ML20247J408

Apla Request for Additional Information LAR 50.69 and TSTF-505 (Epids L-2020-LLA-0017) and

ML20247J408
Person / Time
Site:	LaSalle
Issue date:	09/29/2020
From:	Bhalchandra Vaidya Plant Licensing Branch III
To:	Bryan Hanson Exelon Generation Co, Exelon Nuclear
Vaidya B, NRR/DORL/LPLIII, 415-3308
References
EPID L-2020-LLA-0017, EPID L-2020-LLA-0018
Download: ML20247J408 (15)

Text

September 29, 2020 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer (CNO)

Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

LASALLE COUNTY STATION, UNIT NOS. 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUESTS FOR AMENDMENTS TO RENEWED FACILITY OPERATING LICENSES TO ADOPT 10 CFR 50.69, "RISK-INFORMED CATEGORIZATION AND TREATMENT OF STRUCTURES, SYSTEMS, AND COMPONENTS FOR NUCLEAR POWER REACTORS" AND TO ADOPT TSTF-505, REVISION 2, "PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4B ((EPID L-2020-LLA-0017 AND EPID-L-2020-LLA-0018)

Dear Mr. Hanson:

By letters dated January 31, 2020, Exelon Generation Company, LLC (Exelon, the licensee) submitted 2 license amendment requests (LARs) to the U.S. Nuclear Regulatory Commission (NRC) for Renewed Facility Operating License Nos. NPF-11 and NPF-18 for La Salle County Station, Units 1 and 2 (La Salle):

(1) Request to modify (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML20031E699) the La Salle licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR)

Section 69 (50.69), Risk-Informed Categorization and Treatment of Structures, Systems and Components [SSCs] for Nuclear Power Reactors. The proposed changes are based on Nuclear Energy Institute (NEI) 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline, dated July 2005 (ADAMS Accession No. ML052910035).

The proposed changes are based on Nuclear Energy Institute (NEI) 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline, dated July 2005 (ADAMS Accession No. ML0713604560).

(2) Request to revise (Agencywide Documents Access and Management System (ADAMS)

Accession No. ML20035E577) the La Salle TS requirements to permit the use of risk-informed completion times for actions to be taken when limiting conditions for operation are not met. The proposed changes are based on Technical Specifications Task Force Traveler (TSTF-505, Revision 2, Provide Risk Informed Extended Completion Times -

RITSTF Initiative 4b, dated July 2, 2018 (ADAMS Package Accession No. ML18269A041).

The NRC staff has reviewed your submittal and conducted a Regulatory Audit on June 15 to June 19, 2020. The NRC staff has identified areas where additional information is needed to complete its review. The request for additional information (RAI) is enclosed.

On September 21, 2020, the NRC staff held a teleconference with Exelon staff to clarify this request to ensure that the licensee clearly understood the NRC staff concerns. By email dated September 24, 2020, from your staff, Mr. Jason Taken, you committed to provide the responses to this request no later than 30 calendar days from the date of this letter.

Please contact Bhalchandra K. Vaidya, Project Manager at 301-415-3308, or by email at Bhalchandra.vaidya@nrc.gov.

Sincerely,

/RA/

Bhalchandra K. Vaidya, Project Manager Plant Licensing Branch III Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-373 and 50-374

Enclosure:

Request for Additional Information cc: Listserv

Enclosure REQUEST FOR ADDITIONAL INFORMATION La Salle County Station, Units 1 and 2 License Amendment Requests to Adopt 10 CFR 50.69, "Risk-informed categorization and treatment of structures, systems, and components for nuclear power reactors" and to adopt TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times -

RITSTF Initiative4b, BACKGROUND:

By letters dated January 31, 2020, Exelon Generation Company, LLC (Exelon, the licensee) submitted two license amendment requests (LARs) to the U.S. Nuclear Regulatory Commission (NRC) for Renewed Facility Operating License Nos. NPF-11 and NPF-18 for La Salle County Station, Units 1 and 2 (LaSalle):

(1) Request to modify the LaSalle licensing basis to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations (10 CFR) Section 69 (50.69),

Risk-Informed Categorization and Treatment of Structures, Systems and Components

[SSCs] for Nuclear Power Reactors (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20031E699).

(2) Request to revise the LaSalle technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation (LCOs) are not met (ADAMS Accession No. ML20035E577).

The proposed changes are based on Technical Specifications Task Force (TSTF)

Traveler TSTF-505, Revision2, Provide Risk-Informed Extended Completion Times -

RITSTF Initiative4b, dated July 2, 2018 (ADAMS Accession No. ML18183A493).

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 SSC Categorization Guideline, Revision 0, as endorsed by Regulatory Guide (RG) 1.201, Guidelines for Categorizing Structures, Systems, and Components in Nuclear Power Plants According to Their Safety Significance. 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, and implement the RICT program in accordance with TSTF-505, Revision 2.

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 [PRA] 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 LWRs. 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 in the 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 Decision-making, 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 LaSalle for consistency with identified RGs and applicable industry guidances.

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

[Applicable for TSTF-505 and 10 CFR 50.69]

LAR Enclosure 2, Table E2-2, presents the dispositions for three facts and observations (F&Os) that remain open after the Independent Assessment (IA) performed for closure of F&Os; two that remain open (i.e., 1-19, 4-17) and one partially resolved (i.e., 6-11). For F&Os 4-17 and 6-11, the licensee states the items will be resolved prior to TSTF-505 implementation, however, of the LAR that includes a table of implementation items to be completed prior to implementation of the RICT program does not include these items. In light of these observations provide the following:

a) Regarding F&O 4-17, the LARs dispositions state that [t]his item will be resolved prior to TSTF-505 implementation [and 10 CFR 50.69 implementation] for each respective LAR. Furthermore, the licensee states that the impact of this issue is

judged to be minimal. However, it is not clear to NRC staff the impact on the RICT calculations. Therefore, address the following:

i.

Provide justification (e.g., description and results of a sensitivity study) that any needed adjustments made to the fire PRA (FPRA) identified from review of the plant-specific data on the fire suppression and detection systems cannot impact the RICT calculations performed to support TSTF-505 or does not adversely impact the importance measures and risk metrics used to assess structures, systems, and components (SSC) categorization for 10 CFR 50.69.

ii.

Alternatively, propose a mechanism that ensures the review of plant-specific data for fire suppression and detection systems is performed and any update needed to the FPRA is completed prior to implementation of the RICT program and 10 CFR 50.69 (e.g., include as an implementation item in the respective LARs).

b) Regarding F&O 6-11, the LARs dispositions discuss that a review will be performed to verify consistency with NEI 00-01, Revision 3, prior to implementation of the RICT program [and 10 CFR 50.69 implementation] for each respective LAR. However, no commitment to complete an implementation item for this F&O is made in either LAR. Also, the NRC staff notes that in the event the review cannot verify the circuit analysis was performed in accordance with the requirements of NEI-00-01, Revision 3, then adjustments to the FPRA model may be needed.

i.

Provide sufficient justification to support the conclusion provided in Table E2-2 that any revisions to the cable selection based upon historical methods used during the time of the analysis development and more recent guidance (i.e., NEI 00-01, Revision 3) has no impact on the TSTF-505 RICT calculations performed to support TSTF-505 or does not adversely impact the importance measures and risk metrics used to assess SSC categorization for 10 CFR 50.69.

ii.

Alternatively, propose a mechanism that ensures the review of the circuit analysis to the requirements of NEI-00-01, Revision 3, is performed and any needed update to the FPRA is completed prior to implementation of the RICT program and 10 CFR 50.69 (e.g., include as an implementation item in the respective LARs).

DRA/APLA(B) RAI 02-Fire Hazards

[Applicable for TSTF-505 and 10 CFR 50.69]

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. Some concerns are not always readily 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 overall risk profile of the plant. The NRC staff notes that the calculated results of the PRAs are used directly to calculate a RICT and evaluate the categorization of SSCs. Specifically, for the TSTF-505 application, the PRA results are relied upon considerably to determine how long SSCs

controlled by a TS can remain inoperable; therefore, the NRC staff requests additional information for the following areas.

TSTF-505 LAR, Enclosure 9, Section 4, and Attachment 6 of the 10 CFR 50.69 LAR discusses that the LaSalle FPRA was guided using consensus methods outlined in NUREG/CR-6850 and interpretations of technical approaches as required by NRC. The licensee further states in both of the LARs that [f]ire PRA methods were based on, other more recent NUREGs, (e.g.,

NUREG-7150 []), and published frequently asked questions (FAQs) for the Fire PRA. Furthermore for the TSTF-505 LAR, Part (e), of the proposed TS 5.5.17 ("Risk Informed Completion Time Program) states, in part, [m]ethods to assess the risk from extending the Completion Times must be PRA methods used to support this license amendment, or other methods approved by the NRC for generic use; and any change in the PRA methods to assess risk that are outside these approval boundaries require prior NRC approval.

The integration of NRC-accepted FPRA methods and studies described below that are relevant to these submittals could potentially impact the TSTF-505 RICT calculations, categorization of SSCs, and risk metrics for total core damage frequency (CDF) and total large early release frequency (LERF):

NUREG-2178, Volume 1, Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), dated April 2016 (ADAMS Accession No. ML16110A140).

NUREG-2180, Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities (DELORES-VEWFIRE), dated December 2016 (ADAMS Accession No. ML16343A058).

a) There have been changes to the FPRA methodology since the last full-scope peer review of the LaSalle FPRA in 2015. For each of the above NRC-accepted FPRA methods and studies, the NRC staff requests the licensee address one of the following for parts (i) through (iii) of this RAI:

i.

Discuss how the FPRA method/study had been incorporated into the LaSalle FPRA and, as applicable, summarize the changes made to the FPRA model. Indicate whether this change was PRA maintenance or a PRA upgrade as defined in ASME/ANS RA-Sa-2009, Section 1-5.4, as qualified by RG 1.200, Revision 2, along with a justification for the determination. If this change constitutes a PRA upgrade, discuss the focused-scope (or full-scope) peer review(s) that has been performed to evaluate the change, and provide any open F&Os and associated dispositions from this peer review(s) in accordance with RG 1.200, Revision 2.

OR ii. If the FPRA method/study has not been incorporated into the LaSalle FPRA, provide a detailed justification for why the integration of the FPRA method/study would not change the conclusions of the LARs, and subsequently not impact the TSTF-505 RICT calculations, categorization of SSCs, and risk metrics for total CDF and total LERF. As part of this justification, identify any FPRA methodologies used in the LaSalle FPRA that are no longer accepted by the NRC staff (e.g., guidance provided in FAQ 08-0046, Closure of National Fire Protection Association 805 Frequently Asked Question 08-0046 Incipient Fire Detection Systems, ADAMS Accession No.

ML093220426, has been retired by letter dated July 1, 2016, (ADAMS Accession No. ML16167A444). Provide technical justification for its use in TSTF-505 RICT calculations and the categorization of SSCs and evaluate the significance of its use on the risk metrics provided in Enclosure 5 of the TSTF-505 LAR and Attachment 2 of the 50.69 LAR.

OR iii. Propose a mechanism that ensures the FPRA method/study (or other NRC acceptable method) will be integrated into the LaSalle FPRA prior to implementation of the RICT program and 10 CFR 50.69. If this FPRA update is determined to be a PRA model upgrade per the ASME/ANS PRA standard, include in this mechanism a process for conducting a focused-scope peer review and ensure any findings are closed by using an approved NRC process.

b) Application of Minimum Joint Human Error Probability (HEP) Values in the Fire PRA Section 6.2 of NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines (ADAMS Accession No. ML12216A104) cites NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), which advises that minimum joint human error probabilities (JHEP), not be below ~1.0E-05 since it is typically hard to defend other independent failure modes that are not usually treated NUREG-1921 also states in part, while it might be reasonable to adopt some sort of limit, it needs to be done carefully, so that the results of PRAs are not distorted by arbitrary assignments of probabilities. As discussed in detail later on, any limiting values should be consistent within the context of the scenarios in which they are applied. Furthermore, Table 4-4 of Electrical Power Research Institute (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. For the NRC staff to assess the HRA and application of JHEPs in the FPRA model for technical acceptability consistent with RG 1.200, Revision 2, provide the following:

i.

If a minimum JHEP value less than 1E-05 was used in the FPRA, then provide sufficient justification (e.g., results of a sensitivity study, basis for inapplicability of NUREG-1792, etc.) to confirm that the minimum JHEP value is reasonable and sufficient for the assigned human failure events (HFEs).

ii.

Alternatively, propose a mechanism that ensures the JHEP values used in the FPRA are consistent with the guidance in NUREG-1792 for JHEP floor values prior to implementation of the risk-informed applications.

c) Well-Sealed Motor Control Center (MCC) Cabinets 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 of NUREG/CR-6850, 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 [motor control center] Electrical Panels Greater than 440V (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.

i.

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

ii.

If well-sealed cabinets less than 440 V are included in the Bin 15 count of ignition sources, provide justification for using this approach. Justification should be sufficient in detail to ascertain that there is no adverse impact to the FRPA model used for performing RICT calculations and the categorization of SSCs.

d) PRA Treatment of Fire Dependencies Between Units 1 and 2 Many plants have Units 1 and 2 adjoined and, thus, have common areas. For these plants, the risk contribution from fires originating in one unit must be addressed for impacts to the other unit given the physical proximity of the other unit and common areas. Therefore, address the following for Units 1 and 2 adjacent and common areas.

i.

Discuss how the risk contribution of fires originating in one unit is addressed for the other unit given impacts due to the physical proximity of equipment and cables. Include identification of locations where a fire in one unit can affect components in the other unit and explain how the risk contributions of such scenarios are allocated for a RICT calculation, the categorization of SSCs, and RG 1.174 risk thresholds.

DRA/APLA RAI 03-PRA Configuration and Control

[Applicable for TSTF-505 and 10 CFR 50.69]

Section 2.3.4 of NEI 06-09, Revision 0-A, 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.

TSTF-505 LAR, Enclosure 7, states that if a plant change or a discovered condition is identified and can have significant impact to the RICT program calculations, then an unscheduled update of the PRA models will be implemented. More specifically, the LAR states that if the plant changes meet specific criteria defined in the plant PRA and update procedures, including criteria associated with consideration of the cumulative risk impact, then the change will be incorporated into applicable PRA models without waiting for the next periodic PRA update.

Neither the TSTF-505 or 10 CFR 50.69 LARs explains under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedures that will be used to initiate the update. Therefore, describe the conditions under which an unscheduled PRA update (i.e., more than once every two refueling cycles) would be performed

and the criteria that would be used to require a PRA update. In the response, describe what is meant by significant impact to the RICT program calculations or SSC categorization results.

DRA/APLA RAI 04 -System and Surrogate Modeling Used in the PRA Models

[Applicable for TSTF-505]

The NRC safety evaluation (SE) to NEI 06-09, Revision 0, 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 to 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 some TS LCO conditions, the table explains that the associated SSCs are not modelled in the PRAs but will be conservatively represented using a surrogate event. For certain LCOs it is unclear to the NRC staff whether the modeling (i.e., depth, surrogates used) will be acceptable to address the associated RICTs calculated. Therefore, address the following:

a) For TS LCO 3.3.6.1 (Primary Containment Isolation Instrumentation), Condition A, (Primary containment instrumentation - one or more channels inoperable) in LAR Table E1-1, the logic for primary containment isolation is not modeled in detail and, therefore, a surrogate event will be used. The table states that the surrogate event will either be failure of containment or failure of the frontline system. It is not clear to NRC staff what system is being referred to by the phrase frontline system. Confirm/identify the frontline system(s) intended to be referred to in the table (e.g., Primary Containment Isolation valves (PCIVs)) and discuss how the failure of the function(s) compares to failure of containment.

b) For TS LCO 3.6.1.3, PCIV, Condition A (One or more penetration flow paths with one PCIV inoperable for reasons other than Condition D), in LAR Table E1-1, it states not all PCIVs are modeled, therefore, a surrogate of a pre-existing containment failure is chosen. It is not clear to the NRC staff what pre-existing containment failure will be chosen and how LaSalle will ensure that it represents the equivalent release as an open PCIV. Therefore, discuss the pre-existing containment failure or criteria used to determine which surrogate will be identified for an inoperable PCIV and provide sufficient justification to ensure that the surrogate will reflect the function covered by the TS LCO condition. In the discussion include whether this failure assumes a large early release.

APLA RAI 05 - PRA Model Uncertainty Analysis

[Applicable for TSTF-505 and 10 CFR 50.69]

The NRC SE to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and key sources of uncertainty and assess/disposition each as to their impact on the risk managed technical specifications (RMTS) application. LAR Enclosure 9, Tables E9-1, E9-2, and E9-3 provided the key assumptions and sources of uncertainty identified for the internal events PRA (IEPRA), transition to the Real Time Risk (RTR) model, and the FPRA and included the dispositions for each source of uncertainty for this TSTF-505 application. Furthermore, as part of its audit of the LAR, the NRC staff reviewed the PRA analyses that support the uncertainty analysis for the LaSalle IEPRA, FPRA, and risk applications cited in the LAR. Upon review of the dispositions provided in LAR Tables E9-1, E9-2 and E9-3, it was unclear to the

NRC staff how the licensee concluded there was no impact on the RICT calculations. In light of these observations, address the following:

a) For the TSTF-505 and 10 CFR 50.69 LARs, the licensee identifies cable selection as a key source of FPRA modeling uncertainty because of conservatisms in the approach (i.e., lack of cable data). The LARs discuss that an informed approach was used in developing the assumed [cable] routing and that other modeling assumptions provide some offsetting effects; therefore, the licensee concludes that Unknown Location (UNL) modeling uncertainty has no impact on the RICT program calculations and SSC categorization. It is not clear to the NRC staff how this assumption (i.e., cable selection) was concluded to have no impact on the risk-informed applications, given a sensitivity study demonstrated that its impact on risk is moderate. Therefore, provide a description and results of the selective sample of sensitivities performed to demonstrate that the conservatisms are bounding and do not adversely impact the results of future SSC categorization and future RICT calculations.

b) For the TSTF-505 and 10 CFR 50.69 LARs, the licensee identifies vapor suppression capability following vessel rupture failure as a key source of uncertainty. The TSTF-505 LAR dispositions this uncertainty by stating, [a]lthough the RICT estimates change as a result of this sensitivity, the bounding sensitivity analysis utilizes the upper bound values, which is not a realistic assumption and use of this bounding assumption would result in overly conservative RICT estimates. The 10 CFR 50.69 LAR dispositions this uncertainty by stating, [t]herefore, the uncertainty associated with this model uncertainty is negligible within the 50.69 application. It appears to the NRC staff that the results of the sensitivity study using upper bound values from NUREG/CR-6595 validate the concern that this source of uncertainty impacts the RICT calculations and potentially the SSC categorization results. In light of the staff concerns, address the following:

i.

Given the uncertainty indicated by recent MAAP [modular accident analysis program] runs cited in the LAR and the results of the sensitivity study on vapor suppression failure following vessel failure, provide sufficient justification to support the conclusion that the upper bound values used are not realistic and would result in overly-conservative RICT estimates and SSC categorization results.

c) The LaSalle FPRA uncertainty for the TSTF-505 LAR and 10 CFR 50.69 does not discuss the uncertainty associated with modeling the Hardened Containment Vent System (HCVS) in the FPRA, nor does the LAR discuss how this uncertainty will be treated in the RICT program. In light of the NRC staff concerns, address the following:

i. Discuss the uncertainty associated with modelling the HCVS in the FPRA along with justification that the modelling of the HCVS is sufficient for the TSTF-505 and 10 CFR 50.69 applications given the impact that credit for HCVS has on the fire CDF.

ii. If in the response to part (i) above, the modelling of the HCVS in the FPRA cannot be justified for this application given the significant impact that HCVS modeling credit has on the fire CDF, propose a mechanism that addresses this source of modelling uncertainty in the RICT and SSC categorization programs.

d) For the TSTF-505 LAR, Enclosure 9, Table E9-1, identifies credit for survivability of the emergency core cooling system (ECCS) after containment venting as a key source of uncertainty. It is not clear to the NRC staff how this key source of uncertainty was concluded to have negligible impact on RICT calculations and no further treatment (i.e.,

risk management actions, RMAs) is needed. The NRC staff acknowledges that increasing the conditional failure probability of the ECCS to 1.0 is bounding. Provide the following additional information:

i.

Discuss any applicable RMAs for the TS LCOs impacted by this uncertainty (e.g.,

TS LCO Conditions 3.3.5.1.C, 3.3.5.1.D, 3.5.1.A, and 3.5.1.C). Include justification for why the RMAs are sufficient to address this uncertainty, specifically discuss why additional RMAs or other measures are not needed to reduce the risk of the impacted TS LCOs.

ii. If in response to part (i) above, it cannot be justified that the applicable RMAs are sufficient to address this uncertainty, then include additional RMAs or other measures that will be used to reduce the risk impact of the cited uncertainty for the applicable TS LCOs (e.g., include as an implementation item in Attachment 5 of the LAR).

APLA RAI 06 - Evaluating SOKC Uncertainty Impact on the Risk Applications

[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. In light of these observations, address the following:

a) Provide a summary of how the SOKC investigation was performed for the base LaSalle 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.

DRA/APLA RAI 07: Key Principle 5 - 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 Regulatory Guide 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 LaSalle 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 LaSalle 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 RAI 08 -Credit of FLEX in the PRA Models

[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 was 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 09 - Implementation Items

[Applicable for TSTF-505 and 10 CFR 50.69]

RG 1.174, Revision 3, and RG 1.200, Revision 2, 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 change and the role the PRA plays in justifying the change. If the responses to any RAIs require any follow-up actions prior to implementation of the risk-informed applications (i.e., 10 CFR 50.69 or TSTF-505), provide a list of those actions and any PRA modeling changes including any items that will not be completed prior to issuing the amendments but must be completed prior to implementing the applications.

Propose a mechanism that ensures these activities and changes will be completed and appropriately reviewed and any issues resolved prior to implementing the applications (for example, a license condition that includes all applicable implementation items and a statement that they will be completed prior to implementation).

ML20247J408

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