ML20307A659

NRR E-mail Capture - Request for Additional Information Regarding TSTF-505 and 50.69 Applications

ML20307A659
Person / Time
Site:	Clinton
Issue date:	10/27/2020
From:	Joel Wiebe NRC/NRR/DORL/LPL3
To:	Nicely K Exelon Generation Co
References
Download: ML20307A659 (16)

Text

From:

Wiebe, Joel Sent:

Tuesday, October 27, 2020 1:33 PM To:

Nicely, Ken M.:(GenCo-Nuc) (ken.nicely@exeloncorp.com)

Subject:

Request for Additional Information Regarding TSTF-505 and 50.69 Applications

==

Introduction:==

The Nuclear Regulatory Commission (NRC) staff is performing an integrated review of the two license amendment requests (LARs) to the extent that information is applicable (e.g.,

probabilistic risk analysis (PRA) acceptability). Request for additional information[1] (RAIs) have been developed to support both applications, in addition to those RAIs that are only applicable for a certain application. This integrated review also includes multiple divisions across the office of Nuclear Reactor Regulation (NRR) to ensure the staff findings are appropriately assessed against all of the five key principles of risk-informed decision making. As follows, to identify a specific RAI for the responsible division, it is denoted as follows:

DRA/APLA RAI XX: Division of Risk Assessment, PRA Licensing Branch A DRA/APLB RAI XX: Division of Risk Assessment PRA Licensing Branch B DRA/APLC RAI XX: Division of Risk Assessment PRA Licensing Branch C DEX/EEOB RAI XX: Electrical Engineering Operating Reactors Branch DSS/STSB RAI XX: Technical Specifications Branch Background and Regulatory Basis:

By letters dated April 30, 2020, Exelon Generation Company, LLC (Exelon, the licensee) submitted two LARs to the NRC for Facility Operating License No. NPF-62 for Clinton Power Station, Unit 1 (Clinton):

(1) Application to adopt 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors" dated April 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20121A241).

(2) Application to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b" dated April 30, 2020 (ADAMS Accession No. ML20121A178).

Section 3.1.1 of the LAR to adopt 10 CFR 50.69 states that Exelon will implement the risk categorization process in accordance with Nuclear Energy Institute (NEI) NEI 00-04, 10 CFR 50.69 structure, system, and components (SSC) Categorization Guideline, Revision 0 (ADAMS Accession No. ML052900163}, as endorsed by Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance (ADAMS Accession No. ML061090627).. Section 1.0 of the LAR to adopt TSTF-505 states that Exelon will adhere to NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines, Revision 0, (ADAMS Accession No. ML12286A322) and implement the Risk Informed Completion Times (RICT) program in accordance with TSTF-505, Revision 2 (ADAMS Accession No. ML18183A493).

[1] As a part of agency efforts to improve efficiency and effectiveness, RAIs have been structured to include sufficient details to render a staff finding and potentially obviate the need for a second round of RAIs.

The NRC staff reviews for risk-informed licensing applications is comprised of ensuring that these proposed changes meet the five key principles of risk-informed decision making outlined in NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light Water Reactor] Edition. The following regulatory guidance documents in addition to the above industry guidance includes accepted methodologies and practices to ensure that the proposed changes meet the technical adequacy the staff has determined is appropriate for a specific risk-informed application.

RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, provides guidance on the use of PRA findings and risk insights in support of changes to a plants licensing basis. Revision 3 of RG 1.174 provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, describes an acceptable approach for determining whether the acceptability of the base PRA, in total or the parts, that is used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light water reactors. It endorses, with clarifications and qualifications, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) PRA standard, ASME/ANS RA-Sa-2009.

RG 1.201, Revision 1, endorses the categorization process described in NEI 00-04, Revision 0, with clarifications, limitations, and conditions. RG 1.201, Revision 1, states that the applicant is expected to document, at a minimum, the technical adequacy of the internal initiating events PRA. Licensees may use either PRAs or alternative approaches for hazards other than internal initiating events.

NUREG 1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, provides guidance on how to treat uncertainties associated with PRA in risk-informed decision making. The guidance fosters an understanding of the uncertainties associated with PRA and their impact on the results of the PRA and provides a pragmatic approach to addressing these uncertainties in the context of the decision making.

The NRC staff requests additional information to further assess the proposed adoption of 10 CFR 50.69 and TSTF-505, Revision 2 at Clinton for consistency with identified RGs and applicable industry guidance.

DRA/APLA RAI 01 - Open Internal Events PRA Facts and Observations (F&O)

[Applicable for TSTF-505 and 10 CFR 50.69]

a. LAR Enclosure 2, Table E2-1 presents the dispositions for two F&Os (i.e., F&Os 1-32 and 1-

34) that remain open after the internal events F&O closure review performed in November 2019. These F&Os both address the same concern, stating in F&O 1-32 that potentially risk significant combinations of HFEs [Human Error Events] are not captured through the current approach, due to the chosen truncation level for the dependency identification (5E-9 / 5E-10 for CDF/LERF) in conjunction with the elevated HEP level chosen (0.1). The LAR states that the existing cutsets were reviewed and only a few dependent Human Error Probability (HEP) combinations with some level of dependency were identified. The LAR concluded that the overall risk results were not substantially impacted by the current treatment.

Lowering the truncation level is likely to reveal further HEP combinations that require dependency analysis and possible adjustment of the joint probability for the combination.

Therefore, it is not clear to NRC staff that the current treatment of HEP dependency has no impact on the RICT calculation for plant configurations and SSC categorization.

In light of the observations above, address the following:

i. Provide justification that the additional HEP combinations using a lower quantification level will not adversely impact the RICT calculations or SSC categorizations for each of the risk-informed applications.

ii.

Alternatively, propose a mechanism that ensures the results of the dependency analysis associated with F&O 1-32 and 1-34 are resolved prior to implementation of the RICT and SSC categorization programs.

b. The disposition for F&O 1-32 further states that a floor value of 1E-06 or 5E-07 may be imposed on the dependent joint HEP depending on the timing of operator actions. Section 4.4.3.2 of NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA) states in part, the PRA analyst can define the significant contributors by using typical PRA criteria [], such as importance measure thresholds as well as other qualitative and quantitative considerations (ADAMS Accession No. ML051160213). It is not clear to the NRC staff what assumptions (i.e., qualitative or quantitative considerations) were made in the internal events PRA for the application of the minimum joint HEP values and how they were determined to be sufficient. In light of the observations above, address the following:

Confirm what minimum joint HEP values were used in the internal events PRA. If a minimum joint HEP value less than 1E-06 was used in the internal events PRA, then provide sufficient justification that demonstrates that for each joint HEP value used below 1E-6 issue-relevant human actions have been appropriately addressed. The justification should support that the internal events PRA minimum values used have no adverse impact on the RICT and SSC risk-informed applications.

DRA/APLA RAI 02 - Peer Review History for the Internal Events, including Internal Flooding, PRA

[Applicable for TSTF-505 and 10 CFR 50.69]

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 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.

LAR Enclosure 2 states that the last full scope peer review for the internal events PRA was conducted in October 2009 and that F&O closure reviews to close out F&Os from the 2009 review were conducted in December 2018 and November 2019. The LAR does not discuss the internal events and internal flood PRA model changes made between October 2009 and when the F&O closure reviews were performed in to improve the model or to incorporate changes to reflect the as-built, as-operated plant. Given the significant length of time between the last full scope peer review and the F&O closure reviews, address the following:

Summarize the significant model changes performed for the internal events (including internal flood) PRA since October 2009 and for each change justify why or why not the change meets the definition of a PRA upgrade as defined in the ASME/ANS RA-Sa-2009 PRA Standard (e.g.,

Changing to different PRA software or a different HRA methods are examples of possible PRA upgrades).

DRA/APLA RAI 03 - System and Surrogate Modeling Used in the PRA Models

[Applicable for TSTF-505]

The NRC SE for NEI 06-09 specifies that the LAR should provide a comparison of the Technical Specifications (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 Limited conditions of operation (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 TS LCO Conditions, the table explains that the associated SSCs are not modeled in the PRAs but will be conservatively represented using a surrogate event. For some LCOs, the LAR did not provide enough description of the surrogate PRA modeling that will be used in the RICT calculations for NRC staff to understand whether the modeling will be acceptable. Therefore, address the following:

LAR Table E1-1 states for TS LCO 3.3.6.5 (Relief and Low-Low Set instrumentation)

Condition A (One trip system inoperable) that the [r]elief function is not modeled and that the Low-Low Set Point values are used as surrogate for the trip system. It is not clear to the NRC staff how Clinton identified that the Low-Low Set Point values proposed to be used as a surrogate reflect the failure of the function covered by the applicable TS LCO Condition. Therefore:

Confirm which SSC (i.e., Low-Low Set instrumentation or over pressurization relief function ) is not explicitly included in the PRA model. For the confirmed SSC that is not explicitly modeled, identify the surrogate proposed to be used in the PRA model and provide justification that ensures the surrogate appropriately represents the failure of the design function associated with TS LCO Condition 3.3.6.4.A.

DRA/APLA RAI 04 - Total Risk Consideration of State-of-Knowledge Correlation and Modeling Updates

[Applicable for TSTF-505 and 10 CFR 50.69]

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 threshold 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 calculated. Mean values 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.

LAR Enclosure 5, Section 2 states that the total CDF and LERF values presented in Enclosure 5 for Clinton Power Station (CPS) are point estimate values. NRC staff notes that for CPS, the total CDF of 8.8E05 per year and begins to approach the RG 1.174, Revision 3 guidelines for total CDF without considering the risk increase due to SOKC. The NRC staff notes based on RG 1.174 and Section 6.4 of NUREG-1855, Revision 1, for a Capability Category II risk evaluation, the mean values of the risk metrics (total and incremental values) need to be compared against the risk acceptance guidelines. Additionally, NRC staff notes that the current PRA models might potentially be updated in response to information requests (e.g., response to DRA/APLA RAI 01 regarding unresolved F&Os, DRA/APLA RAI 05 regarding FLEX modeling, etc.). Accordingly, an increase in CDF due to any updates in combination with an increase resulting from SOKC could potentially impact the conclusions of these risk-informed applications.

In light of the observations above, address the following:

a) Provide a summary of how the SOKC investigation was performed for the base Clinton PRA models used to support the risk-informed applications (i.e., TSTF-505 and 10 CFR 50.69).

b) Provide a summary of how the SOKC will be addressed for the risk-informed applications (i.e., based upon the risk metrics to be considered), and explain how this process/approach is consistent with NUREG-1855, Revision 1 (i.e., this should include

increases due to potential updates to PRA models that may be identified in response to NRC staff RAIs affecting the total risk for Unit 1 for conformance with the RG 1.174 risk acceptance criteria for CDF and LERF. Also, include identification of the fire PRA parameters for which SOKC was applied in the parametric uncertainty analysis of fire events.

DRA/APLA RAI 05 - Credit for FLEX Equipment and Actions

[Applicable for TSTF-505 and 10 CFR 50.69]

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 NRC staffs position concerning incorporating mitigating strategies (FLEX) into a PRA in support of risk-informed decision making in accordance with the guidance in RG 1.200, Revision 2 (ADAMS Accession No. ML090410014).

To complete the NRC staffs review of the FLEX strategies modeled in the PRA, the NRC staff requests the following information for the IEPRA (includes internal floods) and FPRA, as appropriate.

a) Clarify whether permanent or portable FLEX equipment and associated operator actions are credited in the PRAs used to support the applications, identifying the specific PRA(s) that include such credit. If FLEX is not credited in the PRAs, then no response to parts (b) and (c) of this RAI is requested. If FLEX is credited in the PRAs and this credit is not expected to impact the PRA results used in the categorization process or RICT program (e.g., permanently installed equipment, hardened vent containment), then provide sufficient justification to confirm this conclusion, and no response to parts (b) and (c) of this question is requested.

b) If the FLEX equipment or operator actions have been credited, and their inclusion is expected to impact the PRA results used in the categorization process and RICT program, provide the following information separately for the IEPRA (includes internal floods) and FPRA, as appropriate:

i.

A discussion detailing the extent of incorporation, i.e. summarize the supplemental equipment and compensatory actions that have been quantitatively credited for each of the PRA models used to support both risk-informed applications.

ii.

Discuss the data and failure probabilities used to support the FLEX modeling and provide the rationale for using the chosen data. Include discussion on whether the uncertainties associated with the parameter values are in accordance with the applicable supporting requirements (SRs) in the ASME/ANS PRA Standard, as endorsed by RG 1.200, Revision 2.

iii.

Discuss the methodology used to assess human error probabilities for the FLEX operator actions. The discussion should include:

1. A summary of how the impact of the plant-specific human error probabilities and associated scenario-specific performance shaping factors listed in (a)-(j) of SR HR-G3 of the ASME/ANS RA-Sa-2009 PRA standard were evaluated.

2. Whether maintenance and testing procedures for the portable equipment were reviewed for possible pre-initiator human failures that renders the equipment unavailable during an event, and whether the probabilities of the pre-initiator human failure events were assessed as described in HLR-HR-D of the ASME/ANS RA-Sa-2009 PRA standard.

3. For licensees procedures governing the initiation or entry into mitigating strategies, identify specific areas which could be ambiguous, vague, or not explicit. Provide a discussion detailing the technical bases for probability of failure to initiate mitigating strategies.

c) 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 ASME/ANS RA-Sa-2009 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.

i.

Provide an evaluation of the model changes associated with incorporating non-safety related SSCs that were included following the FLEX mitigation strategies (permanently installed and/or portable), 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, (3) change in capability that impacts the significant accident sequences or the significant accident progression sequences, OR ii.

Propose a mechanism to ensure that a focused-scope peer review is performed on the model changes associated with incorporating mitigating strategies, and associated F&Os are resolved to Capability Category II prior to implementation of the 10 CFR 50.69 categorization process and RICT program.

DRA/APLA RAI 06 - Performance Monitoring

[Applicable for TSTF-505 and 10 CFR 50.69]

For the TSTF-505 LAR, Section 2.3 of LAR Attachment 1 states that the application of a RICT will be evaluated using the guidance provided in NEI 06-09, Revision 0-A, which was approved by the NRC on May 17, 2007 (ADAMS Accession No. ML071200238). The NRC SE for NEI 06-09, Revision 0-A, states, [t]he impact of the proposed change should be monitored using performance measurement strategies. Furthermore, for the adoption of 10 CFR 50.69 using NEI 00-04, the guidance discusses the use of 10 CFR 50.65, the Maintenance Rule, as a way to monitor RISC-1 and RISC-2 SSCs with the clarifications listed in Section 12 of NEI 00-04. Both NEI 00-04 and NEI 06-09 consider the use of NUMARC 93-01, Revision F (ADAMS Accession No. ML18120A069), as endorsed by RG 1.160, Revision 4 (ADAMS Accession No. ML18220B281), for the implementation of the Maintenance Rule. NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.

Furthermore, Section 2.3 of the TSTF-505 LAR Attachment 1 states:

In addition, the NEI 06-09-A, Revision 0 methodology satisfies the five key safety principles specified in RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decision-making: Technical Specifications," dated August 1998 (ADAMS Accession No. ML003740176), relative to the risk impact due to the application of a RICT.

Section 3.4 of the LAR to adopt 10 CFR 50.69 states in part, [s]ubsequent performance monitoring and PRA updates required by the rule will continue to capture this data and provide timely insights into the need to account for any important new degradation mechanisms.

The staff position C.3.2 provided in RG 1.177 for meeting the fifth key safety principle (specifically for TSTF-505), in addition to the endorsed guidance provided in NEI 00-04 (applicable for 10 CFR 50.69), acknowledges the use of performance criteria to assess degradation of operational safety over a period of time. It is unclear to NRC staff how the

licensees processes for each of the risk-informed applications captures performance monitoring for the SSCs within-scope of each application. In light of these observations, address either (i) or (ii) below:

i) Confirm that the Clinton Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in the NRC-endorsed guidance in NUMARC 93-01.

OR ii) Describe the approach/method used by Clinton for SSC performance monitoring as described in Regulatory Position C.3.2 referenced in RG 1.177 for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative or quantitative) along with the appropriate risk metrics for each application, and explain how the approach and criteria demonstrates the intent to monitor the potential degradation of SSCs for the applicable process of the risk-informed application (i.e., for 10 CFR 50.69, paragraphs (d)(1) and (e)(2) and for TSTF-505, Section 3.4 of NEI 06-09, Revision 0).

DRA/APLA PRA MODEL UNCERTAINTY ANALYSIS RESULTS

[Applicable for TSTF-505 and 10 CFR 50.69]

The NRC staff safety evaluation to NEI 06-09, Revision O, specifies that the LAR should identify key assumptions and sources of uncertainty and to assess/disposition each as to their impact on the RMTS application. LAR Enclosure 9, Tables E9-1, E9-2, and E9-3 identify the key assumptions and sources of uncertainty for the internal events PRA, transition to the RTR model, and the fire PRA and provides dispositions for each source of uncertainty for this TSTF-505 application. NRC staff reviewed the dispositions provided in LAR Tables E9-1, E9-2 and E9-3 to the key assumptions and sources of modeling uncertainty and noted that not all uncertainties that appeared to have the potential to impact the RICT calculations seemed fully resolved. NRC staff notes that LAR Tables E9-1, E9-2, and E9-3 do not address certain sources of uncertainty identified in those reports that appear to NRC staff to have the potential to impact the RICT application. Therefore, address the following:

a) LAR Enclosure 9, Table E9-3 identifies cable selection as a source of fire PRA modeling uncertainty because of conservatisms in the approach. The LAR states that some components were conservatively assumed to be failed based on lack of cable data in certain locations. Such components were referred to as Unknown Location (UNL) components which were stated to be mostly limited to Balance of Plant systems. The LAR states that two sensitivity analysis were performed to determine the impact of the treatment of UNL components. Based on the sensitivity studies, the LAR states it is concluded that the methodology for the Cable selection task does not introduce any epistemic uncertainties. The LAR does not provide the results of the uncertainty analyses. NRC staff notes that conservatism in PRA modeling could have a nonconservative impact on the RICT calculations while having only a small or modest impact on total CDF or LERF. It is not clear to NRC staff that the modeling assumption applied to untraced cables has no impact on the RICT program calculations. If an SSC is part of a system not credited in the fire PRA or it is supported by a system that is assumed to always fail, then the risk increase due to taking that SSC out of service is masked. Therefore, address the following:

i. Identify the systems or components that are assumed to be always failed in the PRA or not included in the PRA (due to lack of cable tracing or other reasons). Justify that this assumption has an inconsequential impact on the RICT calculations.

Ii If in part (i) above, it is be determined that the cited assumption has a consequential impact on the estimated RICTs, then identify what programmatic changes will be considered to compensate for this uncertainty and the basis for their consideration.

b) LAR Enclosure 9, Table E9-3 identifies post-fire HRA as a source of fire PRA modeling uncertainty because fire HEPs must be adjusted to consider the additional challenges present given a fire. The LAR states that industry consensus modeling approaches are used and concludes that this source of uncertainty does not introduce any epistemic uncertainties that would require sensitivity treatment and that additional RMAs were, therefore, not required.

Accordingly, it Is not clear why the need for additional RMAs to mitigate the impact of this uncertainty was not considered. Therefore, address the following:

i. Justify that the uncertainty associated with post-fire HRA modeling does not have a consequential impact on calculated RICTs for components supporting TS LCO Conditions in the RICT program.

ii. If in response to part (i) above: it is determined that the cited source of uncertainty has a consequential impact on the estimated RICTs, then identify what programmatic changes will be considered to compensate for this uncertainty.

PRA Licensing Branch B (APLB)

Fire PRA Questions 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 relatively extensive and detailed reviews of fire PRAs undertaken in support of LARs to transition to NFPA-805 determined that implementation of some of the complex fire PRA methods often used nonconservative and over-simplified assumptions to apply the method to specific plant configurations. Some of these issues were not always identified in F&Os by the peer review teams but are considered potential key assumptions by the NRC staff because using more defensible and less simplified assumptions could substantively affect the fire risk and fire risk profile of the plant. 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 decision making may vary with the relative weight given to the risk assessment element of the decision making process. The NRC staff notes that the calculated results of the PRA are used directly to calculate a RICT which subsequently determines how long SSCs (both individual SSCs and multiple, unrelated SSCs) controlled by TS can remain inoperable. Therefore, the PRA results are given a very high weight in a TSTF-505 application and the NRC staff requests additional information on the following issues that have been previously identified as potentially key fire PRA assumptions.

DRA/APLB RAI 01 - Reduced Transient Heat Release Rates (HRRs)

The key factors used to justify using transient fire reduced HRRs below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, letter from Joseph Giitter, U.S. NRC, to Biff Bradley, NEI, "Recent Fire PRA Methods Review Panel Decisions and Electrical Power

Research Institute (EPRI) 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires," (ADAMS Package Accession No. ML12172A406).

If any reduced transient HRRs below the bounding 98% HRR of 317 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.

DRA/APLB RAI 02 - Treatment of Sensitive Electronics 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) As an alternative to item b above, add an implementation item to replace the current approach with an acceptable approach prior to the implementation of the RICT program.

Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.

DRA/APLB RAI 03 - Minimum Joint Human Error Probability 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 (HFEs) in HRAs.

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 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 LAR does not provide this information and does not explain what minimum joint HEP value is currently assumed in the FPRA. Also, even if the assumed minimum joint HEP values are shown to have no impact on the current FPRA 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. In light of these observations:

a) Explain what minimum joint HEP value was assumed in the FPRA.

b) If a minimum joint HEP value less than 1E-05 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 joint HEP value has no impact on the RICT application.

c) If, in response part (b), if it cannot be justified that the minimum joint HEP value has no impact on the application, then provide the following:

i. Confirm that each joint HEP value used in the FPRA below 1E-5 includes its own justification that demonstrates the inapplicability of the NUREG-1792 lower value guideline (i.e., using such criteria as the dependency factors identified in NUREG-1921 to assess level of dependence). Provide an estimate of the number of these joint HEP values below 1.0E-5, discuss the range of values, and provide at least two different examples where this justification is applied.

ii. If joint HEP values used in the FPRA below 1E-5 cannot be justified, add an implementation item to set these joint HEPs to 1E-5 in the FPRA prior to the implementation of the RICT program.

DRA/APLB RAI 04 - Well-Sealed MCC Cabinets Guidance in Frequently Asked Question 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440 V. With respect to Bin 15 as discussed in Chapter 6, it clarifies the meaning of "robustly or well-sealed." Thus, for cabinets of 440 V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440 V and higher, the original guidance in Chapter 6 remains and requires that Bin 15 panels which 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). Fire PRA FAQ 14-0009, Treatment of Well-Sealed MCC Electrical Panels Greater than 440 V (ADAMS Accession No. ML15119A176) provides the technique for evaluating fire damage from 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, provide justification for using this approach as this is contrary to the guidance.

PRA Licensing Branch C (APLC) TSTF-505 RAIs DRA/APLC RAI 01 - Determination of Seismic LERF Penalty Section 2.3.1, Item 7, of NEI 06-09, Revision 0-A (ADAMS Accession No. ML12286A322),

states that the impact of other external events risk shall be addressed in the [Risk Managed TS] RMTS program, and explains that one method to do this is by performing a reasonable bounding analysis and applying it along with the internal events risk contribution in calculating the configuration risk and the associated RICT. The NRC staffs safety evaluation for NEI 06-09 (ADAMS Accession No. ML071200238) states that [w]here [probabilistic risk assessment] PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

A seismic PRA (SPRA) model is not available for CPS, Unit 1, and the seismic hazard cannot be screened out for the RICT application. Details of the approach for determining the seismic penalty are provided in Section 3 of Enclosure 4 to the LAR which states that the seismic Large Early Release Frequency (SLERF) estimate is conservatively used in the RICT process and that the proposed SLERF penalty was determined by multiplying the estimated seismic Core Damage Frequency (SCDF) by an estimated average seismic conditional large early release probability (CLERP). The LAR explains that the average seismic CLERP (SCLERP) was determined using: (1) an estimation of the breakdown of SCDF by accident sequence type using fragility information from industry seismic PRAs, and (2) the CLERP for each accident type based on the current CPS internal events PRA model. The percent contribution of different accident types to SCDF was then multiplied by the internal events PRA CLERP to produce a sequence weighted average SCLERP. The LAR indicates that, in general, for a given plant, the SPRA does not necessarily produce the same distribution of accident scenarios as the internal events PRA. The LAR explains that containment isolation failures were evaluated to ensure that

random and seismically-induced failures leading directly to LERF are reasonably addressed in the estimation of the average SCLERP.

The LAR presents a graphical approach in Figure E4-1, Contribution of SCDF By Accident Sequence Type, for determining the contribution of different accident types to SCDF based on fragility information from cited industry seismic PRAs. The graphic resembles an event tree in which the top events and end states are used to define the accident types of interest and the branch point probabilities are used to calculate the percent contribution of each accident type to the total SCDF. The branch point probabilities appear to be determined by the seismic failure of components relevant to the branch points based on review of industry fragility information. The details involved in assigning a failure probability to each branch point in the graphic based on review of industry fragility information is not clear. Given this lack of clarity, address the following:

a) Regarding Figure E4-1, describe how the branch point probabilities were determined.

Discuss the information used from the cited industry seismic PRA reports, the basis for the selection of representative SSCs for developing the branch point probabilities and the basis for the applicability of that information to CPS. The discussion should provide the range of seismic capacities collected from industry seismic PRA reports and the basis to support the licensees selections.

b) Explain how the contribution of different scenarios and the average seismic CLERP developed for CPS compares to the seismic CLERP for plants most like CPS when the seismic CLERP is directly computed from the SCDF and SLERF for those plants.

c) Justify that the seismic LERF penalties provided in the submittal to support RICT calculations for CPS are conservative. The justification should include an explanation of the use of the internal events SCLERP values given that the internal events results only reflect random failures and do not capture the seismic failures of SSCs important for containment performance. The justification should also include a discussion of the impact of the estimate seismic penalties on the calculations for the proposed RICTs. If the approach to estimating seismic LERF cannot be justified as conservative for this application, then provide, with justification, the conservative SLERF penalties for use in RICT calculations.

DRA/APLC RAI 02 - External Hazard Screening Section 2.3.1, Configuration Risk Management Process & Application of Technical Specifications, Item 7, of NEI 06-09, Revision 0-A, states that the impact of other external events risk shall be addressed in the RMTS program, and explains that one method to do this is by documenting prior to the RMTS program that external events that are not modeled in the PRA are not significant contributors to configuration risk. The SE for NEI 06-09 (ADAMS Accession No. ML071200238) states that [o]ther external events are also treated quantitatively, unless it is demonstrated that these risk sources are insignificant contributors to configuration-specific risk.

LAR Enclosure 4, Section 6 provides an evaluation of other external hazards for screening in the RICT program and Section 7 concludes that no additional external hazards other than seismic events need to be added to the existing PRA model. The LAR also states that hazards are evaluated for plant configurations allowed under the RICT program. LAR Enclosure 4, Table E4-7, indicates that criterion C1 (event damage potential is less than events for which plant is designed) was used to screen the snow hazard. LAR Enclosure 4, Table E4-7, also indicates that criterion C1 was used to screen the sand or dust storm hazard. The LAR states that per the IPEEE, snow storm need not be considered per the guidance contained in NUREG 1407 and per the IPEEE, sand or dust storm need not be considered per the guidance contained in NUREG 1407, but does not explain why it does not need to be considered since NUREG-1407,

Procedural and Submittal Guidance for the Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities is IPEEE guidance and not specifically written for this application. It is unclear to the NRC staff that updated hazard information, since the IPEEE, has been considered in the screening analysis, and therefore, that the assumptions that resulted in the screening based on criterion C1 will continue to remain valid during the proposed RICTs.

LAR Enclosure 4, Table E4-7, indicates that criterion C4 (event is included in the definition of another event) was used to screen the ice cover hazard. It is unclear to the staff that the primary effect of potential accumulation of frozen water on a lake would be to cause a loss of offsite power.

Justify the screening of the following hazards and discuss how the assumptions that resulted in the screening will continue to remain valid during the proposed RICTs.

(1) snow hazard (2) sand or dust storm hazard (3) ice cover hazard.

PRA Licensing Branch C (APLC) 50.69 RAIs DRA/APLC RAI 03 - Alternate Seismic Approach Paragraph (b)(2)(ii) of 10 CFR 50.69 requires, for license amendment, a description of measures taken to assure the level of detail of the systematic processes that evaluate the plant.

This includes the internal events at power PRA required by 10 CFR 50.69(c)(1)(i) as well as the risk analyses used to address external events.

The proposed alternative seismic approach for Tier 1 plants is based on insights from EPRI 3002017583, which describes the risk insights derived from four case studies. These case studies compare the High Safety Significance (HSS) SSCs determined based on a seismic PRA (SPRA) against the HSS SSCs determined from other PRAs used for categorization. All four case studies included a full-power internal events (FPIE) PRA but only two included fire PRA information. Sections 3.3 through 3.5 of the EPRI report provide general information about the peer reviews conducted for the PRAs used in the case studies. However, the level of information provided is insufficient to determine whether the PRAs used in the case studies have been performed in a technically acceptable manner.

The staff has previously used additional supporting information to support its decision on the technical acceptability of the PRAs used in the case studies as well as details of the conduct of the case studies. This information is included in the supplements to the Calvert Cliffs Nuclear Power Plant, Units 1 and 2, LAR for the adoption of 10 CFR 50.69. The supplement dated July 1, 2019 (ADAMS Accession No. ML19183A012) clarified the alternate seismic approach (see response to RAI 4). The supplement dated July 19, 2019 (ADAMS Accession No. ML19200A216), supported the technical acceptability of the PRAs used for Plants A, C, and D as well as the technical adequacy of how the case studies were conducted; and included modifications to the content of the EPRI report. The supplement dated August 5, 2019 (ADAMS Accession No. ML19217A143), clarified a response in the July 19, 2019 supplement. The above-mentioned information is necessary for the staff to make its regulatory finding on the licensee's proposed alternative seismic approach and has not been provided by the licensee.

1. Provide an explanation of the differences between EPRI Report 3002017583 and EPRI Report 3002012988 which was used to support the staffs review of Calvert Cliffs 10 CFR 50.69 LAR wherein an alternative seismic approach was proposed that is similar to that proposed by

this licensee. The explanation should include justification for why the staff can rely on its review of EPRI Report 3002012988 and does not need to review EPRI Report 3002017583 separately.

2. Provide the above-mentioned information to support the staff's regulatory finding on the alternate seismic approach. This information can be provided by incorporating by reference the identified supplements, except any Calvert Cliffs specific information, as part of the licensees LAR or responding to the RAIs in the identified supplements.

3. Identify differences (if any) between the licensee's proposed alternative seismic approach and the alternative seismic approach previously approved by the staff (ADAMS Accession No. ML19330D909). Incorporate the differences in the licensees proposed approach or justify their exclusion.

DRA/APLC RAI 04 - External Hazard Screening NEI 00-04, Rev. 0, Section 5, Component Safety Significance Assessment states, If the plant does not have an external hazards PRA, then it is likely to have an external hazards screening evaluation that was performed to support the requirements of the IPEEE. NEI 00-04, Rev. 0, also states in Section 3.3.2, Other Risk Information (including other PRAs and screening methods), that the characterization of the adequacy of risk information should include a basis for why the other risk information adequately reflects the as-built, as-operated plant.

Attachments 4 and 5 of the LAR provide the screening results of other external hazards for 10 CFR 50.69 categorization and Section 3.2.4 concludes that all external hazards, except for seismic, were screened for applicability to CPS. Attachment 4 of the LAR indicates that criterion C1 (event damage potential is less than events for which plant is designed) was used to screen the snow hazard and the sand or dust storm hazard. The LAR states that per the IPEEE, snow storm need not be considered per the guidance contained in NUREG 1407 and per the IPEEE, sand or dust storm need not be considered per the guidance contained in NUREG 1407, but does not include the basis for why these evaluations adequately reflect the as-built, as-operated plant. It is unclear to the NRC staff that updated hazard information, since the IPEEE, has been considered in the screening analysis, and therefore, that the assumptions that resulted in the screening based on criterion C1 are appropriate for screening these external hazards from 10 CFR 50.69 categorization. to the LAR indicates that criterion C4 (event is included in the definition of another event) was used to screen the ice cover hazard. It is unclear to the staff that the primary effect of potential accumulation of frozen water on a lake would be to cause a loss of offsite power.

Address the following hazards and explain the basis for determining that the screening analysis adequately reflects the as-built, as-operated plant. Also, discuss how the assumptions that resulted in the screening will continue to remain valid during the implementation of 10 CFR 50.69.

(1) snow hazard (2) sand or dust storm hazard (3) ice cover hazard.

Electrical Engineering Branch By application dated April 30, 2020 (ADAMS Accession No. ML20121A178, Exelon Generation Company, LLC (EGC) (the licensee) requested a license amendment to revise the Clinton TS.

In particular, the amendment would permit the use of RICTs in accordance with Technical Specifications Task Force (TSTF)-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493).

Regulatory Requirement The regulations at 10 CFR 50.36, Technical specifications, establish the requirements related to the content of the TS. Pursuant to 10 CFR 50.36(c), TS are required to include items in five specific categories related to station operation: (1) Safety limits, limiting safety system settings, and limiting control settings, (2) LCOs, (3) Surveillance requirements (SR), (4) Design features; and (5) Administrative controls. The proposed changes in the LAR relate to the LCO and Administrative control categories.

DEX/EEOB RAI 01 According to TS 3.3.8.1, Action A.1, the current TS requires the licensee to place any inoperable channel relating to Loss of Voltage or Degraded Voltage Function in the trip position within one hour. According to Action A.2, the current TS requires the licensee to restore any channel relating to Degraded Voltage Function within 7 days. The licensee has proposed to extend the completion times as per the RICT program.

Page E1-11 of the LAR, for TS 3.3.8.1.A, states that the individual instrument channels for Loss of Power Instrumentation are not modeled. Therefore, a surrogate relay is chosen that fails the Diesel Generator (DG) start mode or undervoltage relay. Clarify whether the Inoperability of One or more channels is modeled as failure of the autostart mode of the associated DG, and that the manual start mode of DG is still considered available in the RICT model.

DEX/EEOB RAI 02 According to TS 3.3.8.1, Action A.1, the current TS requires the licensee to place any inoperable channel relating to Loss of Voltage or Degraded Voltage Function in the trip position within one hour. The RICT program would allow this completion time (CT) to be extended up to a maximum of 30 days. The associated NOTE states that the RICT program is Not applicable when trip capability is not maintained.

Explain the impact on the trip logic and trip capability of each function, if the inoperable channel is not placed in the trip position during the CT allowed by RICT program.

DEX/EEOB RAI 03 In the LAR, Enclosure 12, Risk Management Action (RMA) Examples, the licensee stated that multiple example RMAs may be considered during a RICT program entry to reduce the risk impact and ensure adequate defense-in-depth.

Provide a list of RMAs that are likely to be considered during the implementation of RICT programs relating to the following TS Conditions and Required Actions:

(a) TS 3.8.1, Condition C, Two offsite circuits inoperable, Action C.2.

(b) TS 3.8.4, Condition B One battery on Division 1 or 2 inoperable, Action B.1.

(c) TS 3.8.4, Condition C Division 1 or 2 DC electrical power subsystem inoperable for reasons other than Condition A or B, Action C.1.

(d) TS 3.8.7, Condition A Division 1 or 2 inverter inoperable, Action A.1 (e) TS 3.8.9, Condition A, One or more Division 1 or 2 AC electrical power distribution subsystems inoperable, Action A.1.

(f) TS 3.8.9, Condition B, One or more Division 1 or 2 uninterruptible AC bus distribution subsystems inoperable, Action B.1.

(g) TS 3.8.9, Condition C, One or more Division 1 or 2 DC electrical distribution subsystems inoperable, Action C.1.

Technical Specification Branch DSS/STSB RAI 01 The regulation under 10 CFR 50.36(c)(2) requires that TSs contain LCOs, which are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO of a nuclear reactor is not met, the licensee shall shut down the reactor or

follow any remedial action permitted by the TSs until the LCO can be met. Typically, the TSs require restoration of equipment in a timeframe commensurate with its safety significance, along with other engineering considerations. The regulation under 10 CFR 50.36(b) requires that TSs be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto.

In determining whether the proposed TS remedial actions should be granted, the Commission will apply the "reasonable assurance" standards of 10 CFR 50.40(a) and 50.57(a)(3). The regulation at 10 CFR 50.40(a) states that in determining whether to grant the licensing request, the Commission will be guided by, among other things, consideration about whether "the processes to be performed, the operating procedures, the facility and equipment, the use of the facility, and other TS, or the proposals, in regard to any of the foregoing collectively provide reasonable assurance that the applicant will comply with the regulations in this chapter, including the regulations in Part 20 of this chapter, and that the health and safety of the public will not be endangered."

TSTF-505, Revision 2 (ADAMS Accession No. ML18183A493) does not allow TS Conditions in the RICT program that represent a loss of a specified safety function or inoperability of all trains of a system required to be deemed OPERABLE. Throughout the TS markup pages in, the licensee proposed the following note for certain TS LCO Conditions to prohibit entering a RICT during a loss of function condition: "Not applicable if loss of function."

However, the proposed change to TS LCO 3.7.6 (Main Turbine Bypass System) Condition A (Requirements of LCO not met) appears to include a TS loss of function and has no such note. to the LAR, Table E1-1, indicates that the design basis success criteria for TS 3.7.6.A requires all six turbine bypass valves; this indicates that the loss of even one valve is a loss of function. Accordingly, provide technical justification that the proposed RICT for TS 3.7.6.A does not include a TS loss of function condition and address any discrepancies within LAR Table E1-1 for LCO 3.7.6.

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