ML22061A004

Audit Questions for License Amendment Request to Revise Technical Specifications to Adopt TSTF-505 and TSTF-439, Revision 2

ML22061A004
Person / Time
Site:	Callaway
Issue date:	03/07/2022
From:	Mahesh Chawla Plant Licensing Branch IV
To:	Diya F Union Electric Co
Chawla M
References
EPID L-2021-LLA-0197
Download: ML22061A004 (19)

Text

March 7, 2022 Mr. Fadi Diya Senior Vice President and Chief Nuclear Officer Ameren Missouri Callaway Energy Center 8315 County Road 459 Fulton, MO 65077

SUBJECT:

CALLAWAY PLANT, UNIT NO. 1 - AUDIT QUESTIONS FOR LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505 AND TSTF-439, REVISION 2 (EPID L-2021-LLA-0197)

Dear Mr. Diya:

By application dated October 21, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21294A393), as supplemented by letter dated November 24, 2021 (ADAMS Accession No. ML21328A182), Union Electric Company, dba Ameren Missouri (the licensee), submitted a license amendment request (LAR) for Callaway Plant, Unit No. 1 (Callaway). The proposed amendment would modify the Callaway Technical Specifications (TSs) to implement risk-informed completion times and the Risk-Informed Completion Time Program in accordance with Technical Specifications Task Force (TSTF)

Traveler TSTF-505, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b, Revision 2. In support of the adoption of TSTF-505, the licensee also would adopt TSTF-439, Revision 2, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO [Limiting Condition for Operation], which involves the elimination of second completion times currently specified in the TSs. In addition, the proposed amendment would remove obsolete one-time completion times contained in the Callaway TSs.

On February 18, 2022 (ADAMS Accession No. ML22048A002), the U.S. Nuclear Regulatory Commission (NRC) staff issued an audit plan that conveyed intent to conduct a regulatory audit to support its review of the subject license amendment. In the audit plan the NRC staff requested an electronic portal setup and provided a list of documents to be added to the portal.

The NRC staff has performed an initial review of these documents and developed a list of audit questions. The proposed date for the audit is from March 22, 2022, through March 25, 2022.

The list of audit questions is enclosed in this letter. The proposed agenda for the audit will be provided at a later date.

If you have any questions, please contact me at 301-415-8371 or via e-mail at Mahesh.Chawla@nrc.gov.

Sincerely,

/RA/

Mahesh L. Chawla, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-483

Enclosure:

Audit Questions cc: Listserv

Enclosure AUDIT QUESTIONS REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2 AND TSTF-439, REVISION 2 UNION ELECTRIC COMPANY DBA AMEREN MISSOURI CALLAWAY PLANT, UNIT NO. 1 DOCKET NO. 50-483 By application dated October 21, 2021 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML21294A393), as supplemented by letter dated November 24, 2021 (ADAMS Accession No. ML21328A182), Union Electric Company, dba Ameren Missouri (the licensee), submitted a license amendment request (LAR) for Callaway Plant, Unit No. 1 (Callaway).

The amendment would revise 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. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision2, Provide Risk-Informed Extended Completion Times -

RITSTF [Risk-Informed TSTF] Initiative4b, dated July 2, 2018 (ADAMS Accession No. ML18183A493).

In support of the adoption of TSTF-505, the licensee also would adopt TSTF-439, Revision 2, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO, which involves the elimination of second completion times currently specified in the TSs. In addition, the proposed amendment would remove obsolete one-time completion times contained in the Callaway TSs. The U.S. Nuclear Regulatory Commission (NRC) issued a final model safety evaluation (SE) approving TSTF 505, Revision 2, on November 21, 2018 (ADAMS Accession No. ML18269A041).

The NRC staff from the Office of Nuclear Reactor Regulation; Division of Engineering and External Hazards (DEX), Division of Risk Assessment (DRA), and Division of Safety Systems (DSS) have identified the following information needed to further understand the licensees proposed adoption of TSTF-505.

DSS, Technical Specifications Branch (STSB)

STSB QUESTION 01 NRC staff observes that TSs 3.7.4.A.1 and 3.7.4.B.1 have notes added to the TSs which state, Not applicable when more than 2 ASD [Atmospheric Steam Dump Valves] lines are inoperable for any reason, and were not listed as variations. These are variations from TSTF-505 and need to be identified in the variation section of the LAR.

STSB QUESTION 02 The NRC staff identified the following editorial issues in Attachment 4 (bases):

For TS 3.3.1, the licensee removed Power Range Neutron Flux - Low from Action E.

Based on Table 3.3.1-1, staff believes this was inadvertently included in the bases and is being removed as a cleanup item. Confirm.

For TS 3.3.1, Action L.1, there is a. missing between L and 1 in the Section title. In addition, staff notes that the last phrase should say below P-7 setpoint instead of below P-7. Please confirm.

For TS 3.3.1, Action R.1, there is a period and space that were inadvertently deleted and needs to be corrected. Please confirm.

For TS 3.3.1, Actions W.1, W.2.1, W.2.2.1, W.2.2.2, and W.2.3, staff notes that in the last paragraph, the licensee inadvertently forgot to delete the V from the three references to W actions.

For TS 3.3.2, Actions C.1 and C.2, staff notes that and is missing and a, needs to be removed between the two actions for the Section titles in three places that they are listed.

For TS 3.3.2, Action I.1, staff notes that and I.2 needs deleted from the Section title, as it was removed from the TS.

For TS 3.3.2, Action R.1, staff notes that, R.2.1, and R.2.2 needs to be deleted from the Section title, as it was removed from the TS.

For TS 3.3.2, Action S.1, staff notes that, S.2.1, and S.2.2 needs to be deleted from the Section title in two places, as it was removed from the TS.

For TS 3.3.2, Action T.1, staff requests whether the last sentence should state In MODE 5 instead of In MODE 4.

For TS 3.4.9, staff notes that the bases were provided with markups to include a RICT; however, the application indicates that this TS was not included in the RICT Program.

Please confirm and correct.

STSB QUESTION 03 TSTF-505, Revision 2, does not allow for TS loss of function (LOF) conditions (i.e., those conditions that represent a loss of a specified safety function or inoperability of all required trains of a system required to be operable) in the RICT Program.

Based on the design success criteria provided in the LAR Enclosure 1, List of Revised Required Actions to Corresponding PRA Functions, Table E1-1, In Scope TS/LCO Conditions to Corresponding PRA [Probabilistic Risk Assessment] Functions, it appears that some LCO Actions may constitute an LOF.

a.

LAR Enclosure 1, Table E1-1, lists in the column of TS LCO Action 3.3.1.U, a condition with One trip mechanism inoperable for one RTB [reactor trip breaker]. The corresponding column of the SSCs [Structures, Systems, and Components] Addressed by TS LCO Condition states, RTB Undervoltage and Shunt Trip Mechanisms, and the column of Design Success Criteria states, One trip mechanism. Table 3.3.1-1 indicates that for Condition U, one trip mechanism is required per RTB. Please explain a condition where there would not be an LOF.

b.

LAR Enclosure 1, Table E1-1 lists in the column of TS LCO Action 3.5.2.A, a condition with One or more trains inoperable AND at least 100% of the ECCS [emergency core cooling system] flow equivalent to a single OPERABLE ECCS train available. The corresponding column of the SSCs Addressed by TS LCO Condition states. Two ECCS trains (ECCS trains consists of one Centrifugal Charging, one Safety Injection, and one Residual Heat Removal subsystem.), and the column of Design Success Criteria, states, 3 ECCS subsystems between two trains such that 100% ECCS flow equivalent to a single operable ECCS train is available. If more than one train is inoperable, this appears to be an LOF. Please explain why it is not, and if it is, please add the note not applicable when an LOF occurs, consistent with TSTF-505.

c.

LAR Enclosure 1, Table E1-1 lists in the column of TS LCO Action 3.8.1.C, a condition with Two offsite circuits inoperable. The corresponding column of the SSCs Addressed by TS LCO Condition states, Two qualified circuits between the offsite transmission network and the onsite 1E AC [alternating current] electrical power distribution system, and the column of Design Success Criteria states. One qualified circuit between the offsite transmission network and the onsite 1E AC electrical power distribution system. If two offsite circuits are inoperable, this appears to be an LOF.

Please explain why it is not and if it is, please remove it from the RICT Program.

STSB QUESTION 04 NRC staff suggestion for licensee consideration: The proposed administrative controls for the RICT Program in TS 5.5.19 paragraph e of Attachment 2 to the LAR was based on the TS markups of TSTF-505, Revision 2 for Callaway. The NRC staff recognizes that the model SE for TSTF-505, Revision 2 contains improved phrasing for the administrative controls for the RICT Program in TS 5.5.19 paragraph e, namely the phrasing approved for use with this program instead of used to support this license amendment. In lieu of the original phrasing in TS 5.5.19 paragraph e, discuss whether the phrases used to support Amendment # xxx or, as discussed in the TSTF-505 model SE, approved for use with this program would provide more clarity for this paragraph.

DSS, Containment and Plant Systems Branch (SCPB)

SCPB QUESTION 01

a.

LAR Enclosure 1, Table E1-1 lists in the column of TS LCO Action 3.6.2.C, a condition with one or more containment air locks inoperable for reasons other than Condition A or B. The corresponding column of the SSCs Addressed by TS LCO Condition states, Containment Airlocks, and the column of Design Success Criteria states, One of two containment air lock doors closed. If more than one containment air lock is inoperable, this appears to be an LOF. Please explain why it is not, and if it is, please add the note not applicable when a LOF occurs, consistent with TSTF-505.

b.

LAR Enclosure 1, Table E1-1 lists in the column of TS LCO Action 3.7.2.B, a condition with Two MSIV [main steam isolation valves] actuator trains inoperable for different MSIVs when the inoperable actuator trains are not in the same separation group. The corresponding column of the SSCs Addressed by TS LCO Condition states, Main Steam Isolation Valves (MSIVs), and the column of Design Success Criteria states, One MSIV closure per steam generator (one of two actuator trains). If two MSIV actuator trains are inoperable (as indicated for Condition B), this appears to be an LOF.

Please explain why it is not, and if it is, please remove it from the RICT Program.

DEX, Electrical Instrumentation and Controls Branch (EICB)

EICB QUESTION 01 Within LAR Enclosure 1, starting on page E1-30, the licensee provides a general description of the redundancy design features for the facility. The NRC staff needs the diversity and redundancy features with respect to each event in the accident analysis to evaluate that the defense-in-depth for the proposed changes is preserved. Please demonstrate the diverse means for each event in Callaway Final Safety Analysis Report (FSAR) Chapter 15, Accident Analysis (ADAMS Accession No. ML21193A202), that is relevant to the proposed changes.

EICB QUESTION 02 Under TS 3.3.2, Condition B, please describe the individual design success criteria for each of the functions listed below:

FUNCTION 1.a FUNCTION 2.a FUNCTION 3.a.(1)

FUNCTION 3.b.(1) and how under Condition B these functions preserve their manual initiation capabilities.

DSS, Nuclear Systems Performance Branch (SNSB)

SNSB QUESTION 01:

TS 3.7.4 - Atmospheric Steam Dump Valves (ASDs)

The design basis of the safety function for the ASDs is established by the capacity to cool the unit to the residual heat removal (RHR) entry conditions for various events such as a steam generator tube rupture (SGTR). Callaway Final Safety Analysis Report (FSAR),

Section 15.6.3.1.2, Analysis of Effects and Consequences (ADAMS Accession No. ML21193A202), discusses the analysis of the SGTR showing that an LOF of the ASDs would not occur. Page 15.6-17 of the Callaway FSAR states, in part, for the dose release limiting case that:

Two of the unaffected steam generators are assumed to continually discharge steam and entrained activity via the atmospheric steam dump valves up to the time initiation of the RHR system can be accomplished.

The licensee proposed to apply the RICT Program to TS 3.7.4 Condition B that allows two required ASD lines inoperable for reasons other than excessive ASD seat leakage. Since the TS LCO 3.7.4 requires four ASD lines to be operable, Condition B would result in only two required ASD lines operable. If an SGTR case (designated as an SGTR with one operable ASD line, thereafter) occurs in the steam generator (SG) with an operable ASD line, there would be one operable ASD line remaining in the unaffected SGs for plant cooldown, which is less than two ASD lines assumed in the SGTR analysis. The effect of using more operable ASD lines in the unaffected SGs for plant cooldown decreases the time it takes to reduce the reactor coolant system temperature to below the ruptured SG saturated temperature and decreases the dose releases. Therefore, an SGTR with one operable ASD line (allowed by Condition B) would not be bounded by the limiting case in Callaway FSAR, Section 15.6.3.1.2.

The NRC-approved Nuclear Energy Institute (NEI) 06-09, Revision 0-A Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS)

Guidelines (ADAMS Accession No. ML12286A322) provides guidelines for application of the RICT Program. Specifically, Condition 3 in the NRC SE approving NEI 06-09 imposes a restriction stating that when an LOF of specific safety function for the affected TS system occurs, the RICT Program cannot be applied.

Provide analysis of the SGTR with one operable ASD line discussed above to show that the two required operable ASD lines would not result in an LOF for mitigating the SGTR event, when applying the RICT Program to TS 3.7.4 Condition B.

DRA, Probabilistic Risk Assessment (PRA) Licensing Branch A (APLA)

APLA(C) QUESTION 01 - PRA Model Uncertainty Analysis Process LAR Enclosure 9, Key Assumptions and Sources of Uncertainty, specifically discusses how plant specific sources of uncertainty were considered from the PRA notebooks for the internal events (IE), fire, seismic and high wind (HW) PRAs and how generic sources of uncertainty were considered from Electric Power Research Institute (EPRI) Topical Report (TR) 1016737, "Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, for the IE PRA and from EPRI TR 1026511, "Practical Guidance of the Use of Probabilistic Risk Assessment in Risk-informed Applications with a Focus on the Treatment of Uncertainty, for the Level 2 and fire PRA. The LAR does not indicate that generic sources of seismic PRA uncertainty were considered though such a list exists In EPRI TR 1026511, Appendix C.

The Callaway Title 10 of the Code of Federal Regulations (10 CFR) Section 50.69, Risk-informed categorization and treatment of structures, systems and components for nuclear power reactors, supplement dated July 29, 2021 (ADAMS Package Accession No. ML21210A025),

confirmed that the following five criteria were used to screen and identify assumptions and sources of uncertainty (including modeling choices and approximations): (1) based on a current industry consensus modeling approach, (2) applied the most recent industry data, (3) no impact on the PRA results and, therefore, no impact on the application, (4) the guidance in NEI 00-04, Revision 0, 10 CFR 50.69 SSC Categorization Guideline, already requires a sensitivity study on human failure events (HFEs), and (5) a sensitivity was performed on the base model demonstrating the source of uncertainty has no impact on the risk results and, therefore, no impact on the application. Though not explicitly stated in the TSTF-505 LAR, these same criteria would appear to also apply to the uncertainty analysis used to perform a RICT with the exception of No. 4.

In consideration of the observations above, address the following:

a.

Clarify whether the generic seismic PRA assumptions and sources of uncertainty listed in the EPRI TR 1026511 Appendix were considered in the uncertainty analysis performed for the RICT application. If they were not considered, provide justification for why the exclusion of considering the generic seismic probabilistic risk analysis assumptions and sources of uncertainty is not applicable or adverse to the probabilistic risk analysis used for the RICT Program.

b.

Confirm whether the screening criteria used to identify key sources of modeling uncertainty for the RICT application were the same used in the 10 CFR 50.69 application (with the exception noted above). Include a brief description of any screening criteria that are different from the criteria used for the uncertainty analysis performed for the Callaway 10 CFR 50.69 application.

APLA(C) QUESTION 02 - Dispositions Model Assumptions and Sources of Uncertainty NRC staff reviewed the dispositions of the sources of modeling uncertainty provided in LAR,, Tables E9-1, E9-2, E9-3, E9-4, and E9-5 along with the licensees assessment of the modeling uncertainties impact on the RICT application. In some instances, there is not enough information for NRC staff to conclude that the assumption or source of modeling uncertainty would have no adverse impact on the RICT calculations.

a.

Concerning the IE PRA, portal report PRA-IE-UNCERT, Revision 1, Probabilistic Risk Assessment (PRA), Uncertainty Analysis Notebook identifies treatment of temperature-induced (TI) and pressure induced (PI) SGTRs as a source of modeling uncertainty.

The report explains that the TI-SGTR and PI-SGTR probabilities are developed based on the assumption that the SG tubes are of pristine quality. The report explains that the results of a sensitivity study performed based on average SG tube conditions using guidance from WCAP-16341-P, Simplified Level 2 Modeling Guidelines, shows a 36.8 percent increase in large early release frequency (LERF) from the base case. The report states that plant conditions are monitored with respect to this underlying assumption, but also states that this is a potentially significant source of modeling uncertainty for LERF. Given this observation, the age of the plant, and the potential impact of this treatment on the RICT calculations, it is not clear why average rather than pristine SG tube conditions are assumed. Therefore, address the following:

i.

Provide the basis for the modeling assumption that the Callaway SG tubes condition are pristine rather than average, along with justification that confirms no adverse impact to the RICT Program.

OR ii.

If response to part i. above is that it cannot be supported, provide justification to demonstrate (e.g., sensitivity study) that the modeling treatment of potential TI-SGTR and PI-SGTR events have no adverse impact on the RICT Program. The sensitivity study should at minimum consider a set of LCOs identified to the most affected by the modeling uncertainty of PI and TI SGTR and an estimated RICT less than the backstop of 30 days.

OR iii.

As an alternative to part i. and ii. above, identify the LCO conditions impacted by the treatment of this modeling uncertainty, along with potential risk management actions (RMAs) that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

b.

LAR Enclosure 9, Table E9-3, Assessment of Supplementary Fire PRA Epistemic Uncertainty, for the fire PRA, identifies the treatment of components without cable routing information as always failed to be a source of modeling uncertainty to address lack of cable data. The LAR further states that components deemed risk insignificant or with cable routing too complex to route are not credited in the PRA. Though the assumption that such components always fail in a fire scenario is conservative, NRC staff notes that this conservatism in fire PRA modeling could have a nonconservative impact on the RICT calculations given certain plant configurations. If an SSC is part of a system not credited in the fire PRA or supports a system that is assumed to always fail, then the increase in risk due to taking that SSC out of service could be masked. The LAR cites a sensitivity study that removes these failures concluding that the impact is non-negligible. The LAR states that to ensure the calculated RICTs for the uncredited functions are not significantly affected as a result of assuming the function always fails, RMAs will be developed for the affected RICT LCOs.

A related entry in LAR Enclosure 9, Table E9-3 labeled Treatment of unknown cable locations states, in part, It is common to not know specifically in a room where every cable is located. The LAR states that as a result, All components and cables located in a Fire Area are assumed to be failed by the fire in that area. The extent of the use of this conservative modelling assumption is unclear, and therefore, its impact on the RICT application is not clear. As described above, modeling conservatism could have a nonconservative impact on the RICT Program.

Based on the observations above, it is not clear to NRC staff whether the level of modeling associated with the Lack of Cable Data and Treatment of unknown cable locations issues have no adverse impact on the RICT Program for the plant configurations allowed by the RICT Program. Furthermore, it is not clear whether and how RMAs will be developed for the affected RICT LCOs. Therefore, address the following:

i.

Identify any systems and components that are assumed to always fail in the fire PRA or fire areas where it is conservatively assumed that all components and cables located in a fire area are failed by the fire in that area (instrument air, main feedwater and condensate are already identified).

OR ii.

For the systems or components identified in response to part i. above, provide justification to demonstrate (e.g., sensitivity study) that the treatment of these systems in the fire PRA or fire areas will not adversely impact the RICT Program.

The sensitivity study should at minimum consider a set of LCOs identified assto the most affected by the modeling uncertainty and an estimated RICT less than the backstop of 30 days.

OR iii.

As an alternative to part i. and ii. above, identify the LCO conditions impacted by the treatment of this modelling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

c.

Concerning the fire PRA, portal report PRA-IE-UNCERT_APP5, Revision 1, Disposition of Key Uncertainties: Risk Informed Completion Times (RITS 4b), indicates that one of the important fire PRA frequently asked questions (FAQs) issued by the NRC for fire PRAs may not have been fully addressed. The report states that FAQ 14-0009, Treatment of Well-Sealed MCC [Motor Control Center] Electrical Panels Greater than 440V [volts], (ADAMS Accession No. ML15040A136) is not incorporated into the scoping fire modeling; partially incorporated in the detailed fire modeling. NRC staff notes that the FAQ 14-0009 guidance was issued after NRC issued its SE for Callaways adoption of the National Fire Protection Association 805 program in January 2015. The guidance in FAQ 14-0009 provides the technique for evaluating fire damage from well-sealed MCC cabinets having a voltage greater than 440 V. Accordingly, propagation of fire outside the ignition source panel must be evaluated for all MCC cabinets that house circuits of 440 V or greater. Based on these observations, provide the following:

i.

Describe how fire propagation outside of well-sealed MCC cabinets greater than 440 V is evaluated for the Callaway fire PRA model.

AND ii.

If fires are not propagated outside of well-sealed MCC cabinets greater than 440 V, provide justification (e.g., sensitivity study) that this treatment has no adverse impact on the RICT Program. The sensitivity study should at minimum consider a set of LCOs identified as the most affected by the modeling uncertainty and an estimated RICT less than the backstop of 30 days.

OR iii As an alternative to part i. and ii. above, identify the LCO conditions impacted by the treatment of this modeling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

d.

LAR Enclosure 9, Table E9-5, Assessment of Supplementary Seismic PRA (SPRA)

Epistemic Uncertainty, for the seismic PRA, concerning Model Sensitivity to Seismic HRA [Human Reliability Analysis] Bin Definitions, identifies the treatment of seismic bin definitions as having a potentially significant impact on the baseline risk results. The LAR states, that there is no reasonable alternative to the modeling used, therefore additional sensitivity studies for the RICT application were not performed. Portal report PRA-IE-UNCERT_APP3, Revision 0, Seismic Uncertainty Analysis and Sensitivities, states that sensitivity studies were performed in which the seismic HRA bins were shifted up and down one level and that the impacts on the seismic results were noticeable but not significant to the point they would distort the results. The meaning of this disposition is not completely clear to NRC staff.

Table 3-6 of the report provides the results of four sensitivity cases indicating that the general uncertainty associated with the failure probability of seismic HFEs have a significant impact on seismic core damage frequency (CDF) (e.g., in Case 3, the seismic CDF increases by a factor of 162 percent if the failure probability of these HFEs is increased by a factor of 10). NRC staff notes that the uncertainty associated with seismic HRA bins is addressed in the Callaway 10 CFR 50.69 SE (ADAMS Accession No. ML21344A005) through HRA sensitivity studies that are part of the 10 CFR 50.69 risk-informed categorization process. In consideration of these observations address the following:

i.

Provide justification (e.g., sensitivity) that the treatment of seismic HRA bins do not have an adverse impact on the RICT Program. The sensitivity study should at minimum consider a set of LCOs identified to the most affected by the modeling uncertainty and an estimated RICT less than the backstop of 30 days.

ii.

As an alternative to part (i) above, identify the LCO conditions impacted by the treatment of this modelling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

e.

LAR Enclosure 9, Table E9-4, Assessment of Supplementary Hight Winds PRA Epistemic Uncertainty, for the HW PRA states that based on sensitivity study results that the HW CDF was found to be sensitive to the modeling uncertainty associated with power grid fragility to HW, but that the fragility used in the HW PRA is a reasonable estimate. The LAR also notes that power grid recovery is not credited at any wind speed though recovery is credible for low to moderate wind speeds, which are the wind speeds for which the power grid fragility is uncertain. PRA-IE-INECERT_APP4, Revision 0, High Wind Uncertainty Analysis and Sensitivities, indicates that the sensitivity of the HW CDF to power grid wind fragility is significant. It is not clear to NRC staff that this source of modeling uncertainty has no adverse impact on the RICT Program. In consideration of these observations address the following:

i.

Provide justification (e.g., sensitivity that the treatment of power grid fragility in the HW PRA does not have an adverse impact on the RICT Program. The sensitivity study should at minimum consider a set of LCOs identified as the most affected by the modeling uncertainty and an estimated RICT less than the backstop of 30 days.

ii.

As an alternative to i. above, identify the LCO conditions impacted by the treatment of this modelling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

f.

Concerning the HW PRA, portal report PRA-IE-UNCERT_APP4, Revision 0, explains for safety equipment vulnerabilities associated with openings in Seismic Category 1 structures that it is assumed HW missile impacts result in core damage (and possibly large early release). The report acknowledges this is a conservative approach because the physical separation of the redundant trains of safety-related equipment makes it unlikely that a postulated HW missile through a small opening in the Seismic Category 1 structure would cause the failure of equipment in both trains. The report presents the results of a sensitivity study in which only one train of the safety equipment is assumed to be failed showing that there is an 89 percent decrease in LERF compared to the base case. The report also states that this modeling uncertainty has a significant impact on risk and should be evaluated in more detail in the future. The NRC staff notes that overly conservative modeling can mask the risk significance, potential for a RICT. In consideration of these observations address the following:

i.

Provide justification (e.g., sensitivity study) that the modeling uncertainty does not have an adverse impact on the RICT Program. The sensitivity study should at minimum consider a set of LCOs identified as the most affected by the modeling uncertainty and an estimated RICT less than the backstop of 30 days.

ii As an alternative to i. above, identify the LCO conditions impacted by the treatment of this modelling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

g.

Concerning the internal flooding (IF) PRA, portal report PRA-IE-UNCERT_APP1, Revision 1, Internal Flooding Uncertainty Analysis and Sensitivities, identifies treatment of major IF scenarios as a source modeling uncertainty and presents the results of a sensitivity study showing it to have a significant impact on IF CDF. The analysis concludes that refinement focus on major flood events can result in a significant reduction in overall CDF, and that major flooding scenarios in Turbine Bay 1 are one of most significant contributors to IF CDF due to this conservative treatment. LAR, Table E9-1 for the IE (including flooding) does not disposition this source of modeling uncertainty (or any other source of IF specific modeling uncertainty). The NRC staff notes that overly conservative modeling can mask the risk significance potential for a RICT. In consideration of these observations address the following:

i.

Provide justification (e.g., sensitivity study) that the IF modeling uncertainty does not have an adverse impact on the RICT Program. The sensitivity study should at minimum consider a set of LCOs identified as the most affected by the modeling uncertainty and an estimated RICT less than the backstop of 30 days.

ii.

As an alternative to i. above, identify the LCO conditions impacted by the treatment of this modelling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

APLA QUESTION 03 - Modeling of Analog Instrumentation and Control (I&C) Systems For many TS LCO conditions listed in LAR Enclosure 1, Table E1-1, the table indicates that I&C modeling is insufficient to reflect the plant configurations in the RICT Program and, therefore, the inoperability of the associated SSC (e.g., a channel) will be modeled using a surrogate event. For several other LCO conditions, the LAR states that the I&C is explicitly modeled.

Accordingly, there appears to be a general variability for the level of detail used to model the I&C systems and corresponding potential modeling uncertainty. It is not clear to NRC staff whether there is sufficient detail to support implementation of the proposed LCO conditions other than those specifically identified in LAR Table E1-1 that will be modeled using a surrogate event. In consideration of this observation address the following:

Briefly describe the explicit PRA modeling for these I&C SSCs and include discussion of the scope of the I&C components (e.g., bistables, relays, sensors, integrated circuit cards).

Provide justification that supports why modeling these components is sufficient to calculate the risk associated with the failure of the I&C system(s).

APLA QUESTION 04 - Modeling of Digital I&C Systems Section 2.3.4 of NEI 06-09, Revision 0-A, states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT Program and that sensitivity studies should be performed on the base model prior to initial implementation of the RICT Program on uncertainties that could potentially impact the results of a RICT calculation.

Regarding digital I&C, there is a lack of consensus industry guidance for modeling these systems in plant PRAs for use to support risk-informed applications. In addition, known modeling challenges exist, such as: (1) lack of industry data, (2) the difference between digital and analog system failure modes, and (3) the complexities associated with modeling software failures including common cause software failures. Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT Program. It is not clear to the NRC staff whether the licensee credits digital systems in the PRA models that will be used in the RICT Program, and if so, how this modeling impacts the PRA used. In consideration of these observations address the following:

a.

Identify any digital I&C system(s) credited in the probabilistic risk analysis models.

b.

For the systems identified in a., briefly describe the modeling. Include the results of a sensitivity study performed for each digital system demonstrating that the uncertainty associated with probabilistic risk analysis modeling has no adverse impact on the RICT Program.

c.

As an alternative to b. above, identify the LCO conditions impacted by the treatment of this modelling uncertainty, along with potential RMAs that will be applied for the associated RICT. Include justification for why the application of RMAs will be sufficient to address the modeling uncertainty.

APLA QUESTION 05 - Impact of Seasonal Variations The Tier 3 requirement of RG 1.177, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, stipulates that a licensee should develop a program that ensures that the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06-09 and the accompanying NRC SE to this guidance state that for the impact of seasonal changes either conservative assumptions should be made or, the PRA should be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration.

LAR Enclosure 8, Attributes of the Real Time Risk Model, Section 2.0, presents an unconnected bullet point near the end of the Section stating: No success criteria change based on seasonal variation. Given that changes in PRA success criteria due to seasonal changes will apparently not be made in the RTR model, confirm that conservative or bounding assumptions are used to encompass the impact of any seasonal variations on PRA modeling consistent with the guidance in NEI 06-09, Revision 0-A.

APLA QUESTION 06 - In-Scope LCOs and Corresponding PRA Modeling LAR Enclosure 1, Table E1-1 identifies each TS LCO proposed to be in the RICT Program, describes whether the systems and components participating in the TS LCO are modeled in the PRA, and compares the design basis and PRA success criteria. For certain TS LCO Conditions, the table explains that the associated SSCs are not explicitly modeled in the PRAs and identifies a surrogate to represent failure of the applicable function. For some LCO conditions, it is not clear how the identified surrogate is representative of the failure of the TS function. Therefore, address the following:

a.

LAR enclosure 1, Table E1-1 states for TS LCO 3.6.6, Condition A (One containment spray train inoperable) and Condition C (One containment cooling train inoperable),

that the design-basis success criterion is One of two trains and the PRA success criterion is None. The LAR states in Table E1-1 that containment cooling is not modeled in the PRA because hydraulic analysis shows that containment cooling does not impact which accident sequences contribute to LERF]. Table E1-1 does not define a surrogate event for modeling the unavailability of a containment spay and cooling system in the RICT calculations, and LAR Enclosure 1, Table E1-2, RICT Estimates, does not provide an example RICT estimate for LCOs 3.6.6.A and 3.6.6.C. Accordingly, it appears that the assumption made in the RICT calculation is that the unavailability of the containment spray and cooling systems has zero impact on CDF and LERF. LAR, Table E1-1, shows that functions addressed by the containment spray and cooling system are containment pressure and temperature control and fission produce retention (which NRC staff understands to be iodine removal according to the TS Basis for LCO 3.6.6.A presented in LAR Attachment 4). To support NRC staff understanding of whether the PRA modeling of the containment spray and cooling system is sufficient to support the RICT application, address the following:

i.

Confirm that the cited hydraulic analysis addresses both the fission product removal as well as containment pressure and temperature control functions credited in the PRA models. Include clarification that the success or failure of the containment spray and cooling system has no impact on the success of heat and fission product removal functions as modeled in the PRAs.

OR ii.

If the cited hydraulic analysis does not address both the of heat and fission product removal functions, provide justification (e.g., sensitivity study) to confirm that the modeling treatment has an no adverse impact on the RICT Program (e.g., provide an estimated RICT for LCOs 3.6.6.A and 3.6.6.C in LAR Table E1-2 that demonstrates the results of the RICT calculation for these LCO conditions).

b.

LAR Enclosure 1, Table E1-1 states for the turbine trip and feedwater function of TS LCO 3.3.2 Condition I (i.e., one channel inoperable, SG water level High-High, P-14) that the design-basis success criterion is [t]wo of four on one of four Steam Generators and the PRA success criterion is [n]one. In Table E1-1 of Enclosure 1 to the LAR, the licensee states that this trip function is not explicitly modeled, but SG level high-high signals for the main feedwater isolation are modeled and will be used as a conservative surrogate in the RICT calculations. The LAR further states that trip of the MFW [main feedwater] pumps and closure of the pump discharge valves is not explicitly modeled. It is not clear how SG level high-high signal for MFW isolation is a sufficient surrogate for the turbine trip and feedwater isolation functions.

Provide the impact the SG level high-high signal has on other modeled SSCs and provide justification for why the surrogate event identified for LCO 3.3.2.I is appropriate.

c.

LAR enclosure 1, Table E1-1 states for the steam line isolation and reactor trip P-4 engineered safety features actuation system interlock function of TS LCO 3.3.2 Condition F (i.e., One channel or train inoperable) that [a] channel of main steam isolation will be used as a conservative surrogate for the steam line isolation function.

The proposed modeling does not appear to be conservative.

Provide justification for why the surrogate described for LCO 3.3.2.F is sufficient for the RICT calculations.

APLA QUESTION 07 - Open Phase Condition The NRC issued Bulletin 2012-01, Design Vulnerability in Electric Power System, dated July 27, 2012 (ADAMS Accession No. ML12074A115). As part of the initial Voluntary Industry Initiative for mitigation of the potential for the occurrence of an open phase condition (OPC) in electrical switchyards, some licensees have installed plant modifications (e.g., Open Phase Isolation System (OPIS)). NEI 19-02, Guidance for Assessing Open Phase Condition Implementation Using Risk Insights (ADAMS Accession No. ML19172A086),

discusses performing estimations of OPC and OPIS risk, and further states that the risk impact of an OPC can vary widely dependent on electrical switchyard configuration and design.

Furthermore, Section 2.3.4 of NEI 06-09, Revision 0-A, states, in part, that Criteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations. Considering these observations, provide the following information:

a.

Discuss the evaluation of the risk impact associated withOPCevents including the likelihood ofOPC-initiated plant trips and the impact of those trips on PRA modeled SSCs.

b.

Confirm whether any installed equipment and associated operator actions are credited in the PRAs to address OPC. If equipment and associated operator actions are credited, then provide the following information:

i.

Describe the equipment and associated actions that are credited in the PRA models.

ii.

Describe the impact that this treatment, if any, has on key assumptions and sources of uncertainty.

iii.

Discuss HRA methods and assumptions used for crediting alarm manual response.

iv.

Discuss howOPC-related scenarios are modeled for non-IE scenarios such as internal floods, fire, and seismic.

v.

Regarding inadvertent actuation of the open phase detection systemopen phase isolation, include a discussion of the risk impact of inadvertent OPIS actuation and justification for its exclusion.

c.

IfOPCand the open phase detection system and ability to manually mitigate an open phase condition are not included in the application PRA models, then provide justification that the exclusion of this failure mode and mitigating systemdoes not impact the RICT Program.

d.

As an alternative to c., propose a mechanism to ensure thatOPC-related scenarios are incorporated into the PRA modelsprior to implementing the RICT Program.

DRA, PRA Licensing Branch C (APLC)

APLC QUESTION 01 - External Event Plant Configurations Impacts Section 2.3.1, Configuration Risk Management Process and Application of Technical Specifications, Item 7, of NEI 06-09, Revision 0-A, states, in part, 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 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.

In Section 3.2, Impacts to RICT, of Enclosure 4, Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models, to the LAR, the licensee concluded that all non-modeled external hazards are screened for all RICT configurations.

The licensee specifically discussed those hazards that are screened using criteria other than C2 through C5. However, the licensee did not include external flooding, intense precipitation, and waves in this list for evaluation of RICT configurations. These hazards are screened based on Criterion C1 (or partially) in LAR Enclosure 4, Table E4-1, External Hazards Screening, but are not discussed for plant configurations in Section 3.2 of Enclosure 4 of the LAR.

Confirm that plant configurations for external flooding, intense precipitation, and waves hazards have no impact on the proposed RICTs.

APLC QUESTION 02 - Considerations of Loss of Offsite Power in IE and HW PRAs Section 2.0, Translation of Baseline PRA Model for Use in Real-Time Risk, of Enclosure 8 to the LAR describes the configuration risk management program (CRMP) software, which is designed to quantify the configuration for IE, IF, fire, HW, and seismic risk when calculating the RMA time and RICT. It is not clear to NRC staff how the software treats some common initiating events in which both IE and HW PRAs consider this event without double counting, such as a loss of offsite power.

Describe how the CRMP software treats some common initiating events from different PRA models.

DEX, Electrical Engineering Branch (EEEB)

EEEB Question 01 In Section 8.3.2.1, Description, of the Callaway FSAR (ADAMS Accession No. ML21193A191), the licensee refers to four independent Class 1E 125-Vdc (volts direct current) subsystems.

Please explain why TSs 3.8.4.A and 3.8.9.C in LAR Enclosure 1, Table E1-1, refers to only to two direct current (DC) electrical power distribution subsystems in Column 3 (SSCs Addressed) and one of two DC electrical power distribution subsystems in Column 6 for Design Success Criteria.

EEEB Question 02 In Table E1-1 of Enclosure 1 to the LAR, for TSs 3.8.1.B and 3.8.1.D, please explain the following:

a.

Why two terms diesel generator (DG) and emergency diesel generator (EDG) are used for diesel generators.

b.

Why Column 6 for Design Success Criteria (DSC) does not use similar information as Column 3 (SSCs Addressed) to indicate what an operable diesel generator must do.

EEEB Question 03 In Table E1-1 of Enclosure 1 to the LAR, please explain if the DSC for TS 3.8.1.D should be corrected to one offsite circuit OR one of two EDGs.

EEEB Question 04 Explain if LAR Enclosure 1, Table E1-1 for TS 3.8.1.F should refer to the following in the design success criteria:

a.

One of two LSELS [load shedder and emergency load sequencer] available for its operable DG and assigned 4.16 kilovolt engineered safety feature bus.

b.

Trains A and B as referenced in required alternating current (AC) electrical sources to be OPERABLE in LCO 3.8.1 EEEB Question 05 Explain why TS 3.8.1.F requires an offsite circuit be declared inoperable.

EEEB Question 06 Explain if LAR Enclosure 1, Table E1-1 for TS 3.8.7.A should provide the following:

Column 3 (SSCs Addressed) and 6 (DSC)

i.

What channels are associated with each train of inverters required and if should be identified in both column 3 and 6.

ii. Relationship of trains to channels per USFAR Section 8.3.1.1.5.

iii. If swing inverter can be supplied by non-Class 1E source for a design-basis accident.

EEEB Question 07 Indicate for LAR Enclosure 1, Table E1-1 for TS 3.8.9.A how subsystems are associated with the terms load groups and trains used in UFSAR 8.3.1.1.2 and LCO 3.8.1.

EEEB Question 08 Indicate for LAR Enclosure 1, Table E1-1 for TS 3.8.9.B how subsystems are associated with the term channels used in UFSAR 8.3.1.1.5.

EEEB Question 09 Indicate for LAR Enclosure 1, Table E1-1 for TS 3.8.9.C how subsystems are associated with the term trains used in LCO 3.8.4.

EEEB Question 10 Is RICT beneficial for LCO 3.8.1.C since RICT estimate is 1.1 days and previous TS completion time was 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> given possibility of fluctuation of conditional risk uncertainty for online maintenance?

EEEB Question 11 In Table E1-2 of the Enclosure to the LAR, explain the potential discrepancy between RICT times for LCO 3.8.1.A or B to LCO 3.8.9.A.

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