ML20246G636

Request for Additonal Information to Support Review of License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times

ML20246G636
Person / Time
Site:	Nine Mile Point
Issue date:	09/02/2020
From:	Marshall M Plant Licensing Branch 1
To:	Reynolds R Exelon Nuclear
Marhsall M, NRR/DORL/LPL, 415-2871
References
EPID L-2019-LLA-0234
Download: ML20246G636 (18)

Text

Marshall, Michael From: Marshall, Michael Sent: Wednesday, September 02, 2020 10:04 AM To: [Licensee] Ron Reynolds (Exelon)

Cc: Danna, James

Subject:

NINE MILE POINT NUCLEAR STATION, UNIT 2 - REQUEST FOR ADDITONAL INFORMATION TO SUPPORT REVIEW OF LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT RISK INFORMED COMPLETION TIMES (EPID L-2019-LLA-0234)

Hello Ron, By letter dated October 31, 2019 (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML19304B653) as supplemented by letters dated December 12, 2019 and August 28, 2020 (ADAMS Accession Nos. ML19346F427 and ML20241A044, respectively), Exelon Generation Company, LLC (Exelon, the licensee) requested that the U.S. Nuclear Regulatory Commission (NRC) amend the Technical Specifications, Appendix A of Renewed Facility Operating License No. NPF-69 for Nine Mile Point Nuclear Station, Unit 2. Exelons proposed license amendment request (LAR) would revise technical specification requirements to permit the use of risk informed completion times (RICTs) for actions to be taken when limiting conditions for operation are not met. The proposed changes are based on Technical Specifications Task Force Traveler 505, Revision 2, Provide Risk Informed Extended Completion Times - RITSTF Initiative 4b, dated July 2, 2018 (ADAMS Accession No. ML18269A041).

The U.S. Nuclear Regulatory Commission staff has reviewed the information provided in the LAR and has determined that additional information is needed to complete its review. The request for additional information was discussed with you on September 1, 2020, and it was agreed that your response would be provided within 30 days of the date of this email.

The RAIs listed below are not a complete listing of the additional information needed to complete the NRC staffs review. RAIs 1 to 5 were sent to you in a separate email dated July 30, 2020 (ADAMS Accession No. ML20213A935).

RAIs Internal PRA Section 50.36(c)(2) of Title 10 of the Code of Federal Regulations (10 CFR) requires technical specifications (TSs) to contain limiting conditions of operations (LCOs) that describe the lowest functional capability of equipment required for safe operation of a plant and requires to follow any remedial actions permitted by the TSs. The remedial actions need to be completed within a set time frame commonly referred to as a completion time (CT) or allowed outage time. The risk informed completion time program that Exelon has requested to adopt at Nine Mile Point 2 is one way of establishing or changing a CT using a risk-informed approach that relies on probabilistic risk assessments (PRAs).

RAI 6

Regulatory Guide (RG) 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, (ADAMS Accession No. ML090410014) provides guidance for addressing probabilistic risk assessment (PRA) acceptability. RG 1.200, Revision 2, describes a peer review process using the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard ASME/ANS-RA-Sa-2009, Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, as one acceptable approach for determining the technical acceptability of the PRA. The primary 1

results of peer review are the findings and observation (F&Os) recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the Nuclear Energy Institute (NEI) guidance documents, NEI 05-04, NEI 07-12, and NEI 12-13, titled NEI 05-04/07-12/12-06, Appendix X: Close Out of Facts and Observations (F&Os), (ADAMS Package Accession No. ML17086A431), which was accepted by the NRC staff in a letter dated May 3, 2017 (ADAMS Accession No. ML17079A427).

The license condition proposed in Attachment 7 of the LAR includes the commitment to complete a number of implementation items prior to implementation of RICT program. One of these implementation items is to address the open F&Os from the internal events PRA F&O closure report. This implementation item does not describe what updates will be made to the internal events PRA models to resolve the three remaining F&Os associated with the support system initiating event (SSIE) fault trees or cite resolutions described elsewhere in the LAR, such as in the descriptions in the Disposition column for F&Os presented in Enclosure 2, Table E2-1, of the TSTF-505 LAR. Therefore, address the following:

a. The disposition for F&O 5-1 states that the cited correction factor will be replaced with improved modeling in the PRA. If available, describe the proposed PRA modeling and any subsequent modifications to the corresponding implementation item.

b. The disposition for F&O 8-1 states that a systemic review of the cutsets produced by the SSIE fault trees will be performed to identify feasible recovery actions that could impact the frequency of the associated SSIE. The disposition does not commit to updating the PRA if feasible recovery actions that could impact the frequency of the associated SSIE are identified. Provide a description of the actions that will be performed upon identifying feasible recovery actions.

c. The disposition for F&O 8-2 appears to indicate that the mission time for common cause factors used in the SSIE fault tress will be adjusted to a year-long mission time. Describe how mission time will be adjusted and, if applicable, provide an update to the associated implementation item.

RAI 7

The ASME/ANS RA-Sa-2009 PRA standard defines PRA upgrade as, the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of the ASME/ANS RA-Sa-2009 PRA standard states that, [u]pgrades of a PRA shall receive a peer review in accordance with the guidance specified in the Peer Review Section of each respective Part of this Standard []. Criteria presented to identify PRA upgrades are (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

LAR Enclosure 2 states that the last full scope peer review for the internal events PRA was conducted in July 2009 and that an F&O closure review to close out F&Os from the 2009 review was conducted in February 2019. The LAR does not indicate what internal events and internal flood PRA model changes were made between July 2009 and February 2019 to improve the model or to incorporate changes to reflect the as-built, as-operated plant. Address the following:

a. Summarize the model changes performed for the internal events, including internal flood PRA since July 2009, and for each change, justify why it does or does not meet the definition of a PRA upgrade as defined in the ASME/ANS RA-Sa-2009 PRA standard.

b. Confirm that focused-scope peer reviews have been conducted for any model change performed for the internal events, including internal flood, PRA model since July 2009 that meets the definition of a PRA upgrade as defined in the ASME/ANS RA-Sa-2009 PRA standard. Describe the peer review and status of the resulting F&Os. Provide any remaining open F&Os, along with dispositions for this application.

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c. During the regulatory audit conducted May 4 to 7, 2020, the licensee identified a number of human reliability analysis (HRA) changes to the internal events PRA model, which included a change from human cognitive reliability to accident sequence evaluation program time reliability correlation, change to using the HRA Calculator joint human error probability (HEP) tool versus manual grouping of HEPs and changes to credit for diverse and flexible mitigation capabilities (FLEX).

Non-mandatory Appendix I-A of the PRA Standard states:

Consideration should be given to the scope or number of PRA maintenances performed.

Although individual changes to a PRA model may be considered PRA maintenances, the integrated nature of several changes may necessitate a peer review. Multiple PRA maintenances can, over time, lead to considerable change in the insights (e.g., importance rankings, relative risk significance of [structures, systems, and components] SSCs).

Discuss whether the cumulative changes in the internal events since 2009 led to considerable change in the risk insights and whether a peer review was performed.

RAI 8

Section 2.3.4 of NEI 06-09, Revision 0-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Industry Guidance Document, (ADAMS Accession No. ML12286A322) specifies that [c]riteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations.

LAR Enclosure 7 states that if, a plant change or a discovered condition is identified that has a significant impact on the RICT Program calculations [], an unscheduled update of the PRA models will be implemented. The LAR does not explain under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedure that will be used to initiate the update.

Therefore, describe the conditions under which an unscheduled PRA update (i.e., less than once every two refueling cycles) would be performed, the criteria that would be used to require a PRA update and how the impact on RICT estimates is considered. In the response, define what is meant by significant impact to the RICT Program calculations.

RAI 9

The NRC staffs safety evaluation (SE) to NEI 06-09, Revision 0-A specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions and that justification be provided to show that the scope of the PRA model is consistent with the licensing basis assumptions. Table E1-1 in Enclosure 1 of the LAR identifies each TS LCO proposed to be included in the RICT program and describes how the systems and components covered in the TS LCO are implicitly or explicitly modeled in the PRA. For certain LCOs, the LAR did not provide sufficient description of the PRA modeling that will be used in the RICT calculations. Therefore, address the following:

a. For TS LCO 3.6.1.7 (Suppression Chamber-to-Drywell Vacuum Breakers), Condition A (One line with one or more suppression chamber-to-drywell vacuum breakers inoperable for opening), LAR Table E1-1 states that the PRA model includes one failure mode: lines fail to close after initially opening. The model will be updated to include this failure mode prior to exercising the RICT program for this TS. The meaning of text is not clear. Based on the text, it appears that the PRA model already includes the failure mode that could be used to calculate the RICT. The implementation item table presented in LAR Attachment 6 has the same wording as used in the comment column of this TS LCO Condition.

i. Explain how failure to open of the Suppression Chamber-to-Drywell Vacuum Breakers impacts the core damage frequency (CDF) or large early release frequency (LERF) and how a change in CDF and LERF can be calculated for the RICT estimate.

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ii. Explain and justify the PRA model changes proposed for the implementation item associated with TS LCO Condition 3.6.1.7.A.

b. For TS LCO 3.7.1 (Service Water System and Ultimate Heat Sink), Condition C (One service water subsystem inoperable for reasons other than Conditions A and B), LAR Table E1-1, states that the:

[] success criteria are consistent with the design basis except when UHS temperature is > 82

[degrees Fahrenheit] °F. The model is being updated to include this condition prior to exercising the RICT program for this TS.

The implementation item table presented in LAR Attachment 6 has the same wording as used in the comment column of this TS LCO condition. This seems to imply that the success criteria that will be used in the PRA models to complete the implementation item associated with TS LCO Condition 3.7.1.C will be the same as the design-basis success criterion, four of six pumps during a loss of coolant accident (LOCA) without a loss of off-site power and ultimate heat sink greater than 82 °F and less than or equal to 84 °F. Describe and justify the PRA model update.

c. For TS LCO 3.3.5.1 (Emergency Core Cooling System (ECCS) Instrumentation), Condition E (ECCS Actuation instrumentation for low pressure core spray (LPCS), low pressure coolant injection (LPCI),

high pressure core spray (HPCS)), LAR Table E1-1, states that the failure of the HPCS minimum flow valve will be used as a surrogate for HPCS discharge instrumentation failure.

i. Explain how failure of the HPCS minimum flow valve is deemed conservative compared to failure of HPCS discharge instrumentation.

ii. Explain how actuation instrumentation for LPCS and LPCI is modeled in the PRA and how a RICT estimate can be calculated.

d. LAR Table E1-1 indicates for TS LCO 3.5.1 (Low Pressure ECCS Injection/Spray), Condition A (One low pressure ECCS injection/spray subsystem inoperable), that the PRA success criterion is One of four subsystems, while the design-basis success criterion is Two of four subsystems. The explanation for this difference was not provided in LAR Table E1-1 and is not clear to the NRC staff. The comment column indicates for this LCO condition that the success criteria are consistent with the design basis for each train. Therefore, address the following:

Clarify and justify the PRA success criteria used to model systems associated with TS LCO Condition 3.5.1.A, Low Pressure ECCS Injection/Spray, and provide justification for the less demanding success criteria.

e. For TS 3.3.7.2 (Mechanical Vacuum Pump Isolation Instrumentation), Condition A (one or more channels inoperable), the first implementation item listed in Attachment 6 of the LAR states that:

SSCs are not modeled. The model will be updated to include these SSCs prior to exercising the RICT program for this TS. The PRA Success Criteria will match the Design Success Criteria.

i. Describe the proposed PRA modeling associated with TS 3.3.7.2.A.

ii. Explain how the inoperability of the mechanical vacuum pump isolation instrumentation impacts CDF or LERF and how a change in CDF and LERF can be calculated for the RICT estimate.

iii. If applicable, provide an update to the associated implementation item.

f. For TS LCO 3.7.1.D (One division of intake deicer heaters inoperable), the fifth implementation item listed in Attachment 6 of the LAR states that the intake deicer heaters are not directly modeled in the PRA. The model will be updated to explicitly include these components prior to its use with RICT [].

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i. Describe the proposed PRA modeling associated with TS 3.7.1.D.

ii. Explain how the inoperability of the deicer heaters impacts the CDF or LERF and how a change in CDF and LERF can be calculated for the RICT estimate.

iii. Describe how success criteria associated with intake deicer heaters will be tracked in the Real-Time Risk (RTR) model.

iv. If applicable, provide an update to the associated implementation item.

g. For TS LCO 3.7.5 (Main Turbine Bypass System), Condition A (Main Turbine Bypass System -

Requirements of the LCO not met), LAR Table E1-1, indicates that the PRA success criterion is Three of five bypass valves, while the design-basis success criterion is Five of five bypass valves. The explanation provided in the comment column of table for this entry states that the PRA success criteria is based on the minimum valves required to prevent major demands on the suppression pool. The function of the main turbine bypass valves as stated in LAR Table E1-1 is to control steam pressure when reactor steam generation exceeds turbine requirements during unit startup, sudden load reduction, and cooldown. Accordingly, it is not clear how preventing major demands on the suppression pool is equivalent to limiting peak pressure in the main streamlines and reactor to acceptable limits. Therefore, address the following:

i. Explain the PRA modeling for the main turbine bypass system and its impact on CDF and LERF.

ii. Justify that successful opening of three of five main turbine bypass valves is sufficient to fulfill the safety function of these valves under TS LCO Condition 3.7.5.A in the accident scenarios modeled in the PRAs.

RAI 10

Regulatory Position 2.3.3 of RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, (ADAMS Accession No. ML17317A256) states that:

[t]he level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements.

Additionally, NEI 06-09, Revision 0-A states:

If the PRA model is constructed using data points or basic events that change as a result of time of year or time of cycle (examples include moderator temperature coefficient, summer versus winter alignments for HVAC, seasonal alignments for service water), then the RICT calculation shall either 1) use the more conservative assumption at all time, or 2) be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration for the feature as modeled in the PRA.

LAR Enclosure 8, Section 2, states that, [t]he impact of outside temperatures on system requirements like seasonal service water pumps were evaluated and found no dependent flags were needed to be addressed in the CRMP model. LAR Enclosure 9, Table E9-1, indicates that the industry data used in the PRA models includes data for weather-related loss-of-offsite power. The NRC staff notes that seasonal variations in weather conditions include environmental factors besides temperature.

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a. Explain how any changes in initiator frequency due to seasonal variations is accounted for in the RTR model used in the RICT calculations. If changes in initiator frequency due to seasonal variation are not addressed in the RTR model, then provide justification for this simplification.

b. Explain how any changes in plant response success criteria based on seasonal variations are accounted for in the RTR model used in the RICT calculations. If changes in plant response success criteria due to seasonal variation are not addressed in the RTR model, then provide justification for this simplification.

c. For any items identified in parts (a) and (b) above, describe any additional Risk Management Actions (RMAs) that will be performed.

RAI 11

The NRC staff SE to NEI 06-09, Revision 0-A, specifies that the LAR should identify key assumptions and sources of uncertainty and assess and disposition each as to their impact on the RMTS application.

Section 5.3 of NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Final Report, (ADAMS Accession No. ML17062A466), presents guidance on the process of identifying, characterizing, and qualitative screening of model uncertainties.

LAR Enclosure 9 states that the process for identifying key assumptions and sources of uncertainties for the internal events PRA and fire PRA (FPRA) was performed using the guidance in NUREG-1855, Revision 1. It states that the internal events and FPRA models and notebooks were reviewed for plant-specific key assumptions and sources of uncertainty. Further, the LAR states that generic sources of uncertainty for the internal events PRA were identified from Electric Power Research Institute (EPRI) Technical Report (TR)-1016737, Treatment of Parameter and Modeling uncertainty for Probabilistic Risk Assessments and for the FPRA from EPRI TR-1026511, Practical Guidance of the Use of Probabilistic Risk Assessment in Risk-informed Applications with a Focus on the Treatment of Uncertainty. The LAR does not describe the process and the criteria used to identify, from the initial comprehensive list of assumptions and sources of uncertainty in the base PRA model(s) (including those associated with plant-specific features, modeling choices, and generic industry concerns), the specific key assumptions and sources of uncertainties presented in the LAR.

Describe, separately for the internal events, internal flooding and the fire PRAs, the process used to identify and evaluate key assumptions and sources of model uncertainty. Address the following in the response:

a. Discuss how a comprehensive list of plant-specific and generic industry key assumptions and sources of uncertainty were identified as a starting point for this evaluation.

b. Explain how the comprehensive list of key assumptions and sources of uncertainty sources was screened to a list of uncertainties that were specifically evaluated for their impact on the RICT application.

c. Explain what criteria, qualitative or quantitative, or what additional analysis were used to evaluate the impact of the key assumptions and sources of uncertainty on the RICT application.

RAI 12

The NRC staff SE to NEI 06-09, Revision 0-A, specifies that the LAR should identify key assumptions and sources of uncertainty and assess and disposition each for impact on the application. LAR Enclosure 9 Tables E9-1 and E9-3 identify the key assumptions and sources of uncertainty for the internal events and fire PRA and provide their associated disposition for the application. The NRC staff reviewed the key assumptions and sources of modeling uncertainty and their dispositions provided in the LAR and noted few items that appeared to have an impact on the RICT calculations. During the regulatory audit conducted May 4 to 7, 2020, the NRC staff also reviewed the Nine Mile Point 2 internal events and fire PRA uncertainty analysis report and noted few 6

additional items, not identified in the LAR, that appeared to have an impact on the RICT application. Therefore, address the following:

a. LAR Enclosure 9, Table E9-1, states that treatment of suppression pool strainers performance is a modeling uncertainty. The disposition to this modeling uncertainty states:

Because suction strainer failures impact all ECCS systems as a common-mode failure, any potential extended unavailability via RICT is not relevant. This item does not represent a key source of uncertainty for the RICT Application.

It is not clear why the assumed individual and common cause failure probabilities for the suppression pool strainers have no impact on the RICT calculations. The NRC staff notes that suppression pool strainer plugging contributes to the failure probability of ECCS systems and that LCOs exist in the RICT program for the ECCS. Accordingly, it appears that if the strainer plugging probability is underestimated, then the RICT for an ECCS system can be overestimated. Therefore, justify the conclusion that the uncertainty associated with suppression pool strainer performance has no impact on the RICT calculations.

b. LAR Enclosure 9, Table E9-1, states that:

Since BWRs are designed to maintain 2/3 core height for a very large break LOCA, injection by one LPCI pump into the shroud area may maintain the covered core sub-cooled. Cooling of the top 1/3 core for a substantial time is questionable, since long-term steam cooling effect may not be ensured. Nine Mile Point 2 assumes that a single LPCI pump is adequate, and there is no real evidence yet that this is not acceptable to prevent core melt.

The LAR also states that a set of sensitivity studies has been performed that shows this uncertainty has a minimal impact on the RICT calculation. However, the LAR does not describe those studies or provide the results.

Describe the sensitivity studies that were performed. Include a description of the assumptions that were made in the sensitivity cases and provide the results of the studies that support the conclusion that this uncertainty only has a minimal impact on the RICT calculations.

c. LAR Enclosure 9, Table E9-3, identifies detailed circuit analysis as a source of fire PRA modeling uncertainty because of conservatisms in the approach. The NRC staff notes that because detailed circuit analysis is resource-intensive, it is not typically performed on all circuits. The disposition to this source of uncertainty presented in Table E9-3 states that [] uncertainty (conservatism) that may remain in the fire (FPRA) is associated with scenarios that do not contribute significantly to the overall fire risk. It is not clear what the phrase contribute significantly to the overall fire risk means quantitatively. The NRC staff notes that uncertainties (e.g., assumed failures or assumed hot shorts) that have some impact on total fire risk could impact the RICT calculations for certain SSCs.

Justify that the conservativism that exists in circuit analysis will not have an impact on RICT calculations.

d. The Nine Mile Point 2 PRA uncertainty analysis report reviewed by the NRC staff during the regulatory audit conducted May 4 to 7, 2020 identifies generic and plant specific sources of uncertainty and dispositions them for this application. The report identifies plugging of the intake from the lake (i.e.,

Ultimate Heat Sink (UHS)) to the Service Water (SW) system to be a source of uncertainty. Plugging can occur due to the existence of Zebra Mussels, frazil ice, high winds, and algae. The report explains that a certain initiating event is modelled as the common cause failure of the intake to the SW system from the lake. The report states that this represents considerable uncertainty, but that this uncertainty is treated conservatively. The NRC staff notes that modeling conservatisms can mask the delta risk associated with taking certain components out of service and, therefore, can lead to underestimation of the delta risk and overestimation of a RICT. The NRC staff notes that LCO 3.7.1 (SW and UHS) 7

Conditions A, C, D, E, and F are proposed to be in the RICT program and that RICTs calculated for these conditions could be impacted by this modeling conservatism. Therefore:

i. Given that risk can be masked in the RICT calculation for LCO 3.7.1 Conditions associated with the SW system, justify that the uncertainty associated with modelling the common cause plugging of the lake intake will have no impact on the RICT calculated for components associated with this LCO (e.g., one way to determine the impact is to perform a sensitivity study).

ii. Also, given that risk can be masked in the RICT calculation for TS LCO SSCs in the RICT the program that depend on the SW system, justify that the uncertainty associated with modelling the common cause plugging of the lake intake has no impact on the RICT estimates for other LCOs in scope of the RICT program.

iii. If in the response to part (i and ii) above, it cannot be justified that the uncertainty associated with modelling the common cause plugging of the lake intake has no impact on the application, then explain how this uncertainty will be treated in the RICT program. Include discussion of any additional RMAs that would be implemented.

e. The Nine Mile Point 2 PRA uncertainty analysis report reviewed by the NRC staff during the regulatory audit conducted May 4 to 7, 2020 identifies the development of special data variables as a source of uncertainty. This appears to refer to failure probabilities that were developed based on judgment for non-typical equipment for which failure data was not available. The disposition for this source of uncertainty states that the treatment is considered reasonable and that no justifiable alternative exists. It is not clear to the NRC staff what non-typical equipment is being referred to by the report or the basis used for the failure probabilities developed using judgement. The NRC staff notes that even though non-typical equipment failure data may not be available, the uncertainty associated using failure probabilities based on judgment, none-the-less, represents a source of uncertainty. In light of these observation, address the following:

i. Briefly describe the non-typical equipment referred to in the uncertainty analysis report and explain the approach used to develop failure probabilities for this equipment based on judgment.

ii. Justify that the uncertainty associated with the non-typical equipment failure probabilities developed using judgment will have no impact on the RICTs calculated for components associated with LCOs proposed to be included in the RICT program (e.g., one way to determine the impact is to perform a sensitivity study).

iii. If in the response to part (ii) above, if it cannot be justified that the uncertainty associated with the non-typical equipment failure probabilities developed using judgment will have no impact on the calculated RICTs, then explain how this uncertainty will be treated in the RICT program. Include discussion of any additional RMAs that would be implemented.

f. The Nine Mile Point 2 PRA uncertainty analysis report reviewed by the NRC staff during the regulatory audit conducted May 4 to 7, 2020 identifies the modelling associated with Main Control Room (MCR) abandonment due to loss of control (LOC) as a source of modelling uncertainty. The report explains that no credit was taken for MCR abandonment due to LOC, and therefore, the approach is considered to be conservative and to have only a small impact on the application. The report indicated that the modeling associated with MCR abandonment due to LOC required additional development beyond the modeling that was performed for MCR abandonment due to loss of habitability. The NRC staff notes that modeling conservatisms can mask the delta risk associated with taking certain components out of service and, therefore, can lead to underestimation of the delta risk and overestimation of a RICT. The NRC staff also notes that the fire risk contribution from MCR abandonment scenarios due to LOC may be significant. The report states that MCR abandonment from a main control board fire is the top CDF contributor and second highest risk LERF contributor. In light of these observation, address the following:

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i. Justify that the uncertainty associated with not modelling or crediting MCR abandonment due to LOC will have no impact on the RICT estimates (e.g., one way to determine the impact is to perform a sensitivity study). Include identification of LCO Conditions that could potentially be impacted by this uncertainty.

ii. If, in the response to part (i) above, it cannot be justified that the uncertainty associated with not modelling or crediting MCR abandonment due to LOC will have no impact on the calculated RICTs, then explain how this uncertainty will be treated in the RICT program. Include discussion of any additional RMAs that would be implemented.

RAI 13

As provided by the guidance in NEI 06-09, Revision 0-A, changes to CDF and LERF calculated by a PRA that models the current operating configuration are used to support the RICT program. The guidance in NEI 06-09, Revision 0-A provides several quantitative risk management thresholds values: the calculated RICT, the calculated instantaneous risk, and the cumulative risk increase. When a risk threshold value is exceeded, specific guidance is provided in Table 2-2 of NEI 06-09, Revision 0-A.

RG 1.174 clarifies that, because of the way the acceptance guidelines in RG 1.174 have been developed, the appropriate numerical measures to use when comparing the PRA results with the risk acceptance guidelines are mean values. The risk management thresholds values for the RICT program have been developed based on RG 1.174 and, therefore, the most appropriate measures with which to make a comparison are also mean values. Point estimates are the most commonly calculated and reported PRA results. Point estimates do not account for the state-of-knowledge correlation (SOKC) between nominally independent basic event probabilities, but they can be quickly and simply calculated. Mean values do reflect the SOKC and are always larger than point estimates but require longer and more complex calculations. NUREG-1855, Revision 1, provides guidance on evaluating how the uncertainty arising from the propagation of the uncertainty in parameter values (SOKC) of the PRA inputs impacts the comparison of the PRA results with the guideline values.

Summarize how the SOKC investigation was performed for all the PRA models used to support the RICT application, and how the SOKC will be addressed for the RICT program.

RAI 14

The NRC memorandum dated May 30, 2017, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (ADAMS Accession No. ML17031A269), provides the NRCs staff assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in support of risk-informed decision-making in accordance with the guidance of RG 1.200, Revision 2 (ADAMS Accession No. ML090410014).

Regarding equipment failure probability in the May 30, 2017, memorandum, the NRC staff concludes (Conclusion 8):

The uncertainty associated with failure rates of portable equipment should be considered in the PRA models consistent with the ASME/ANS PRA Standard as endorsed by RG 1.200. Risk-informed applications should address whether and how these uncertainties are evaluated.

Regarding human reliability analysis (HRA), NEI 16-06, Section 7.5, recognizes that the current HRA methods do not translate directly to human actions required for implementing mitigating strategies. Sections 7.5.4 and 7.5.5 of NEI 16-06 describe such actions to which the current HRA methods cannot be directly applied, such as debris removal, transportation of portable equipment, installation of equipment at a staging location, routing of cables and hoses, and those complex actions that require many steps over an extended period, multiple personnel and locations, evolving command and control, and extended time delays. In the May 30, 2017, memorandum, the NRC staff concludes (Conclusion 11):

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[] Until gaps in the human reliability analysis methodologies are addressed by improved industry guidance, HEPs associated with actions for which the existing approaches are not explicitly applicable, such as actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, along with assumptions and assessments, should be submitted to NRC for review.

Regarding uncertainty, Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA-based model results to the RICT program and that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties that could potentially impact the results of an RICT calculation. NEI 06-09, Revision 0-A, also states that the insights from the sensitivity studies should be used to develop appropriate RMAs, including highlighting risk significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions. Uncertainty exists in PRA modeling of FLEX, related to the equipment failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and pre-initiator failure probabilities. Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for RICTs proposed in this application.

The LAR does not address whether FLEX equipment or actions have been credited in the PRA models. The NRC staff notes that the LAR Enclosure 4, Section 5 credits FLEX features for defense-in-depth for the impact of Local Intense Precipitation. To understand the credit that will be taken for FLEX equipment and actions in the RICT Program, address the following separately for the internal events PRA, internal flooding PRA, and FPRA:

a. Discuss whether Exelon has credited FLEX equipment or mitigating actions into the Nine Mile Point Nuclear Station, Unit 2 (Nine Mile Point 2) internal events, including internal flooding, or FPRA models.

If not incorporated or their inclusion is not expected to impact the PRA results used in the RICT program, no additional response is requested and remainder of this question is not applicable.

b. Summarize the supplemental equipment and compensatory actions, including FLEX strategies that have been quantitatively credited for each of the PRA models used to support this application. Include discussion of whether the credited FLEX equipment is portable or permanently installed equipment.

c. Regarding the credited equipment:

i. Discuss whether the credited equipment (regardless of whether it is portable or permanently installed) are like other plant equipment (i.e. SSCs with sufficient plant-specific or generic industry data).

ii. If all credited FLEX equipment is similar to other plant equipment credited in the PRA (i.e.,

SSCs with sufficient plant-specific or generic industry data), responses to items ii and iii below are not necessary.

iii. Provide the failure rates and discuss the data used to support the modeling and provide the rationale for using the chosen data.

iv. Discuss how the failure rates assumed in the PRA for the FLEX equipment are consistent with the most recent industry operational experience.

v. Detail the plant-specific operational experience (e.g., number of failures, number of demands, operational hours) of the Nine Mile Point portable FLEX equipment that are credited in the PRA. Discuss any screening or disregarding of plant-specific data (e.g., design modifications, changes in operating practices). Discuss how the failure rates assumed in the PRA for the FLEX equipment are consistent with relevant plant-specific evidence and operational experience.

10

vi. Discuss whether the uncertainties associated with the parameter values are in accordance with the ASME/ANS PRA standard, as endorsed by RG 1.200, Revision 2.

vii. Perform, justify, and provide results of LCO specific sensitivity studies that assess impact on RICT due to FLEX equipment data and failure probabilities. As part of the response, include the following:

1. Justify values selected for the sensitivity studies, including justification of why the chosen values constitute bounding realistic estimates.

2. Provide numerical results on specific selected RICTs and discussion of the results.

3. Describe how the results of the sensitivity studies will be used to identify RMAs prior to the implementation of the RICT program, consistent with the guidance in Section 2.3.4 of NEI 06-09, Revision 0-A.

d. Regarding HRA, address the following:

i. Discuss whether any credited operator actions related to FLEX equipment contain actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06.

If any credited operator actions related to FLEX equipment contain actions described in Sections 7.5.4 and 7.5.5 of NEI 16-06, answer either item ii or iii below:

ii. Perform, justify, and provide results of LCO specific sensitivity studies that assess impact from the FLEX independent and dependent HEPs associated with deploying and staging FLEX portable equipment on the RICTs proposed in this application. As part of the response, include the following:

1. Justify independent and joint HEP values selected for the sensitivity studies, including justification of why the chosen values constitute bounding realistic estimates.

2. Provide numerical results on specific selected RICTs and discussion of the results.

3. Discuss composite sensitivity studies of the RICT results to the operator action HEPs and the equipment reliability uncertainty sensitivity study provided in response to item c.iii above.

4. Describe how the source of uncertainty due to the uncertainty in FLEX operator actions HEPs will be addressed in the RICT program. Describe specific RMAs being proposed and how these RMAs are expected to reduce the risk associated with this source of uncertainty.

iii. Alternatively, for item ii above, provide information associated with the following items listed in supporting requirements HR-G3 and HR-G7 of the ASME/ANS RA-Sa- 2009 PRA standard to support detailed NRC review:

1. the level and frequency of training that the operators and non-operators receive for deployment of the FLEX equipment (performance shaping factor (a));

2. performance shaping factor (f) regarding estimates of time available and time required to execute the response;

3. performance shaping factor (g) regarding complexity of detection, diagnosis, and decisionmaking, and executing the required response;

4. performance shaping factor (h) regarding consideration of environmental conditions; and 11

5. human action dependencies as listed in supporting requirement HR-G7 of the ASME/ANS RA-Sa-2009 PRA standard.

e. The ASME/ANS RA-Sa-2009 PRA standard defines PRA upgrade as, the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impacts the significant accident sequences or the significant accident progression sequences. Section 1-5 of Part 1 of ASME/ANS RA-Sa-2009 PRA standard states that, upgrades of a PRA shall receive a peer review in accordance with the requirements specified in the Peer Review Section of each respective Part of this Standard [].

Regarding human reliability analysis (HRA), NEI 16-06, Section 7.5, recognizes that the current HRA methods do not translate directly to human actions required for implementing mitigating strategies. Sections 7.5.4 and 7.5.5 of NEI 16-06 describe such actions to which the current HRA methods cannot be directly applied, such as debris removal, transportation of portable equipment, installation of equipment at a staging location, routing of cables and hoses, and those complex actions that require many steps over an extended period, multiple personnel and locations, evolving command and control, and extended time delays. During the regulatory audit conducted May 4 to 7, 2020, the licensee indicated FLEX HRA was developed by following EPRI 3002013018 Human Reliability for Diverse and Flexible Mitigation Strategies and Use of Portable Equipment. This methodology indicates modeling human actions for transporting equipment, connecting hoses, and verifying portable pump operability are categorized as tasks not covered by THERP.

Non-mandatory Appendix I-A of the PRA Standard cites "a different HRA approach to human error analysis..." as a potential PRA upgrade.

Considering the above, applying the current HRA methods to those operator actions for staging and deploying FLEX mitigating strategies could constitute a PRA upgrade. Therefore:

i. Provide an evaluation of the model changes associated with incorporating FLEX mitigating strategies, which demonstrates that none of the following criteria is satisfied: (1) use of new methodology, (2) change in scope that impacts the significant accident sequences or the significant accident progression sequences, and (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences.

ii. As an alternative to item i above, confirm that a focused-scope peer review has been conducted for crediting FLEX in the PRA models. Describe the peer review and status of the resulting F&Os. Provide any remaining open F&Os, along with dispositions for this application.

RAI 15

The LAR proposed TS LCOs include those related to instrumentation and controls (I&C). PRA technical acceptability attributes are provided in Section 2.3.4 of NEI 06-09, Revision 0-A, and in RG 1.200, Revision

2. The LAR does not address whether the I&C is modeled in sufficient detail to support implementation of TSTF-505, Revision 2. The following additional information is requested:

a. Explain how instrumentation is modeled in the PRA. This should include, but not be limited to, the scope of the I&C equipment (e.g., channel, relays logic) and associated TS functions for which an RICT would be applied, and PRA modeling of I&C and associated functions, including the level of detail and inclusion of plant-specific data, etc.

b. For any I&C design basis functions or I&C components in scope of the RICT program not modeled in the PRA, justify why the lack of modeling has no impact on the RICT program, or alternatively, describe any proposed surrogates and justify why the proposed surrogate adequately captures the configuration risk; 12

c. Regarding digital I&C, the NRC staff notes the lack of consensus industry guidance for modeling these systems for plant PRAs to be used in risk-informed applications. In addition, known modeling challenges exist due to the lack of industry data for digital I&C components and the complexities associated with modeling software failures, including common cause software failures. Given these needs and challenges, if the modeling of digital I&C system is included in the RTR model, then address the following:

i. Provide the results of a sensitivity study on the SSCs in the RICT program demonstrating that the uncertainty associated with modeling digital I&C systems has inconsequential impact on the RICT calculations.

ii. Alternatively, identify which LCOs are determined to be impacted by the digital I&C system modeling for which RMAs will be applied during an RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation require additional RMAs.

RAI 16

NEI 06-09, Revision 0-A states concerning the quality of the PRA model that:

RG 1.174, Revision 1, and RG 1.200, Revision 1 define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change.

LAR Attachment 6 lists six implementation items that must be complete prior to implementation of the RICT program. The attachment table (1) identifies for each implementation item the modeling that needs to be incorporated (2) makes the commitment to update the PRA model and (3) provides other details such as clarification that the success criteria associated with the PRA update will match the design basis success criteria. The LAR attachments do not explicitly state whether the fire PRA models as well as the internal events PRA models will be updated or whether the RTR model will be updated.

In light of the observations above, clarify that the internal events and fire PRA models and the RTR model will all be updated to incorporate the implementation items prior to implementation of the RICT program. Otherwise, justify not incorporating the update into the fire PRA or RTR model.

RAI 17

Key Principle 5 of RG 1.177, Revision 1 (ADAMS Accession No. ML100910008) pertains to performance monitoring (ADAMS Accession No. ML100910008). It states:

The licensee should consider implementation and performance monitoring strategies formulated to ensure (1) that no adverse safety degradation occurs because of the changes to the TS and (2) that the engineering evaluation conducted to examine the impact of the proposed changes continues to reflect the actual reliability and availability of TS equipment that has been evaluated. This will ensure that the conclusions that have been drawn from the evaluation remain valid.

Similarly, RG 1.174, Revision 3 (ADAMS Accession No. ML17317A256) provides guidance on implementation and monitoring for any risk-informed licensing basis changes:

The licensee should propose monitoring programs that adequately track the performance of equipment that, when degraded, can affect the conclusions of the licensees engineering evaluation and integrated decision-making that support the change to the licensing basis. The program should be capable of trending equipment performance after a change has been implemented to demonstrate that performance is consistent with the assumptions in the traditional engineering and probabilistic analyses conducted to justify the change. . . . The program should be structured such that (1) SSCs are monitored commensurate with their safety 13

importance . . . , (2) feedback of information and corrective actions is timely, and (3) degradation in SSC performance is detected and corrected before plant safety can be compromised.

LAR Attachment 1, Section 2.3 states that the application of a risk-informed completion time (RICT) will be evaluated using the guidance provided in NEI 06-09, Revision 0-A (ADAMS Accession No. ML122860402). The LAR further states:

The . . . methodology satisfies the five key safety principles specified in Regulatory Guide 1.177 relative to the risk impact due to the application of a RICT.

The LAR did not explain how the RICT program satisfies the fifth key safety principle of RG 1.177 and RG 1.174, specifically with respect to the ability of the monitoring program to adequately and timely track the performance of equipment that, when degraded, can affect the conclusions of the licensees RICT evaluations.

Describe the performance monitoring strategies proposed for the RICT program. Justify how the program meets Key Principle 5 of RG 1.177 and RG 1.174. Include description and justification on:

a. Whether and how SSCs are monitored commensurate with their safety importance;

b. Timeliness of the feedback of information and corrective actions,

c. How degradation in SSC performance is detected and corrected before plant safety can be compromised;

d. How it is ensured that the PRA used in the RICT program continues to reflect the actual reliability and availability of TS equipment.

Fire PRA Section 50.36(c)(2) of 10 CFR requires TSs to contain LCOs that describe the lowest functional capability of equipment required for safe operation of a plant and requires to follow any remedial actions permitted by the TSs. The remedial actions need to be completed within a set time frame commonly referred to as a CT or allowed outage time. The risk informed completion time program that Exelon has requested to adopt at Nine Mile Point 2 is one way of establishing or changing a CT using a risk-informed approach that relies on PRAs, including the fire PRA (FPRA).

RG 1.200 states, NRC reviewers, [will] focus their review on key assumptions and areas identified by peer reviewers as being of concern and relevant to the application. The NRC staff evaluates the acceptability of the PRA for each new risk-informed application and, as discussed in RG 1.174, recognizes that the acceptable technical adequacy of risk analyses necessary to support regulatory decisionmaking may vary with the relative weight given to the risk assessment element of the decisionmaking process. The NRC staff notes that the calculated results of the PRA are used directly to calculate a RICT. The NRC staff requests additional information on the following FPRA issues that have been previously identified as potentially key FPRA assumptions.

RAI 18

The LAR provides the history of the FPRA peer review but does not discuss methods used in the FPRA. In Section 4 of Enclosure 9 to the LAR, it is stated that NUREG/CR-6850, EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (ADAMS Accession Nos. ML052580075, ML052580118, and ML103090242) was used to guide the development of the FPRA. It is unclear in the LAR whether methods may have been used in the FPRA that deviate from guidance in NUREG/CR-6850 or other acceptable guidance (e.g., frequently asked questions (FAQs), NUREGs, or interim guidance documents).

a. Identify methods used in the FPRA that deviate from guidance in NUREG/CR-6850 or other acceptable guidance.

14

b. If such deviations exist, then justify their use in the FPRA, any impact on the RICT, and describe and justify any replacement methods to be used.

RAI 19

The key factors used to justify using transient fire-reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, NRC letter to Nuclear Energy Institute (NEI), Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires (ADAMS Package Accession No. ML12172A406).

In Table E9-3 of Enclosure 9 to the LAR, it is stated that HRR is an input parameter to the analysis to translate a fire initiating event into a set of consequences. If any reduced transient HRRs below the bounding 98 percent HRR of 317 kilowatts (kW) from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs. Include in this discussion:

a. Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.

b. A description for each location where a reduced HRR is credited and a description of the administrative controls that justify the reduced HRR, including how location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations and the types and quantities of combustible materials needed to perform maintenance. Also, include discussion of the personnel traffic that would be expected through each location.

c. The results of a review of records related to compliance with the transient combustible and hot work controls.

RAI 20

In Section 4 of Enclosure 9 to the LAR, it is stated that the FPRA methods were based, in part, on published FPRA FAQs, but the specific FAQs were not identified. FAQ 13-0004, Clarifications on Treatment of Sensitive Electronics (ADAMS Accession No. ML13322A085), provides supplemental guidance for application of the damage criteria provided in Sections 8.5.1.2 and H.2 of NUREG/CR-6850, Volume 2, for solid-state and sensitive electronics.

a. Describe the treatment of sensitive electronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents).

b. If the approach cannot be justified to be consistent with FAQ 13-0004, then justify that the treatment of sensitive electronics has no impact on the RICT calculations.

c. If the approach cannot be justified as consistent with FAQ13-004, and it has an impact on the RICT calculations, describe and justify how this issue will be resolved.

RAI 21

In Table E9-3 of Enclosure 9 to the LAR, the human error probabilities (HEP) used in the FPRA are discussed, but does not explicitly identify the method (i.e., guidance) used to determine the GEP values or list the minimum joint HEP value. NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines - Final Report, (ADAMS Accession No. ML12216A104), discusses the need to consider a minimum value for the joint probability of multiple human failure events in HRAs.

15

NUREG-1921 refers to Table 2-1 of NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA) (ADAMS Accession No. ML051160213), which recommends that joint HEP values should not be below 1E-5. Table 4-4 of EPRI TR-1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. Therefore, the guidance in NUREG-1921 allows for assigning joint HEPs that are less than 1E-5, but only through assigning proper levels of dependency. The NRC staff notes that underestimation of maximum joint probabilities could result in non-conservative RICTs of varying degrees for different inoperable SSCs.

The LAR does not provide this information and does not explain what minimum joint HEP value is currently assumed in the internal events PRA or FPRA. Also, even if the assumed minimum joint HEP values are shown to have no impact on the current risk estimates, it is not clear to the NRC staff how it will be ensured that the impact remains minimal for future PRA model revisions.

a. Explain what minimum joint HEP (JHEP) value was assumed in the internal events and the fire PRAs.

b. If a minimum JHEP value less than 1E-6 was used in the internal events PRA, or less than 1E-5 was used in the FPRA, then provide a description of the sensitivity study that was performed and the quantitative results that justify that the minimum JHEP value has no impact on the RICT application.

c. If, in response part (b), it cannot be justified that the minimum JHEP value has no impact on the application, then confirm that each JHEP value used in the internal events PRA below 1E-6 and each JHEP used in the FPRA below 1E-5 includes its own justification that demonstrates the inapplicability of the NUREG-1792 lower value guideline (i.e., utilizing the dependency factors identified in NUREG 1921). Provide an estimate of the number of these JHEP values below 1E-6 for the internal events PRA and below 1E-5 for the FPRA, discuss the range of values, and provide at least two different examples, separately for the internal events and the fire PRAs, where this justification is applied.

d. During the regulatory audit conducted May 4 to 7, 2020, the licensee identified that no JHEPs less than 1.0E-05 were used in the FPRA for the RICT although no lower limit on the JHEP was imposed. Absent an imposed lower limit, indicate how do you intend to track JHEPs in your FPRA as it evolves. Also, indicate a plan for your evolving FPRA to address JHEPs lower than 1.0E-05 consistent with part b) or part c) of this RAI.

RAI 22

In Table E9-3 of Enclosure 9 to the LAR, it is stated that the heat release rates from NUREG-2178, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE -FIRE), Volume 1, Peak Heat Release Rates and Effect of Obstructed Plume, (ADAMS Accession No. ML16110A14016) are used. NUREG-2178, Volume 1 contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction. Additionally, NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet.

a. If obstructed plume modeling was used, then indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet.

b. Justify any modeling in which the base of an obstructed plume is located at less than one half of the cabinets height.

RAI 23

In Section 4 of Enclosure 9 to the LAR, it is stated that the FPRA methods were based, in part, on published FPRA FAQs, but the specific FAQs were not identified. Guidance in FAQ 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440 volts (V). With respect to Bin 15, as discussed in Chapter 6, it clarifies the meaning of robustly or well-sealed. Thus, for cabinets of less than 440 V, fires from 16

well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440 V and higher, the original guidance in Chapter 6 indicates that Bin 15 panels that house circuit voltages of 440 V or greater are counted because an arcing fault could compromise panel integrity (an arcing fault could burn through the panel sides, but this should not be confused with the high energy arcing fault type fires). FPRA FAQ 14-0009, Treatment of Well-Sealed MCC Electrical Panels Greater than 440V (ADAMS Accession No. ML15119A176), provides the technique for evaluating fire damage from motor control center (MCC) cabinets having a voltage greater than 440 V. Therefore, propagation of fire outside the ignition source panel must be evaluated for all MCC cabinets that house circuits of 440 V or greater.

a. Describe how fire propagation outside of well-sealed MCC cabinets greater than 440 V is evaluated.

b. If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, justify that this approach has an inconsequential impact on the RICT calculations.

RAI 24

In Section 4 of Enclosure 9 to the LAR, it is stated that the FPRA methods were based, in part, on published FPRA FAQs, but the specific FAQs were not identified. NUREG/CR-6850, Section 6, and FAQ 12-0064, Hot Work/Transient Fire Frequency Influence Factors" (ADAMS Accession No. ML12346A488), describe the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance:

a. Indicate whether the methodology used to calculate hot work and transient fire frequencies applies influencing factors using NUREG/CR-6850 guidance or FAQ 12 0064 guidance.

b. Indicate whether administrative controls are used to reduce transient fire frequency, and if so, describe and justify these controls.

c. Indicate whether you have any combustible administrative control that were not meet and discuss your treatment of not meeting these administrative controls for the assignment of transient fire frequency influence factors. For those cases where you have violations and have assigned an influence factor of 1 (low) or less, indicate the value of the influence factors you have assigned and provide your justification.

d. If you have assigned an influencing factor of 0 to maintenance, occupancy, storage, or hot work for any fire physical analysis units provide justification.

e. If a weighting factor of 50 was not used in any fire physical analysis unit, justify this in light of the guidance in FAQ 12-0084.

RAI 25

In Section 4 of Enclosure 9 to the LAR, it is stated that the FPRA methods were based, in part, on published FPRA FAQs, but the specific FAQs were not identified. Traditionally, the cabinets on the front face of the main control board (MCB) have been referred to as the MCB for purposes of FPRA. Appendix L of NUREG/CR-6850, EPRI/NRC Fire PRA Methodology for Nuclear Power Facilities (ADAMS Accession Nos.

ML052580075), provides a refined approach for developing and evaluating those fire scenarios. Fire PRA FAQ 14-0008, Main Control Board Treatment, dated July 22, 2014 (ADAMS Accession No. ML14190B307),

clarifies the definition of the MCB and effectively provides guidance for when to include the cabinets on the back side of the MCB as part of the MCB for FPRA. It is important to distinguish between MCB and non-MCB cabinets, because misinterpretation of the configuration of these cabinets can lead to incomplete or incorrect fire scenario development. This FAQ also provides several alternatives to NUREG/CR-6850 for using Appendix L to treat partitions in an MCB enclosure. Therefore, address the following:

a. Briefly describe the MCB configuration and describe whether cabinets on the rear side of the MCB are a part of the MCB.

17

b. If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure, describe and justify the approach used to develop fire scenarios in the MCB and determine the frequency of those scenarios.

c. If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure, describe and justify how the fire scenarios for the backside cabinets are developed.

d. Describe and justify the impact of the current treatment of the MCB and those cabinets on the rear side of the MCB on the RICT calculations.

Best Regards, Michael L. Marshall, Jr.

Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 301-415-2871 Docket Number: 050-410 18