 27 DAEC RAI PRA 27 In response to F&Os 1-5, 2-12, 4-14, 4-28, 5-18 and 5-36 a discussion of uncertainty and assumptions for the technical elements was added to each of the FPRA development documents.
These are generally large lists of issues of a range of importance.
Provide a consolidated discussion of the key uncertainties and assumptions for this application.
Include those important for the CDF and LERF baseline model, and any sensitivity analyses performed for them.RESPONSE: Appendix B of the FPIE PRA Summary Notebook (DAEC-PSA-QU-14) characterizes sources of model uncertainty for DAEC based on guidance in NUREG-1 855 and on additional industry guidance.
Table B-1 of DAEC-PSA-QU-14 implements the process of identifying generic sources of model uncertainty for DAEC. The results of this assessment lead to the following list of candidate sources of modeling uncertainty to be considered in applications.
Each candidate source is discussed for the NFPA 805 application.
* LOOP frequency and fail to recover probabilities (includes grid stability):
The methodology used in the baseline model uses current industry guidance and is considered to be good practice.
For the FPRA, fire induced LOOP sequences are the significant contributor to CDF and LERF. Given fire induced LOOP, the FPRA does not credited recovery of offsite power. The sensitivity to LOOP frequency and fail to recover probabilities would be negligible considering fire impacts.* FW/CRD injection capability after containment failure: Injection post containment failure is credited in the baseline model with conditional probabilities that containment failure size and location disrupt the injection path. Given the significant contribution of fire induced LOOP sequences with no recovery of offsite power, the sensitivity of these conditional probabilities would be negligible considering fire impacts. In addition, the fire model provides limited credit to FW and CRD as identified in the equipment and cable selection tasks." AC Switchgear Room Cooling dependencies:
AC switchgear room cooling was assessed in the FPRA in support of the transition to NFPA 805 and consistent with the NSCA. The model includes conditional probabilities for equipment failure given the loss of room cooling. The conditional probabilities are discussed in Appendix G.6 of the DAEC Component Data Notebook.
Given cable or equipment failure due to fire, the model may be sensitive to the uncertainties in the conditional probabilities.
o DCBHV-NNRCLNEED-CE--:
conditional probability of failure given loss of room cooling for non-essential switchgear rooms is assumed to be 0.1.The model is not sensitive to the uncertainty in the conditional probability.
Rev A. Page 1 of 13 Rev A.Page 1 of 13 RAI -PRA 27 o DCBHV-NN--RMCLGFCE--:
conditional probability of failure given loss of room cooling for essential switchgear rooms is assumed to be 0.05. The model is sensitive to the uncertainty in the conditional probability.
An increase in the conditional probability to 0.1 by assuming equipment failure would increase FPRA CDF/LERF by approximately 3%.Operability of equipment in beyond design basis environments:
To maintain realism, the baseline model credits the possibility of equipment operability beyond design basis conditions.
Table B-1 identified four failure mechanisms that may be impacted: o LPCI/Core Spray for loss of net positive suction head o LPCI/Core Spray for loss of component cooling o HPCI for loss of lube oil cooling o Control room equipment for loss of room cooling For the FPRA, the significant contribution of fire induced LOOP sequences with no recovery of offsite power would result in negligible sensitivity to the modeling of these systems.* Internal flood initiating event frequencies and failure modes: Not applicable to the FPRA.* Internal flood propagation paths: Not applicable to the FPRA.* ISLOCA frequency:
The methodology used in the baseline model used current industry guidance and is considered good practice.
Increase in pipe rupture or leakage probabilities would have a negligible impact on fire risk. ISLOCA sequences contributed one percent to the total fire risk.Section B.5 of DAEC-PSA-QU-14 provides a list of plant specific sources of model uncertainty.
Table B-2 provides the results of the search in identifying the DAEC specific features to be initially considered as potential candidate modeling uncertainties.
The results of this assessment lead to the following list of candidate sources of modeling uncertainty to be considered in applications.
Each candidate source is discussed for the NFPA 805 application.
* Diesel generator repair probability use in the PRA: Repair probabilities were modeled consistent with industry guidance in the baseline model. For the FPRA, repair was not credited given the potential for fire induced damage to equipment or cables.* Digital FW control failure probabilities:
FW is given limited credit in the FPRA as identified in the equipment and cable selection tasks. Therefore, model uncertainty associated with FW control failure probabilities would be negligible.
Rev A. Page 2 of 13 Rev A.Page 2 of 13 RAI -PRA 27* Credit for motor driven FW pumps: FW is given limited credit in the FPRA as identified in the equipment and cable selection tasks. Therefore, model uncertainty with credit for motor driven FW would be negligible.
Section B.6 of DAEC-PSA-QU-14, Rev. 7 summarizes the potential model sensitivity to uncertainty in human error probabilities (HEP) and common cause failure (CCF)probabilities.
It was identified that these probabilities will be candidate sources of uncertainty for many applications.
For the NFPA 805 application, these are not considered significant sources of uncertainty.
Each HEP was reviewed for fire impacts resulting in increased fire HEPs or no credit given to the operator action. Therefore, in most cases the FPRA included HEPs greater than those in the baseline model. Given this, there remain very few significant HEPs in the FPRA given that no credit was given to offsite power recovery.
CCF probabilities are not altered based on fire impacts. Given equipment or cable fire damage, CCF basic events are not significant in the FPRA.The FPRA reports provide lists of assumptions used in the FPRA. NUREG/CR-6850, Appendix V, was used to identify key uncertainties for the NFPA 805 application.
Table 1 provides the identified NUREG/CR-6850 uncertainty issues and the sensitivity to the uncertainty in the DAEC FPRA.Rev A. Page 3 of 13 Rev A.Page 3 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 1. Plant Boundary The task is not By performing single and multi-compartment analyses, the FPRA Definition and explicitly considered as results are insensitive to variability in plant partitioning.
The plant Partitioning a source of uncertainty was partitioned consistent with the fire protection program which ensures consistency among single and multi-compartment scenarios.
: 2. Fire PRA No treatment of The FPRA equipment list is based on best available information and Component uncertainty is judgment and the results are not sensitive to the equipment list.Selection necessary The selection of components requires not only the consideration of failure modes (active versus passive) but an understanding of the Appendix R functions not previously considered risk significant in the FPIE model. The potential for uncertainty in this task is reduced as a result of multiple overlapping tasks, internal reviews, and the MSO expert panel.Rev A. Page 4 of 13 Rev A.Page 4 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 3. Fire PRA Cable No treatment of The DAEC FPRA cable selection used the FHA-500 methodology Selection uncertainty is and included all cables for all schemes. Therefore, the FPRA necessary potentially fails a component given damage to a cable that would not cause the failure. However, additional circuit analysis was performed as needed during the FPRA development to ensure the fire failures in significant area were reflective of the cable damage.Cable selection was not performed for several PRA components.
These components were included using assumed routing or credit by exclusion which helped to reduce unnecessary conservatism.
A sensitivity quantification was performed in which the Y3 components were assumed to be available (as opposed to damaged) for all fire scenarios.
The sensitivity run concludes that cable selection for Y3 components would at most result in a small reduction of CDF.4. Qualitative No treatment of In the event that a structure which could lead to a plant trip was Screening uncertainty is screened incorrectly, its contribution to CDF would be small (with a necessary CCDP commensurate with base risk) and would likely be offset by inclusion of the additional ignition sources on the reduction of other scenario frequencies.
Rev A.Page 5 of 13 RAI -PRA 27 Table I Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 T Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 5. Plant Fire- Uncertainties are A reactor trip is assumed as the initiating event for all quantification.
Induced Risk related to the model The model logic is then transferred to the appropriate accident Model structure, accident sequence given the fire induced failures.
Several model logic sequences, and changes were made for the FPRA related to MSOs and safe frequencies shutdown components.
Fire induced failures (i.e., 1.0 probability) or cable failure likelihood probabilities represent the significant failures in the model.FPIE and fire PRA peer reviews (including the F&O resolution process), and internal assessments are useful in exercising the model and identifying weaknesses with respect to FPRA logic model.Rev A. Page 6 of 13 Rev A.Page 6 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 6. Fire Ignition Fire ignition frequency Ignition source counting and transient ignition source weighting Frequency distributions should be factors are an area with inherent uncertainty; however, the results propagated through the are not particularly sensitive to changes in ignition source counts or FPRA. Generic fire weighting factors. The primary source of uncertainty for this task is ignition frequencies, associated with the frequency values from NUREG/CR-6850 which plant specific data, result in uncertainty due to variability among plants along with some equipment counting, significant conservatism in defining the frequencies, and their and transient ignition associated heat release rates, based on limited fire events and fire source weighting test data.factors all contribute to A Bayesian update process for events after 2000 was applied to the the uncertainty in fire generic frequencies taken from NUREG/CR-6850.
The uncertainty ignition frequencies.
in the fire ignition frequencies was propagated through the FPRA. A sensitivity quantification was performed using the EPRI 1016735 (see FAQ 08-0048) ignition frequencies and resulted in a significant decrease in CDF and LERF.7. Quantitative No treatment of Quantitative screening was not performed for the FPRA. The fire Screening uncertainty is scenario results are maintained in the cumulative CDF/LERF.necessary Rev A. Page 7 of 13 Rev A.Page 7 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 T Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 8. Scoping Fire No treatment of See task 11 discussion.
Fire scenarios were not screened based on Modeling uncertainty is probability.
necessary 9. Detailed Circuit No treatment of Circuit analysis was performed as part of the NSCA. Refinements Failure uncertainty is in the application of the circuit analysis results to the fire PRA were Analysis necessary performed on a case by case basis where the scenario risk quantification was large enough to warrant further analysis.Therefore, the uncertainty/conservatism which remains in the evaluation is associated with scenarios which do not contribute significantly to the overall fire risk.Rev A. Page 8 of 13 Rev A.Page 8 of 13 RAI -PRA 27 Table I Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 t Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 10. Circuit Failure Circuit failure mode The FPRA used the NUREG/CR-6850 "Option 1" guidance when Mode probability distributions applying cable failure mode probabilities.
The error factor was Likelihood should be propagated applied and the uncertainty propagated through the FPRA. Circuit Analysis through the FPRA. analysis was performed and the circuit failure mode probability was applied for risk significant components in the FPRA when necessary to remove unnecessary conservatism.
A sensitivity quantification was performed in which all spurious events were assigned the appropriate NUREG/CR-6850 probability.
The results showed that the significant events were assigned the appropriate probability and additional assessment would not contribute significantly to the overall fire risk.Rev A. Page 9 of 13 Rev A.Page 9 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 11. Detailed Fire Modeling Fire model parameter uncertainty distributions should be propagated through the FPRA.Uncertainties in fire modeling arise from input parameters (e.g., HRR and damage criteria) to the fire models used.The primary uncertainty in this task is in the area of target failure probabilities.
Conservative heat release rates may result in additional target damage. Non-conservative heat release rates would have an opposite effect.The FPRA used the Generic Fire Modeling Treatments and the Main Control Room Abandonment time calculations.
Each of these reports identifies sources of uncertainty in the fire model and provides sensitivity calculations to show that the fire models provide conservative estimates of critical separation distance, time to hot gas layer, and MCR abandonment times.The fire models used NUREG/CR-6850 guidance in selecting key input parameters like HRR, damage criteria, and MCR abandonment criteria.
Additional calculations were used to include NUREG/CR-6850 parameters for fire growth and time to damage for significant scenarios.
Manual suppression was credited only in the case that a room had automatic detection.
Automatic suppression was credited in preventing hot gas layer and multi-compartment interactions.
Given the fire model reports show that the fire models result in conservative estimates and the use of industry guidance in NUREG/CR-6850, further sensitivity quantifications were not performed and the uncertainty in the parameters was not propagated through the model.Rev A. Page 10 of 13 Rev A.Page 10 of 13 RAI -PRA 27 Table I Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 i Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 12. Post-Fire Human error Human error probabilities represent a potentially large uncertainty Human probabilities for the fire PRA given the importance of human actions in the base Reliability distributions should be model. Since many of the HEP values were adjusted for fire, the Analysis propagated through the joint dependency multipliers developed for the FPIE model also FPRA. Uncertainties in represent a potential for introducing a degree of uncertainty.
the FPRA HRA are the Best estimate HEP adjustments were made to the nominal HEP same as with any HRA values (e.g., cue, time available and performance shaping factors)analysis.
used in the FPIE model then revisited to address unique fire considerations using available industry guidance.
Given the application, fire recovery actions were not modeled to the extent possible.
Recovery actions were only modeled for Main control room abandonment scenarios.
Recovery of fire induced loss of offsite power was not credited in the FPRA.Given the use of available guidance in adjusting performance shaping factors and limited credit for recovery actions, the fire HRA represents an additional source of conservatism in the FPRA. Given the lack of industry guidance and application, a sensitivity quantification was not performed.
The human error probability distributions were propagated through the FPRA.Rev A. Page II ofI3 Rev A.Page 11 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 Task No. Uncertainty Issues Sensitivity of the Results to the Source(s) of Uncertainty
: 13. Seismic Fire Since this is a Seismic fire interaction has no impact on fire risk quantification.
Interactions qualitative evaluation, Uncertainties associated with this task are discussed qualitatively.
there is no quantitative impact with respect to the uncertainty of this task.14. Fire Risk As the culmination of The other source of uncertainty is the selection of the truncation Quantification other tasks, most of the limit. Since the fire PRA solves for CCDP (prior to the application of uncertainty associated frequency) at a truncation limit of 1 E-9, there should not be a with quantification has significant truncation contribution.
This truncation limit is several already been orders of magnitude below the typical CDF value calculated.
A addressed.
sensitivity evaluation of the truncation limit used in the analysis is provided.15. Uncertainty and This task does not N/A Sensitivity introduce any new Analyses uncertainties but is intended to address how uncertainties may impact the fire risk.Rev A.Page 12 of 13 RAI -PRA 27 Table 1 Summary of Key Sources of Uncertainty and Sensitivity for the NFPA 805 Application NUREG/CR-6850 NUREG/CR-6850 Sensitivity of the Results to the Source(s) of Uncertainty Task No. Uncertainty Issues 16. Fire PRA This task does not The documentation task compiles the results of the other tasks.Documentation introduce any new See specific technical tasks for a discussion of their associated uncertainties to the fire uncertainty and sensitivity.
risk.Rev A. Page 13 of 13 Rev A.Page 13 of 13 RAI -PRA 28 DAEC RAI PRA 28 Finding SY-C2-02 (SR SY-C2) of the focused peer review of the internal events PRA notes the lack of evidence that the system level cutsets were reviewed to validate the completeness and accuracy of the system models. The F&O states that based on DAEC discussions the cut sets were reviewed but the review was not documented.
The disposition states that, "given the conservatism of the Fire PRA model and the reviews performed in the process of developing the Fire PRA model, the system cutset review is not expected to identify modeling issues that would impact this LAR submittal." A general reference to conservative models does not resolve the issue. Provide confirmation that the system level cutsets were reviewed and provide a summary of the review of the system models performed in the process of developing the FPRA.RESPONSE: System level cutsets were reviewed prior to issuing the Revision 6 Full Power Internal Events (FPIE) model as the "model of record." Cutset reviews were also performed for specific applications of the FPIE PRA. Cumulatively, these cutset reviews validated the completeness and accuracy of the system models. The results of each cutset review was verified, approved, and documented in accordance with the DAEC fleet PRA procedure, PSAG-2 revision 4. FPIE PRA applications included MSPI, 50.65 (a)(4) -Risk Monitor, Maintenance Rule, and MOV Ranking. System cutsets were also reviewed in the process of developing and updating the fleet required risk ranking report for DAEC management; the ranking includes Initiating Events, SSC and Operator action importance to the DAEC organization.
Insights and inaccuracies that merited changes to the model were processed and dispositioned in accordance with the priorities defined in the DAEC PRA procedure, PSAG-2 revision 4. The issues were documented, processed and tracked in the PRA Change Database and ranked in accordance with the following grading priority from PSAG-2: Rev A. Page 1 of 2 Rev A.Page 1 of 2 RAI -PRA 28 GRADE DEFINITION A Extremely important and necessary to address to assure the technical adequacy of the PSA, the quality of the PSA, or the quality of the PSA update process.B Important and necessary to address, but may be deferred until the next PSA update or an interim update based on the cumulative risk impact of open "B" priority changes.C Considered desirable to maintain maximum flexibility in PSA applications and consistency in the industry, but not likely to significantly affect results or conclusions.
D Editorial or minor technical items, left to the discretion of the model custodian and/or PRAG Manager.No significant issues were identified that merited immediate update or upgrade of the FPIE PRA; all issues were graded as B or lower. In summary, most of the issues are related to potential reductions in conservatism, improvements to the model logic structure, minor logic errors or omissions, and data inaccuracies.
Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -PRA 29 DAEC RAI PRA 29 Finding DA-D4-01A (SR DA-D4) of the focused peer review of the internal events PRA concerns the lack of a discussion of reasonableness of posterior distributions.
While the disposition indicates there will be an enhancement to documentation, no assurance is provided that unreasonable distributions do not adversely impact the results used for the LAR. A review of the FPIE data notebook during the audit identified 5 component types (ARCR FR, ARCR FS, AS1K FR, DFCV FC, and DFCV FO including the 1 identified in the finding) where the posterior distributions differ significantly from the prior distributions.
Capability Category II for SR DA-D4 requires a check that the posterior distribution is reasonable given the weight of evidence provided by the prior and the plant specific data. Provide an assessment about the potential impact of differences in these distributions on the change in risk results used to support the submittal.
If the change in risk could be significant, provide an assessment of the reasonableness of the plant specific data distributions and the impact on FPRA results for any structure, system, and component (SSC) where the posterior distributions are found unreasonable.
RESPONSE: The five identified component types (ARCR FR, ARCR FS, AS1 K FR, DFCV FC, and DFCV FO) include the one identified in the finding (AS1 K FR). The current posterior mean value was substituted with the highest mean value from either the prior or plant data for four of the five component types (ARCR FR, ARCR FS, AS1 K FR, and DFCV FC.) Total increase in core damage frequency (CDF) and total increase in large early release frequency (LERF) were found to be much less than Regulatory Guide 1.174 thresholds for risk significance (1 E-6 /yr and 1 E-7/yr respectively) as a result of the changes. The substituted values are listed in the table below.Type Code Current Posterior Replacement Value Origin of replacement value Value ARCR FR 9.26 E-5 8.6 E-4 Prior (Generic data)ARCR FS 1.73 E-2 2.19 E-2 Plant Data ASIK FR 2.75 E-4 5.5 E-3 Plant Data DFCV FC 2.35 E-3 2.83 E-3 Plant Data The current posterior mean value was substituted with the plant data value for DFCV FO and increases in CDF and LERF were found to be close to Regulatory Guide 1.174 thresholds for risk significance.
The substituted value is listed below.Type Code Current Posterior Replacement Value Origin of replacement value Value DFCV FO 3.40 E-3 4.24 E-3 Plant Data The posterior distribution of DFCV FO was judged to be reasonable by use of examples presented in the PRA standard for supporting requirement DA-D4. For instance, the posterior distribution shape for DFCV FO is not a single bin histogram and does not Rev A.Page 1 of 2 RAI -PRA 29 have a multimodal shape. The posterior distribution mean value appears reasonable since it is within one order of magnitude of the plant specific data mean value.A review of the PRA model results was completed to identify the significant components that would be impacted from changes to the service water air-operated valve failure to open frequency (DFCV FO) type code. The significant components were the emergency diesel generator cooler inlet valves and the river water supply inlet valves to the Pump House basin. Since these components are not subject to any variances from deterministic requirements, changing the failure to open frequency for these components would not change any conclusions in the NFPA 805 application.
Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -PRA 35 DAEC RAI PRA 35 Section 6 of the Fire Quantification Report describes the development and application of a 0.02 failure probability for incipient detection in modeling several CR panels based on guidance in FAQ 08-0046. The methodology in FAQ 08-0046 was developed for implementing incipient detection in electrical cabinets in rooms that are not constantly manned, contrary to the CR. Provide justification for the risk reduction credit obtained from implementing incipient detection.
Discuss any deviations in the incipient detection risk analysis from NUREG/CR-6850 or the FAQ.RESPONSE: The DAEC FPRA credited the committed incipient detection modification in control room panels 1C03, 1004, 1005, 1C06, 1C08, 1C15, 1C17, 1C26, 1031, 1032, 1033, and 10C44. For fire scenarios at these panels, the FPRA applied a credit using the FAQ 08-0046 simplified event tree approach which resulted in a non suppression probability of 0.02 (see Section 6 of the FPRA Quantification Report, 049080001.004).
Per this RAI, given the control room is constantly manned, the applied incipient detection non suppression probability was revised.The following event tree was developed for the FPRA based on the Figure 13-2 of Supplement 1 to NUREG/CR-6850.
The point estimates of each branch of the simplified event tree are conservative estimates from the FAQ. The FAQ event tree nodes are: 1 Fire initiating event: Set to 1.0 in the event tree. The fire initiating event frequency is applied during quantification.
2 Fraction that Have an Incipient Phase Detectable by System: This event tree node is not included in the simplified event tree. As such, the event tree assumes that the panel components have an incipient stage that is detectable.
3 Detector System Availability and Reliability:
Assigned a 0.99 probability based on FAQ 08-0046.4 Successful Operator Response to Alert: Assigned a 0.99 probability based on FAQ 08-0046.5 Technician Successful in Preventing Fire in Incipient Stage: This event tree node is not included in the simplified event tree. As such, the event does not give credit in preventing fire in the incipient stage.6 Fire Suppressed:
Assigned a 0.999 probability based on FAQ 08-0046.From above, the methodology from FAQ08-0046 resulted in a 0.02 non suppression probability given incipient detection when using the simplified event tree. Per this RAI, the methodology was not developed for rooms constantly manned.Rev A.Page 1 of 3 RAI -PRA 35 In determining the non suppression probability to be applied in the control room for panels with incipient detection, the following assumptions are required:* Per this RAI, incipient detection provides more benefit to plant locations not constantly manned. Therefore, the incipient detection non suppression probability for rooms constantly manned is assumed not to be less than 0.02.* The non suppression probability for incipient detection should be less than the prompt suppression probability of 0.189 (based on Supplement 1 to NUREG/CR-6850) allowed by NUREG/CR-6850.
* The total non suppression probability for the control room should be less than other plant locations when crediting the same detection systems with manual suppression.
Therefore, the total control room non suppression probability should be 0.02 or less." As a result, the applied non suppression probability for incipient detection in the control room is revised to be 0.1 (0.02/0.189).
With the above assumptions, the control room non suppression probability (nsp) applied for panels with incipient detection is: Incipient detection nsp (0.1)
* prompt manual suppression nsp (0.189) = 0.02 The change in credit provided for incipient detection does not change the conclusions in the License Amendment Request (ML1 1221A280) (LAR), and the FPRA results for CDF/LERF and A CDF/LERF still meet the acceptance criteria.Rev A. Page 2 of 3 Rev A.Page 2 of 3 RAI -PRA 35 Detector Fire System Initiating Availability Event and Reliability Successful Operator Response to Alert Fire Suppressed End Point Probability 0.98 0.999 No Fire 0.99 0.001 0.99 0 No Fire Damage To 0.00098 Targets Outside of Cabinet No Fire Damage To 0 Targets Outside of Cabinet 0.0099 Fire Damage No Fire Damage To 0 Targets Outside of Cabinet 0.01 Fire Damage 1 0 0.01 Rev A. Page 3 of 3 Rev A.Page 3 of 3 RAI PRA 38 DAEC RAI PRA 38 F&O 2-6 noted that the Revision 6 FPIE model used for the FPRA model had not been approved and that there were draft system notebooks that had a section for fire impact that needed updating.
While two FPIE system notebooks reviewed were approved, a third one (HPCI notebook, DAEC-PSA-SY-05.05, Rev. 3) had not been approved.Confirm that the FPIE model on which the FPRA is based and the supporting documentation was fully approved.RESPONSE: The Revision 6 FPIE PRA model used as the base model for the FPRA was approved by DAEC through the Engineering Change Process (DCR 0000156431).
The system notebooks were accepted as final. The HPCI notebook, DAEC-PSA-SY-05.05, Revision 4 includes fire impact information in Section 5. The FPIE model on which the FPRA was based, and the supporting documentation was fully approved.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 43 DAEC RAI PRA 43 The LAR Table W-1, "DAEC Fire PRA CDF Significant Fire Initiating Events (Individually Representing 
> 1% of the Calculated CDF)," describes fire scenarios involving failure of HPCI and reactor core isolation cooling (RCIC). Respond to the following:
: a. Why are the HPCI and RCIC systems both failed in these fire scenarios?
Is the ability to depressurize the reactor modeled for these fire scenarios, including dependencies on DC power and operator actions?b. Are both systems failed by the same fire(s)? If so, is it (are they) associated with VFDR(s)?c. Were operator actions, recovery actions (RAs), or other actions to address risk, defense-in-depth, or safety margins, considered or modeled for these fire scenarios?
For the corresponding fire areas, Table B-3 states that Fire Risk Evaluations determined no further action was determined to be necessary.
Discuss the rationale for this conclusion.
RESPONSE: In several fire scenarios identified in LAR Table W-1, the HPCI and RCIC systems were failed. These fire scenarios were in the essential switchgear rooms. In these fire scenarios, fire damage was postulated to result in a loss of offsite power. In addition, the applicable switchgear was postulated to be failed from fire damage. These fire scenarios result in a loss of all AC power given postulated random failures of the opposite division diesel generator or support systems. Due to the specific fire damage resulting in loss of offsite power, recovery of loss of offsite power was not credited.Without recovery of loss of offsite power, these fire scenarios resulted in the loss of HPCI and RCIC.Offsite power or the HPCI and RCIC systems are not credited in the NSCA for fire in the essential switchgear rooms. Therefore, failures associated with the HPCI and RCIC system in these fire scenarios are not associated with VFDRs.The fire risk evaluations determined that the plant total risk and delta risk met the acceptance criteria for risk, defense-in-depth, and safety margin. The fire scenarios were reviewed to determine actions that may reduce plant overall risk. However, no specific actions were identified and credited in the FPRA.Quantification of fire scenarios discussed in this question are affected by the logic error discussed in RAI PRA 1. The FPRA model was corrected and re-quantified in response to RAI PRA 1.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 50 DAEC RAI PRA 50 As shown in Table W-3, explain why the A LERF is greater than the A CDF for Fire Area CB3.RESPONSE: In response to RAI PRA 01, the Fire PRA was updated and requantified.
The new model resulted in an estimated CDF of 2.28E-5/yr and an estimated LERF of 6.45E-6/yr for Fire Area CB3. The update also resulted in a delta CDF of 1.43E-8/yr and a delta LERF of 1.28E-8/yr for the same area.The LERF model contains a relatively large number of basic events that serve as flags by being set to values greater than 0.1. This practice is inconsistent with the 'minimum cutset upper bound' simplification method employed by the quantification engine used at DAEC, and is the reason that LERF is over-estimated relative to associated CDF values.In the updated LERF model, over-estimation of results has been softened by setting several basic events to "TRUE" instead of to 1.0. As such, the updated delta CDF is greater than the updated delta LERF. Attachment W of the enclosure to the License Amendment Request (ML11221A280) (LAR) discusses the over-estimation of calculated risk in the FPRA.Rev A. Page 1 of I Rev A.Page 1 of 1 RAI -PRA 51 DAEC RAI PRA 51 Table S-I, "Plant Modifications Committed" of the LAR identifies a modification to pump circuitry to correct a VFDR to result in a deterministically compliant condition and to assist in a reduction in cumulative A LERF to support meeting Regulatory Guide (R.G.)1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision 1, November 2002", criteria.
Address the following for this modification:
: a. The entry in Table S1 states that this modification will "assist in a reduction in cumulative A LERF to support meeting R.G. 1.174 criteria." Explain what this means.b. Correcting the FPRA model error discussed during the NRC audit could change the results. Re-evaluate the reduction in LERF as a result of the modification for the applicable fire area, the reduction of cumulative A LERF, and the cumulative A CDF.c. For the FPRA developed for this modification, confirm that the evaluations were done using peer reviewed methods for the FPRA and the supporting FPIE PRA.d. Verify that fire(s) relevant to the modification, for the associated fire area noted in Table S-1, do not result in a scenario that should deterministically assume the loss of offsite power. Explain whether or not the conditional probability of loss of offsite power given a plant trip is included in the probabilistic evaluations.
If not, why not?RESPONSE: a. Section 5.4.1 of Report Number 0027-0042-000-004 Duane Arnold Energy Center Fire Risk Evaluations discusses the risk acceptance criteria associated with the transition to NFPA 805 application.
Given the quantified LERF, a modification to the ESW B pump control circuit was identified in fire area TB1 to reduce overall LERF and delta LERF. The risk reduction from the modification was identified as 3.90E-8/yr in Table W-4 of Attachment W of the enclosure to the License Amendment Request (ML1 1221A280) (LAR).b. In response to RAI PRA 01 the FPRA was updated to correct the identified model error. Additionally, in response to RAI PRA 11 the fire large oil fire scenario in the turbine building was updated. The conditional probability of a catastrophic turbine oil fire was refined based on further justification in the RAI response.
Given the changes, the TB1 fire area CDF was calculated to be 1.94E-6/yr and LERF was calculated to be 4.99E-7/yr.
The delta CDF and LERF was calculated to be 3.80E-10/yr and 1.10E-10/yr, respectively.
These results consider the modification.
Without the modification, the updated model results in an estimated CDF of 2.74E-6/yr and LERF of 6.62E-7/yr and delta CDF of 8.08E-7/yr and delta LERF of 1.64E-7/yr.
Given the results of the Rev A. Page 1 of 2 Rev A.Page 1 of 2 RAI -PRA 51 updated model quantification, the ESW B pump control circuit modification will still assist in a reduction of overall LERF to support meeting the acceptance criteria.c. The FPRA scenarios for the fire area TB1 used peer reviewed methods as updated by the response in RAI PRA 11.d. Report number FPLDA013-PR-021, At-Power Analysis for Fire Area TB1 for Duane Arnold Energy Center, deterministically assesses the availability of offsite power for the fire area. Based on the conclusions of the report, offsite power is deterministically available for the fire area. The PRA model logic does include the conditional probability of a loss of offsite power given a plant trip.Rev A. Page 2 of 2 Rev A.Page 2 of 2 RAI -PRA 61 DAEC RAI PRA 61 As indicated by F&O 4-42, as the PRA model evolved HFEs may have been added and the HRA dependency analysis needs to consider all of the multiple HFEs that appear in the final model. Confirm that the dependency analysis appropriately considered all of the final models' multiple HFEs.RESPONSE: The FPRA model was changed in response to RAI PRA #01. Additionally, other RAI responses impacted the HRA and dependency analysis.
Therefore, the dependency analysis was performed again after changes were made to the FPRA model. The dependency analysis was performed consistent with the previous dependency analysis and identified several new HFE combinations which were added to the FPRA recovery file. The new combinations are included in the updated recovery file (see FPRA model updated report 0493080001.006).
Rev A. Page 1 of I Rev A.Page I of I Attachment 2 to Revised Pages to Response For Additional Information Fire Modeling RAI FM 3 and RAI FM 4 5 pages follow 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]: where cp is the heat capacity of ambient air (kJ/kg-K [Btu/Ib-&deg;R]), Too is the ambient temperature (K [&deg;R ]), g is the acceleration of gravity (m/s 2 [ft/s 2]), p.o 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 AHc 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 L ratio for which the D5 plume equations have validation basis is between 7 -700 kW 2/5/m (4 -9 Btu 2 1 5/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/m 2 (8.81 Btu/s-ft 2) to well over 3,000 kW/m 2 (264 Btu/s-ft 2). For weaker fires (heat release rates less than 100 kW/m 2 [8.81 Btu/s-ft 2], 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 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 release rate per unit length range (z) 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 -3290C (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 predictions when the target separation to fire diameter ( -R) ratio is between 2.2 and 5.7 Df[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-ft 2) 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 Shokri and Fire Size Rate Per Unit Fire Diameter Model Heat Beyler Heat Fire [Size Areat Per Ui Flux Range Flux Range (kW [Btu/s]) Area (kW/m 2  (m [ft]) (kW/m 2 [Btu/s- (kW/m 2 [Btu/s-[Btu/s-ft 2]) ft 2]) ft 2])0.07 -0.45 0.36 -3.8 25 (24) 100 (8.8) 0.56 (1.9) (0.006 -0.04) (0.03 -0.4)2-13.6 2.84-10 25 (24) 3,000 (264) 0.1 (0.3) (0.2 -1.2) (0.3 -0.9)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 0.45-4.6 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-ft 2). 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-ft 2). Clearly, the steady state ZOI dimensions based on critical heat fluxes of 5.7 -11.4 kW/2 (0.5 -1 Btu/s-ft 2) 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-ft 2]). 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 Rev D.Page 5 of 8 RAI -Fire Modeling 4 AT = 5.38 (1)where AT is the maximum temperature within the ceiling jet (0C) at a distance r from the centerline of the fire (m), &#xfd; 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 (3090C [556&deg;F]) as computed using Equation I 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 Peak Heat Horizontal ZOI Base for Ignition Release Rate Flame Heightt Dimension Bae the (kW [Btu/s]) (m [ft]) used in 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 _tFrom 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. One 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 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-ft 2), the internal fire is adjacent to one boundary, and all energy is direction 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: Qb,max = AbE (2)where Qb,max 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-ft 2]). Assuming the maximum flame emissive power is 120 kW/m 2 (10.6 Btu/s-ft 2) 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 fully 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}}