ML20337A301

Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - Ritstf.

ML20337A301
Person / Time
Site:	Peach Bottom
Issue date:	12/02/2020
From:	David Helker Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2020-LLA-0120
Download: ML20337A301 (59)

Text

200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90 December 2, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Subsequent Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b."

References:

1. Letter from David P. Helker, Exelon Generation Company, LLC, to the U.S. Nuclear Regulatory Commission, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b,'" dated May 29, 2020 (ADAMS Accession No. ML20150A007).

2. Letter from Jennifer C. Tobin, U.S. Nuclear Regulatory Commission, "Peach Bottom Atomic Power Station, Units 2 and 3 Regulatory Virtual Audit Plan Regarding License Amendment Request to Adopt TSTF-505, Revision 2 (EPID L-2020-LLA-0120)," dated October 21, 2020 (ADAMS Accession No. ML20290A524).

By letter dated May 29, 2020 (Reference 1), Exelon Generation Company, LLC (Exelon) requested approval for proposed changes to the Technical Specifications (TS), Appendix A of Subsequent Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, respectively.

The proposed amendment would modify TS requirements to permit the use of Risk Informed Completion Times (RICT) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493).

By letter dated October 21, 2020 (Reference 2), the NRC notified Exelon of their intent to conduct a regulatory virtual audit the week of November 9, 2020 with Exelon staff and associated contractors in support of the license amendment request (LAR) in Reference 1.

The letter contained a regulatory virtual audit plan with attached audit questions.

This letter is a supplement to the Reference 1 LAR. The attachment to this letter provides a response to several of the audit questions posed by the NRC staff during the regulatory virtual audit.

Supplement to License Amendment Request Adopt Risk Informed Completion Times TSTF-505 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 December 2, 2020 Page 2 Exelon has reviewed the information supporting the No Significant Hazards Consideration and the Environmental Consideration that was previously provided to the NRC in Reference 1. The additional information provided in this LAR supplement does not impact the conclusion that the proposed license amendment does not involve a significant hazards consideration. The additional information also does not impact the conclusion that there is no need for an environmental assessment to be prepared in support of the proposed amendment.

There are no regulatory commitments contained in this supplement.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), Exelon is notifying the Commonwealth of Pennsylvania of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

Should you have any questions concerning this submittal, please contact Glenn Stewart at (610) 765-5529.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2nd day of December 2020.

Respectfully, David P. Helker Sr. Manager, Licensing and Regulatory Affairs Exelon Generation Company, LLC

Attachment:

Response to NRC Audit Questions cc: USNRC Region I, Regional Administrator w/ attachments USNRC Project Manager, PBAPS "

USNRC Senior Resident Inspector, PBAPS "

Director, Bureau of Radiation Protection - Pennsylvania Department of Environmental Protection "

S. Seaman - State of Maryland "

ATTACHMENT Supplement to License Amendment Request Peach Bottom Atomic Power Station, Units 2 and 3 Docket Nos. 50-277 and 50-278 Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b."

Response to NRC Audit Questions

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 1 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

References:

1. Letter from David P. Helker, Exelon Generation Company, LLC, to the U.S.

Nuclear Regulatory Commission, "License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, 'Provide Risk-Informed Extended Completion Times -

RITSTF Initiative 4b,'" dated May 29, 2020 (ADAMS Accession No. ML20150A007).

2. Letter from Jennifer C. Tobin, U.S. Nuclear Regulatory Commission, "Peach Bottom Atomic Power Station, Units 2 and 3 Regulatory Virtual Audit Plan Regarding License Amendment Request to Adopt TSTF-505, Revision 2 (EPID L-2020-LLA-0120)," dated October 21, 2020 (ADAMS Accession No. ML20290A524).

By Letter dated May 29, 2020 (Reference 1), Exelon Generation Company, LLC (Exelon) requested approval for proposed changes to the Technical Specifications (TS), Appendix A of Subsequent Renewed Facility Operating License Nos. DPR-44 and DPR-56 for Peach Bottom Atomic Power Station (PBAPS), Units 2 and 3, respectively.

The proposed amendments would modify TS requirements to permit the use of Risk Informed Completion Times (RICT) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493).

By letter dated October 21, 2020 (Reference 2), the NRC notified Exelon of their intent to conduct a regulatory virtual audit the week of November 9, 2020 with Exelon staff and associated contractors in support of the license amendment request (LAR) in Reference 1. The letter contained a regulatory virtual audit plan with attached audit questions.

This attachment provides a response to several of the audit questions posed by the NRC staff during the regulatory virtual audit. NOTE: The NRC staffs questions are in italics throughout this attachment to distinguish from the Exelon responses.

Probabilistic Risk Assessment Licensing Branch A (APLA) Audit Questions APLA QUESTION 04 - Probabilistic Risk Assessment Modeling of Vacuum Breakers (Implementation Items)

LAR Attachment 6 lists the following implementation items that must be completed prior to implementation of the RICT program to satisfy the guidance in NEI 06-09 that the PRA reflect the as-built, as-operated plant and that the PRA technical adequacy is acceptable:

• Exelon will ensure that the reactor building-to-suppression chamber vacuum breakers are modeled in the Peach Bottom PRA with sufficient detail to accurately calculate the RICT.

• Exelon will ensure that the suppression chamber-to-drywell vacuum breakers are modeled in the Peach Bottom PRA with sufficient detail to accurately calculate the RICT.

LAR Attachment 6 also states that if implementation of any of these changes constitutes a PRA upgrade as defined in the PRA standard, as endorsed by RG 1.200, then a focused-scope peer

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 2 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 review will be performed on these changes, and any findings will be resolved and incorporated in the PRA prior to the implementation of the RICT program. However, it is unclear to the NRC staff how the addition of these system models will meet CC-II of the PRA standard, as endorsed by RG 1.200. In light of these observations, provide the following information:

Regarding the implementation items identified above, describe how the associated systems will be adequately modeled in the PRA to CC-II. Include in this discussion:

i. How mechanical components, instrument channels, logic components, and other relevant system components will be modeled.

Response

Both the Full Power Internal Events (FPIE) and Fire PRA (FPRA) models as well as the Real-Time Risk (RTR) tool will be updated with the items identified in LAR Attachment 6 prior to implementation of the RICT program at PBAPS.

Any mechanical components, instrument channels, logic components, and other relevant system components will be modeled in a manner analogous to the already existing fault trees and consistent with the Capability Category II requirements of the PRA standard.

No new methods are expected to be employed.

ii. Provide details of the success criteria for these systems. If the PRA success criteria do not match the DSC, then provide a justification for the PRA success criteria.

Response

The success criteria will be modeled consistent with the design basis as it is described in LAR Enclosure 1 (see Table E1-1, TS 3.6.1.5.C and TS 3.6.1.6.A).

iii. Confirm whether these implementation items apply to both the internal events PRA (IEPRA) and the fire PRA (FPRA). Accordingly, adjust the wording for each of the affected implementation items in LAR Attachment 6. If any of these implementation items will not be applied to the FPRA, then justify the position that the FPRA model will be sufficient to support the RICT program.

Response

These implementation items apply to both the IEPRA and FPRA. Because the FPRA model uses the IEPRA model details and structure as its baseline, the changes incorporated into the IEPRA for these implementation items are also incorporated into the FPRA model.

LAR Attachment 6 states that, "logic will be added to the PRA to model the impact of these vacuum breakers." The implementation writeup does not state specifically that the items will be incorporated into the IEPRA or the FPRA since the changes will impact both models which collectively are simply referred to as the PRA model in Attachment 6 of the LAR.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 3 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 APLA QUESTION 05 - Probabilistic Risk Assessment Modeling and Uncertainty of FLEX Equipment and Actions The NRC memorandum dated May 30, 2017, Assessment of the Nuclear Energy Institute 16-06, Crediting Mitigating Strategies in Risk-Informed Decision Making, Guidance for Risk-Informed Changes to Plants Licensing Basis (ADAMS Accession No. ML17031A269),

provides the NRCs staff assessment of challenges to incorporating FLEX equipment and strategies into a PRA model in support of risk-informed decisionmaking in accordance with the guidance of RG 1.200.

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

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

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

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

Regarding uncertainty, Section 2.3.4 of NEI 06-09 states that PRA modeling uncertainties shall be considered in the 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 an RICT calculation.

NEI 06-09 also states that the insights from the sensitivity studies should be used to develop appropriate risk management actions (RMAs), including highlighting risk-significant operator actions, confirming availability and operability of important standby equipment, and assessing the presence of severe or unusual environmental conditions. Uncertainty exists in PRA modeling of FLEX related to the equipment failure probabilities for FLEX equipment used in the model, the corresponding operator actions, and pre-initiator failure probabilities.

Therefore, FLEX modeling assumptions can be key assumptions and sources of uncertainty for the RICTs proposed in this application.

LAR Enclosure 9, Table E9-1, indicates that FLEX equipment and actions have been credited in the IEPRA. The LAR states that a sensitivity study was performed for the IEPRA to address this issue. The LAR stated that the sensitivity did not significantly impact the RICT values. As

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 4 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 part of its audit (ADAMS Accession No. ML20217L346), the NRC staff noted that Section 8 of PRA Notebook PB-MISC-043 provided results of a sensitivity study where the failure probability of the FLEX injection pump and diesel generator was significantly increased.

However, the NRC staff notes the significant challenges of modeling FLEX equipment and actions without sufficient industry data and without a consensus HRA approach to address unique aspects of FLEX actions.

The NRC staff also notes that the difference between failure rates associated with permanently installed safety-related diesel generators and portable non-safety-related diesel generators could be greater than a factor of 10 without consideration of further uncertainty. It is unclear to the NRC staff whether the stated sensitivity study addressed the uncertainties associated with estimating HEP values for FLEX actions, especially for non-operator trained actions. Given the observations above, it is not clear whether the sensitivity study performed to assess the impact of crediting FLEX equipment and actions is sufficient to conclude that the impact to the RICT program of the uncertainties associated with modeling FLEX is negligible. For this reason, and to understand the credit that will be taken for FLEX equipment and actions in the RICT program, address the following separately for the IEPRA, internal flooding PRA, and FPRA:

a) Provide results of LCO-specific sensitivity studies that assess the removal of FLEX credit on RICT calculations.

Response

In the base PRA model, a factor of 2 is applied to the PBAPS-specific unreliability failure probabilities for similar equipment in order to estimate the FLEX equipment unreliability failure probabilities. This escalation is a reasonable approximation of the unreliability of FLEX equipment until industry data is published.

As documented in Section 8.9 of PB-MISC-043 [1], a sensitivity analysis was performed to assess the potential impact on the TSTF-505 RICT calculations. The FLEX equipment failure probabilities were escalated by a factor of 5 when compared to the base non-FLEX equipment. For the FLEX diesel generators, the failure to start (FTS) failure rate increased to 1.64E-02 and the failure to run (FTR) failure rate increased to 8.95E-02. For the FLEX diesel-driven pumps, the FTS failure rate increased to 1.58E-02 and the FTR failure rate increased to 2.38E-01. Given the magnitude of these failure rates, a factor of 5 on the non-FLEX equipment failure rates is assessed to be suitable for assessing potential impacts on TSTF-505 RICT calculations.

However, an additional sensitivity analysis was performed, where the FLEX equipment is not credited for the RICT calculations. The results of this sensitivity are provided in Table 5.a-1 below.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 5 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 5.a-1: Example RICT Calculations Original RICT Sensitivity w/ No Tech Spec TS Condition Estimate FLEX Credit (days) (days) 3-5-1-C HPCI system inoperable 29.7 29.7 3-6-2-3-A One RHR suppression pool 30 30 cooling subsystem inoperable 3-8-1-A One offsite AC power circuit 30 30 inoperable 3-8-1-B One EDG inoperable 30 30 3-8-1-D Two or more offsite AC power 15.4 8.8 circuits inoperable 3-8-1-E One offsite AC power circuit 25.7 20.9 AND one EDG inoperable HPCI - High Pressure Coolant Injection RHR - Residual Heat Removal EDG - Emergency Diesel Generator b) Regarding HRA, address the following items:

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

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

Response

There are PBAPS operator actions related to the implementation of FLEX strategies that contain the activities described in Sections 7.5.4 and 7.5.5 of NEI 16-06. Credit is taken for the following in each of the PBAPS PRA models.

1) Deploying and aligning a portable FLEX 480V generator to restore battery chargers (limited to extended loss of AC power (ELAP) scenarios).

2) Deploying and aligning a portable FLEX pump for reactor pressure vessel (RPV) injection (limited to ELAP scenarios).

3) Prolonged reactor core isolation cooling (RCIC) operation without Suppression Pool Cooling via partial RPV depressurization and containment venting using the permanently installed Hardened Containment Vent System (HCVS).

The credited operator actions for each of the mitigating strategies listed above are shown below. These actions are similar to other operator actions included in the PRA models and are evaluated using approaches consistent with the

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 6 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 endorsed ASME/ANS RA-Sa-2009 PRA standard as documented in the PBAPS Internal Events HRA notebook.

1) Success of the FLEX generators includes required operator actions for DC Load Shed (QHULS-ACDXI2), deploying and starting the FLEX generators and aligning them to supply the battery chargers (QHUFXL13DXI2), and refueling the FLEX generators (QHUDFUELDXI2).

2) Success of the diesel-driven FLEX pumps includes required operator actions for deploying and aligning a FLEX pump to take suction on the Emergency Cooling Tower (ECT) supply and inject to the RPV (QHUFXRPVDXI2) and refueling the FLEX pump (QHUPFUELDXI2).

3) Success of prolonged RCIC operation includes required operator actions for performing partial RPV depressurization and early containment venting via the HCVS (RHUVENT1DXI2), bypassing interlocks to allow continued RCIC operation after containment venting (RHUVENT2DXI2), aligning a FLEX pump for suppression pool makeup from the ECT supply (QHUFXTRSDXI2), and refueling the FLEX pump (QHUPFUELDXI2).

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

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

Response

In order to provide "bounding realistic estimates" of the HEPs and joint HEPs (JHEPs) for the actions identified in the response to question 5.b.i, the 95th percentile values for the events are proposed as appropriate values.

While other options are available to estimate bounding values, such as applying a factor of 10 to the base failure probabilities, the 95th percentile values are intended to represent the likely upper bound HEPs for the actions, the values were developed using the human reliability analysis (HRA) methodology, and they have a documented quantitative basis. The Electric Power Research Institute (EPRI) Human Reliability Analysis Calculator (HRAC) v 5.2 methods were used to calculate the 95th percentile values for independent and dependent HEPs. While these estimates may be lower than the result obtained by applying a factor of 10 multiplier to the base values, EPRI 3002013018 has provided a systematic approach to addressing the issues addressed in NEI 16-06 Sections 7.5.4 and 7.5.5 and

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 7 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 insights from the guidance indicate that for some of the activities identified, the available methodologies may overestimate the task failure probabilities.

• Debris removal: The suggested treatment is to account for the task in the timeline rather than to quantify a HEP, because commonly used HRA methodologies do not address this type of work. While this could be considered to be a non-addressed contributor to the FLEX strategy failure probability, debris removal is not expected to be a major factor for the FPIE, internal flooding, or internal Fire scenarios. In addition, the recovery times available for the potentially impacted actions QHUFXRPVDXI2 and QHUFXL13DXI2 are extensive. For example, for QHUFXRPVDXI2, the time available for recovery is about 430 minutes and for QHUFXL13DXI2 the time is about 180 minutes. Additional time spent working on debris removal is not expected to be a significant issue for these FLEX actions.

• Transportation of Portable Equipment: EPRI 3002013018 provided an approach to assessing the risk from potential transportation errors and they were determined to be negligible contributors for PBAPS. No reasonable variations in the probability of failure for these tasks is expected to impact the action HEPs.

• Installation of Equipment at Staging Location/Addressing Complex Actions in Mitigating Strategies: EPRI 3002013018 indicates that a weakness of using Technique for Human Error Rate Prediction (THERP) to assess tasks comprised of many steps is that that the aggregate HEP can be unrealistically high. This implies that the use of a factor of 10 multiplier would provide a further, undesirable bias to this sensitivity case.

o For those tasks that are not directly represented by THERP data, EPRI 3002013018 requires a basis to be developed for surrogate values used. This may lead to a greater degree of uncertainty in the HEPs for sub-steps, but the surrogate values are generally applied to those actions that are comprised of many steps (e.g.,

making hose connections), which would reduce the likelihood of underestimating the FLEX action HEP. The QHUFXRPVDXI2 and QHUFXL13DXI2 actions contain subtasks that have stated bases for the surrogate values used in place of THERP data.

• Routing of Hoses and Cables: The treatment of these tasks fits into the category of self-revealing errors in EPRI 3002013018. If the hoses or cables are incorrectly routed such that they cannot be connected to equipment, the action could not progress. If there is adequate time for recovery (true for the relevant PBAPS actions) and no irreversible consequence occurs, the errors are treated as negligible contributors to risk. No reasonable variations in the probability of failure for these tasks is expected to impact the action HEPs.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 8 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 The same approach of applying the 95th percentile values to the base failure probabilities was used for the JHEPs. For JHEPs that were set to the minimum JHEP value of 1E-6 or 5E-7, if the calculated 95th percentile value was still below the minimum JHEP value, the minimum JHEP value of 1E-6 or 1E-7 was retained as the 95th percentile value. The baseline values and the 95th percentile values for each hazard are provided in Table 5.b.ii-1 below for the independent HFEs along with examples of the JHEPs that were used in the models. The entire list of JHEPs containing one or more of the HFEs identified in the response to 5.b.i has not been provided in this response due to the large number of events that comprise the list (over 100 dependent events between the FPIE and FPRA models):

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 9 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 5.b.ii-1: Overview of FLEX HEP and Joint HEP 95th Percentile Values (1,2)

BEID Description Dist. FPIE / Flood Fire Type Base 95th Base 95th QHUDFUELDXI2 OPS FAILS TO REFUEL FLEX DIESEL B 5.0E-02 1.3E-01 5.0E-02 1.3E-01 QHUFXL13DXI2 OPERATOR FAILS TO ALIGN FLEX GENERATOR TO LC B 3.7E-02 1.0E-01 3.7E-02 1.0E-01 E134 AND LC E334 QHUFXRPVDXI2 OPERATOR FAILS TO ALIGN FLEX FLOW PATH TO B 2.5E-02 7.1E-02 2.5E-02 7.2E-023 RPV QHUFXTRSDXI2 OPERATOR FAILS TO ALIGN FLEX FLOW PATH TO B 2.6E-02 7.3E-02 2.6E-02 7.4E-023 TORUS QHULS-ACDXI2 DEEP DC LOAD SHED WHEN ELAP DECLARED (STEPS B 1.9E-02 5.3E-02 2.1E-02 6.0E-02 FOR RCIC)

QHUPFUELDXI2 OPS FAILS TO REFUEL FLEX PUMP B 5.0E-02 1.3E-01 5.0E-02 1.3E-01 RHUVENT1DXI2 OPS FAILS TO PARTIALLY DEPRESS RPV AND B 2.7E-02 7.5E-02 1.0E-01 2.5E-01 INITIATE EARLY CONTAINMENT VENT RHUVENT2DXI2 OPS FAILS TO BYPASS INTERLOCKS TO ALLOW B 9.7E-02 2.4E-01 1.00 1.00 CONTINUED RCIC OPERATION IN EL JHEP-2C1-EV-00375 Dep HEP for BHU-PUMPDXI2, QHUFXRPVDXI2, L 1.5E-06 1.5E-05 N/A N/A JHUHWINJDXD2 JHEP-2C1-EV-00105 Dep HEP for JHU--ECTDXI2, QHUFXRPVDXI2, L 5.0E-07 2.2E-06 N/A N/A JHUHWINJDXD2 JHEP-2C1-EV-00085 Dep HEP for DHU--SPCDXI2, DHU--SPCDXD2, ZHU--

L 5.0E-07 5.0E-07 N/A N/A CSTDXI2, QHUFXRPVDXI2 JHEP-2HF-NE-00984 Dep HEP for 2ISOP-ISOLATEH---F, EHU-SE11DXI0-F, L N/A N/A 1.9E-06 1.9E-05 DHU-ASDCDXI2-FRA, QHUDFUELDXI2-F JHEP-2CF-NE-00606 Dep HEP for ZHU1089XDXI2-F, ZHU--CSTDXI2-F, L N/A N/A 5.0E-07 1.2E-06 JHUHWINJDXD2-F, QHUFXRPVDXI2-F JHEP-2CF-EV-01043 Dep HEP for EHUOSP12DXI2-FRA, DHU--SPCDXI2-F, DHU--SPCDXD2-F, BHU-PUMPDXI2-F, JHUHWINJDXD2- L N/A N/A 5.0E-07 5.0E-07 F, QHUFXRPVDXI2-F

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 10 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 5.b.ii-1: Overview of FLEX HEP and Joint HEP 95th Percentile Values BEID(1,2) Description Dist. FPIE / Flood Fire Type Base 95th Base 95th Notes:

1. FPRA HFEs are named similar to FPIE PRA HFEs but have a -F or -FRA suffix.

2. Key to Operator HEPs included in Table:

• 2ISOP-ISOLATEH---F - OPERATOR FAILS TO ISOLATE PATH GIVEN SIGNALS FAIL

• BHU-PUMPDXI2 - OPERATORS FAIL TO RESTART PREV RUNNING CRD PUMP OR START THE STANDBY

• DHU-ASDCDXI2-FRA - FAILURE TO INITIATE ALTERNATE SDC PER SE-10 (MCRAB)

• DHU--SPCDXD2 - OPERATORS FAIL TO INITIATE RHR IN SPC MODE (NON-ATWS) TO AVOID PCPL (LATE, CONDITIONAL)

• DHU--SPCDXI2 - OPERATORS FAIL TO INITIATE RHR IN SPC MODE (NON-ATWS) AVOID HCTL

• EHUOSP12DXI2-FRA - OPERATORS FAIL TO MANUALLY MANIPULATE BREAKERS FOR OSP TO 4KV BUS E12

• EHU-SE11DXI0-F - FAILURE TO X-TIE EMERGENCY AC POWER PER SE-11

• JHU--ECTDXI2 - OPERATOR FAILS TO CORRECTLY ALIGN HPSW FOR COOLING TOWER FLOW

• JHUHWINJDXD2 - OPERATORS FAIL TO OPEN HPSW CROSS-TIE (CONDITIONAL)

• JHUHWINJDXD2-F - OPERATOR FAILS TO INJECT WITH HPSW THRU RHR (LATE)

• QHUDFUELDXI2-F - OPS FAILS TO REFUEL FLEX DIESEL

• QHUFXRPVDXI2 - OPERATOR FAILS TO ALIGN FLEX FLOW PATH TO RPV

• ZHU1089XDXI2-F - MANUAL ACTION TO ALIGN ALTERNATE POWER TO HPSW MOVs

• ZHU--CSTDXI2 - OPERATORS FAIL TO DIAGNOSE NEED TO REFILL CST (ANY MEANS; COGNITIVE-ONLY)

3. While the independent HEPs for the given action match in the FPIE and Fire results, there are slight rounding differences inherent to the HRAC-supplied 95th percentile values.

SDC - Shutdown Cooling

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 11 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

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

Response

The results of the HEP sensitivities are provided as composite sensitivity cases (Equipment and HEP adjustments applied) in the response to audit question 5.b.ii.3 below. Note that for the purposes of this sensitivity, the equipment failures were set to 5x the generic data and is consistent with the original sensitivity evaluated in PB-MISC-043 [1].

3. Discuss composite sensitivity studies of the RICT results to the operator action HEPs and the FLEX equipment reliability uncertainty sensitivity study.

Response

The results of the FLEX HEP and equipment sensitivity studies are shown in Table 5.b.ii-2 below. Sensitivities were performed for LCOs with FLEX portable equipment failure rates modified and FLEX strategy HEP/JHEPs modified. The cases were developed by setting equipment failures to 5x generic data and Independent and Joint HEPs to the 95th percentile values.

As shown in Table 5.b.ii-2, the number of RICT days for each LCO are not highly sensitive to the reliabilities of the FLEX equipment or operator actions associated with the FLEX equipment.

Table 5.b.ii-2: Example RICT Calculations Sensitivity w/

Original RICT Equipment Tech Spec TS Condition Estimate Reliability and (days) HEP adjustments (days) 3-5-1-C HPCI system inoperable 29.7 29.7 3-6-2-3-A One RHR suppression pool 30 30 cooling subsystem inoperable 3-8-1-A One offsite AC power circuit 30 30 inoperable 3-8-1-B One EDG inoperable 30 30 3-8-1-D Two or more offsite AC power 15.4 8.8 circuits inoperable 3-8-1-E One offsite AC power circuit 25.7 20.6 AND one EDG inoperable

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

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 12 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

Response

Based on the results of the sensitivity studies, no specific global RMAs were identified related to FLEX HEPs. If FLEX actions are identified as important during a certain plant configuration based on the Real-Time Risk tool (PARAGON), configuration-specific RMA candidates would be identified.

In general, determination of RMAs involves the use of both qualitative and quantitative considerations for the specific plant configuration and the practical means available to manage risk. The scope and number of RMAs developed and implemented are reached in a graded manner.

Exelon Risk Management procedures contain guidance for development of RMAs in support of the RICT program. Development of RMAs considers those developed for other processes, such as the RMAs developed under the 10 CFR 50.65(a)(4) program and the protected equipment program.

Additionally, Common Cause RMAs are developed to address the potential impact of common cause failures.

RMAs are identified based on the configuration-specific risk. There are three categories of RICT RMAs:

1) Actions to increase risk awareness and control, such as briefing of crews on risk important operator actions and procedures.

2) Actions to reduce the duration of maintenance activities, such as performing activities around the clock.

3) Actions to minimize the magnitude of the risk increase, such as protecting risk important equipment or minimizing fire risk in risk important rooms.

General RMAs are developed for input into the site-specific RICT system guidelines. These guidelines are developed using a graded approach.

Consideration is given for system functionality. These RMAs include:

• Consideration of rescheduling maintenance to reduce risk.

• Discussion of RICT in pre-job briefs.

• Consideration of proactive return-to-service of other equipment.

• Efficient execution of maintenance.

In addition to the RMAs developed qualitatively for the system guidelines, RMAs are developed based on the Real-Time Risk tool to identify configuration-specific RMA candidates to manage the risk associated with internal events, internal flooding, and fire events. These actions include:

• Identification of important equipment or trains for protection.

• Identification of important Operator Actions for briefings.

• Identification of key fire initiators and fire zones for RMAs in accordance with the site Fire RMA process.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 13 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

• Identification of dominant initiating events and actions to minimize potential for initiators.

• Consideration of insights from PRA model cutsets, through comparison of importances.

Common cause RMAs are also developed to ensure availability of redundant structures, systems and components (SSCs), to ensure availability of diverse or alternate systems, to reduce the likelihood of initiating events that require operation of the out-of-service components, and to prepare plant personnel to respond to additional failures. Common cause RMAs are developed by considering the impact of loss of function for the affected SSCs. Examples of common cause RMAs include:

• Performance of non-intrusive inspections on alternate trains.

• Confidence runs performed for standby SSCs.

• Increased monitoring for running components.

• Expansion of monitoring for running components.

• Deferring maintenance and testing activities that could generate an initiating event which would require operation of potentially affected SSCs.

• Readiness of operators and maintenance to respond to additional failures.

• Shift briefs or standing orders which focus on initiating event response or loss of potentially affected SSCs.

iii. Alternatively to item (b)ii above, provide information associated with the following items listed in supporting requirements (SR) HR-G3 and HR-G7 of the PRA standard to support the NRC staffs detailed review of the LAR:

1. the level and frequency of training that the operators and non-operators receive for deployment of the FLEX equipment (performance shaping factor (a) in SR HR-G3),

Response

Because a response to audit question 5.b.ii is provided above, no response to this question is required.

2. performance shaping factor (f) in SR HR-G3 regarding estimates of time available and time required to execute the response,

Response

Because a response to audit question 5.b.ii is provided above, no response to this question is required.

3. performance shaping factor (g) in SR HR-G3 regarding complexity of detection, diagnosis, and decisionmaking and executing the required response,

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 14 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

Response

Because a response to audit question 5.b.ii is provided above, no response to this question is required.

4. performance shaping factor (h) in SR HR-G3 regarding consideration of environmental conditions, and

Response

Because a response to audit question 5.b.ii is provided above, no response to this question is required.

5. human action dependencies as listed in SR HR-G7 of the PRA standard.

Response

Because a response to audit question 5.b.ii is provided above, no response to this question is required.

c) The PRA standard defines PRA upgrade as the incorporation into a PRA model of a new methodology or significant changes in scope or capability that impact the significant accident sequences or the significant accident progression sequences.

Section 1-5 of Part 1 of the 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.

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

Response

Incorporation of FLEX into the PBAPS PRA model is a reflection of plant modifications and procedure changes. Updating the model to reflect such a change is necessary to maintain the model as representative of the as-built, as-operated plant. Accident sequences progress in the same manner as before, except there is the possibility of extended time for power to be available and alternate injection sources. Risk estimation capability is not changed, all FLEX system implementations were made utilizing the existing PRA methodology.

The model changes associated with incorporating FLEX mitigating strategies and their disposition regarding (1) new methodology, (2) change in scope and (3) change in capability are noted in Table 5.c.i-1 below. The term "new method" used in this disposition is consistent with Table A-1 of RG 1.200, Rev.

2.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 15 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 The Scope attribute is defined consistent with Section C of RG 1.200[1], i.e., "The scope of the PRA ...is defined in terms of (1) the metrics used to characterize risk, (2) the plant operating states for which the risk is to be evaluated, and (3) the causes of initiating events (hazard groups) that can potentially challenge and disrupt the normal operation of the plant and, if not prevented or mitigated, would eventually result in core damage and/or a large release."

Consistent with concepts in RG 1.200, Rev. 2 as well as the basis for Capability Category distinctions in the PRA Standard, the term capability used in this disposition is defined in terms of degree of analysis detail and plant-specific realism. Implementation of this criterion in the context of determining whether a specific PRA change represents an upgrade is whether the change would increase the Capability Category (from Not Met or CC-I to CC-II) for one or more SRs.

Table 5.c.i-1: Summary of FLEX Model Changes in PBAPS PRA Change Significant New in Change in Impact on PRA Model Change Method(1) Scope(2) Capability(3) Sequences(4) Comment Creation of FLEX No No No --- The FLEX system fault tree logic system fault tree logic modeling was created in a manner analogous to the already existing fault trees. Failure rate values for new FLEX equipment used in the PRA model is assumed to be twice the corresponding generic equipment failure probability of the similar non-FLEX equipment.

No new methods were employed.

The scope of the model remains identical and no change in the capability categories for any supporting requirement apply.

Creation of FLEX No No No --- All the human error probabilities for HFEs FLEX components were evaluated with the same methodology used for all human error probabilities in the PBAPS PRA models as documented in the PBAPS HRA notebook.

No new methods were employed.

The scope of the model remains identical and no change in the capability categories for any supporting requirement apply.

1 REGULATORY GUIDE 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, March 2009, Rev. 2.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 16 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 5.c.i-1: Summary of FLEX Model Changes in PBAPS PRA Change Significant New in Change in Impact on PRA Model Change Method(1) Scope(2) Capability(3) Sequences(4) Comment Edits to Alternate No No No --- This is an edit to an existing Fault Injection and AC Trees.

power fault trees to implement FLEX No new methods were employed.

capability. The scope of the model remains identical and no change in the capability categories for any supporting requirement apply.

1. New Method: Consistent with Table A-1 of RG 1.200, Rev. 2, the term "new method" refers to an analysis method (i.e., not documentation method) that is new to the subject PRA even if the method itself is not new and has been applied in other PRAs. This term also encompasses newly developed methods in the industry that have been implemented in the base PRA in question.

2. Change in Scope: Consistent with Section C of RG 1.200, Rev. 2, the term PRA scope is defined in terms of the following three attributes: (1) the metrics used to characterize risk, (2) the plant operating states for which the risk is to be evaluated, and (3) the causes of initiating events (hazard groups) that can potentially challenge and disrupt the normal operation of the plant and, if not prevented or mitigated, would eventually result in core damage and/or a large release.

3. Change in Capability: Consistent with concepts in RG 1.200, Rev. 2 as well as the basis for Capability Category distinctions in the PRA Standard, this term is defined in terms of degree of analysis detail and plant-specific realism.

Implementation of this criterion in the context of determining whether a specific PRA change represents an upgrade is whether the change would increase the Capability Category (from Not Met or CC-I to CC-II) for one or more supporting requirements (SRs).

4. Impact on Significant Accident Sequences or Significant Accident Progression Sequences: This term encompasses both Level 1 (core damage) and Level 2 (post-core damage) accident sequences. This criterion is interpreted in this context of "PRA Upgrade" as the top 95% of sequences and whether the makeup of those sequences have been significantly impacted. Whether the makeup of the top 95% of the sequences is determined to be significantly impacted is based on a qualitative consideration as to whether the change in the sequences would likely change decision making when applying the PRA in risk applications. For example, top sequences in the top 95% that for the model change drop out of the top 95% would be a case where justification should be provided as to why the change in question is not considered an upgrade or it should be identified as an upgrade. NOTE: Per the ASME PRA Standard Addenda A and RG 1.200, Rev.

2 definition of PRA upgrade, this criterion is logically ANDed with the other criteria of first having to be a change in scope or a change in capability.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 17 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 ii. Alternatively to item (c)i above, confirm that the modeling of FLEX equipment and FLEX actions in the PRA has been peer reviewed in accordance with NRC-accepted methods. Provide the findings of the peer review performed on the FLEX modeling and the disposition of the findings as they pertain to the impact on this LAR.

Response

Because a response to audit question 5.c.i is provided above, no response to this question is required.

APLA 05 References

1. PB-MISC-043, Assessment of Key Assumptions and Sources of Uncertainty for the Peach Bottom Atomic Power Station PRA, Revision 1, April 2020.

APLA QUESTION 06 - Probabilistic Risk Assessment Modeling and Uncertainty of Digital Instrumentation and Controls Section 2.3.4 of NEI 06-09 states that PRA modeling uncertainties be considered in application of the PRA base model results to the RICT program. The NRC SE for NEI 06-09 states that this consideration is consistent with Section 2.3.5 of RG 1.177, Revision 1, An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications (ADAMS Accession No. ML100910008). NEI 06-09 further states that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties that could potentially impact the results of an RICT calculation and that sensitivity studies should be used to develop appropriate compensatory RMAs.

a) A TS LCO condition listed in LAR Table E1-1 indicates that instrumentation and control (I&C) modeling in the PRA is insufficient to model the condition, and therefore, the inoperability of the associated equipment (e.g., channel) will be modeled using a surrogate event. Furthermore, based on documentation in the LAR for other TS LCO conditions in the RICT program, it is not clear to NRC staff whether I&C is modeled in sufficient detail to support implementation of TSTF-505, Revision 2.

Describe how I&C equipment that is applicable or that impacts the RICT calculations is modeled/considered in the PRA. Include in this discussion: (1) the scope of the I&C equipment that is explicitly modeled (e.g., bistables, relays, sensors, integrated circuit cards), (2) description of the level of detail that the PRA model supports (e.g., are all channels of an actuation circuit modeled), (3) discussion of the generic data and plant-specific data used, and (4) discussion of the associated TS functions for which an RICT can be applied.

Response

The I&C Equipment Functions applicable to the I&C TS LCO conditions listed in LAR Table E1-1 are presented in Table 6.a-1 below. The Table correlates the I&C equipment TS Function to a discussion I&C equipment modeling in the PRA and a discussion of the level of detail supported by the PRA.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 18 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 The PRA-modeled I&C components use the generic data from NUREG/CR-6928 to capture the industry-average performance for components and initiating events at U.S.

commercial nuclear power plants. Plant-specific data updates include component failure probability increases as modified by Surveillance Frequency Control Program (SFCP) interval extension evaluations, and from plant specific component failure data, and average yearly system maintenance.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 19 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

3.3.1.1 3.3.1.1-1 RPS I&C Functions found in Table 3.3.1.1-1 are not Yes RPS Channel Logic is modeled, This I&C function is NOT modeled to include all required channel inputs included as a detailed listing. however individual RPS signals are within the scope of the T.S. The impact of the loss of channel input within the not modeled scope of the applicable T.S. can be "upmapped" to an RPS channel level .

CPH--1A1DWI2 - RPS LOGIC CHANNEL 1A1 OR RELAYS FAIL CPH--1A2DWI2 - RPS LOGIC CHANNEL 1A2 OR RELAYS FAIL CPH--1B1DWI2 - RPS LOGIC CHANNEL 1B1 OR RELAYS FAIL CPH--1B2DWI2 - RPS LOGIC CHANNEL 1B2 OR RELAYS FAIL 3.3.2.2 N/A feed water /main turbine Two channels per trip system of the Digital Yes Digital Feedwater Computers, This I&C equipment function is modeled consistent with T.S. scope for high level trip Feedwater Control System (DFCS) high water reactor level instrumentation required channels. The impact of the loss of channel within the scope of the level trip instrumentation Function shall be applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

OPERABLE. FLC-DCCXHWI2- DIGITAL FEEDWATER COMPUTER DCC-X FAILURE FLC-DCCYHWI2- DIGITAL FEEDWATER COMPUTER DCC-Y FAILURE NTL-072CDWI2- LEVEL TRANSMITTER LT-2-3-72C FAILS TO OPERATE NTL-072DDWI2- LEVEL TRANSMITTER LT-2-3-72D FAILS TO OPERATE 3.3.4.1 N/A ARI Two channels per trip system for each ATWS- Yes The I&C modeling for ARI ATWS- This I&C equipment function is modeled consistent with T.S. scope for RPT instrumentation Function listed below RPT Function includes the Power, required channels. The impact of the loss of channel within the scope of the shall be OPERABLE: channel Logic, Relays, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

a. Reactor Vessel Water Level Low Low Transmitters, switches NTP-404AHWI2 and NTL-072ADWI2 (Level 2); and NTP-404BHWI2 and NTL-072BDWI2

b. Reactor Pressure High. NTP-404CHWI2 and NTL-072CDWI2 NTP-404DHWI2 and NTL-072DDWI2 3.3.4.2 N/A ARI a. Two channels per trip system for each EOC- No I&C Equipment Not Modeled Not explicitly modeled in the PRA.

RPT instrumentation Function listed below consistent with the Function ATWS RPT will be used as a surrogate shall be OPERABLE:

1. Turbine Stop Valve (TSV)Closure; and

2. Turbine Control Valve (TCV) Fast Closure, Trip Oil PressureLow.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 20 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

3.3.5.1 3.3.5.1-1 1- Core Spray System a- Reactor Vessel Water Level Low Low Low Yes The I&C modeling for LPCS system This I&C equipment function is modeled consistent with T.S. scope for (Level 1) Low Low Low Level initiation required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, LRE---7ADWI2- CORE SPRAY RELAY 3-14A-K7A FAILS Transmitters, switches LRE---7BDWI2- CORE SPRAY RELAY 3-14A-K7B FAILS LRE---8ADWI2- CORE SPRAY RELAY 2-14A-K8A FAILS LRE---8BDWI2- CORE SPRAY RELAY 2-14A-K8B FAILS b- Drywell Pressure High Yes The I&C modeling for LPCS system This I&C equipment function is modeled consistent with T.S. scope for DW Press. initiation Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NRE-158ADWI2- RELAY 10A-158A FAILS (HI DWP)

NRE-158BDWI2- RELAY 10A-158B FAILS (HI DWP)

NRE-158CDWI2- RELAY 10A-158C FAILS (HI DWP)

NRE-158DDWI2- RELAY 10A-158D FAILS (HI DWP) c- Reactor PressureLow (Injection Yes The I&C modeling for LPCS system This I&C equipment function is modeled consistent with T.S. scope for Permissive) RX Press. Initiation and injection required channels. The impact of the loss of channel within the scope of the permissive Function includes the applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Power, Division Logic, Relays, LRE---9ADWI2- CORE SPRAY RELAY 2-14A-K9A FAILS Transmitters, switches LRE---9BDWI2- CORE SPRAY RELAY 2-14A-K9B FAILS LRE--23ADWI2- CORE SPRAY RELAY 2-14A-K23A FAILS LRE--23BDWI2- CORE SPRAY RELAY 2-14A-K23B FAILS d- Core Spray Pump Discharge Flow Low Yes The I&C modeling for LPCS system This I&C equipment function is NOT modeled to include all required channels (Bypass) Pump Discharge Min Flow within the scope of the T.S. The impact of the loss of channel within the operation Function includes the scope of the applicable T.S. can be represented by the LPCS pump associated Power and switches with each signal.

LPS-081ADWI2 - DIFFERENTIAL PRESS.SWITCH 14-081A FAILS TO ACTUATE LPS-081BDWI2 - DIFFERENTIAL PRESS.SWITCH 14-081B FAILS TO ACTUATE LPS-081CDWI2 - DIFFERENTIAL PRESS.SWITCH 14-081C FAILS TO ACTUATE LPS-081DDWI2 - DIFFERENTIAL PRESS.SWITCH 14-081D FAILS TO ACTUATE

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 21 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

e- Core Spray Pump Start-Time Delay Relay Yes The I&C modeling for LPCS system This I&C equipment function is NOT modeled to include all required channels (loss of offsite power) Time Delay Relay start Function within the scope of the T.S. The impact of the loss of channel within the includes load sequencing relays. scope of the applicable T.S. can be represented by the LPCS pump associated The specific TD relays are not with each signal.

explicitly modeled ESQ14415DWI2 - LOAD SEQUENCE PROCESS FOR E12 BUS FAILS (2-144X-15 RELAY)

ESQ14416DWI2 - LOAD SEQUENCE PROCESS FOR E22 BUS FAILS (2-144X-16 RELAY)

ESQ14417DWI2 - LOAD SEQUENCE PROCESS FOR E32 BUS FAILS (2-144X-17 RELAY)

ESQ14418DWI2 - LOAD SEQUENCE PROCESS FOR E42 BUS FAILS (2-144X-18 RELAY) f- Core Spray Pump Start-Time Delay Relay Yes The I&C modeling for LPCS system This I&C equipment function is NOT modeled to include all required channels (offsite power available) Time Delay Relay start Function within the scope of the T.S. The impact of the loss of channel within the Pumps A,C includes load sequencing relays. scope of the applicable T.S. can be represented by the LPCS pump associated Pumps B,D The specific TD relays are not with each signal.

explicitly modeled ESQ14415DWI2 - LOAD SEQUENCE PROCESS FOR E12 BUS FAILS (2-144X-15 RELAY)

ESQ14416DWI2 - LOAD SEQUENCE PROCESS FOR E22 BUS FAILS (2-144X-16 RELAY)

ESQ14417DWI2 - LOAD SEQUENCE PROCESS FOR E32 BUS FAILS (2-144X-17 RELAY)

ESQ14418DWI2 - LOAD SEQUENCE PROCESS FOR E42 BUS FAILS (2-144X-18 RELAY) 2- Low Pressure Coolant a- Reactor Vessel Water Level Low Low Low Yes The I&C modeling for LPCI system This I&C equipment function is modeled consistent with T.S. scope for Injection System (Level 1) Low Low Low Level initiation required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, NTP-404AHWI2- PRESSURE TRANSMITERPT-404A FAILS TO OPERATE Transmitters, switches NTP-404BHWI2- PRESSURE TRANSMITERPT-404B FAILS TO OPERATE NTP-404CHWI2- PRESSURE TRANSMITERPT-404C FAILS TO OPERATE NTP-404DHWI2- PRESSURE TRANSMITERPT-404D FAILS TO OPERATE

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 22 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

b- Drywell Pressure High Yes The I&C modeling for LPCI system This I&C equipment function is modeled consistent with T.S. scope for DW Press. initiation Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NRE-158ADWI2- RELAY 10A-158A FAILS (HI DWP)

NRE-158BDWI2- RELAY 10A-158B FAILS (HI DWP)

NRE-158CDWI2- RELAY 10A-158C FAILS (HI DWP)

NRE-158DDWI2- RELAY 10A-158D FAILS (HI DWP) c- Reactor PressureLow (Injection Yes The I&C modeling for LPCI system This I&C equipment function is modeled consistent with T.S. scope for Permissive) RX Press. Initiation and injection required channels. The impact of the loss of channel within the scope of the permissive Function includes the applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Power, Division Logic, Relays, LRE---9ADWI2- CORE SPRAY RELAY 2-14A-K9A FAILS Transmitters, switches LRE---9BDWI2- CORE SPRAY RELAY 2-14A-K9B FAILS LRE--23ADWI2- CORE SPRAY RELAY 2-14A-K23A FAILS LRE--23BDWI2- CORE SPRAY RELAY 2-14A-K23B FAILS d- Reactor Pressure Low Low (Recirculation Yes The I&C modeling for LPCI system This I&C equipment function is modeled consistent with T.S. scope for Discharge Valve Permissive) RX Press. Recirc valve isolation required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, DRE-K90ADWI2 Transmitters, switches DRE-K90BDWI2 DRE-K101ADWI2 DRE-K101BDWI2 e- Reactor Vessel Shroud Level Level 0 Yes The I&C modeling for LPCI system This I&C equipment function is modeled consistent with T.S. scope for RX Level Containment Cooling required channels. The impact of the loss of channel within the scope of the permissive Function includes the applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Power, Division Logic, Relays, NTP-404AHWI2 and NTL-072ADWI2 Transmitters, switches NTP-404BHWI2 and NTL-072BDWI2

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 23 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

f- Low Pressure Coolant Injection Pump Start Yes The I&C modeling for LPCI system This I&C equipment function is NOT modeled to include all required channels Time Delay Relay (offsite power available) Time Delay Relay start Function within the scope of the T.S. The impact of the loss of channel within the Pumps A,C includes load sequencing relays. scope of the applicable T.S. can be represented by LPCI pump associated with Pumps B,D The specific TD relays are not each signal.

explicitly modeled ESQ14415DWI2 - LOAD SEQUENCE PROCESS FOR E12 BUS FAILS (2-144X-15 RELAY)

ESQ14416DWI2 - LOAD SEQUENCE PROCESS FOR E22 BUS FAILS (2-144X-16 RELAY)

ESQ14417DWI2 - LOAD SEQUENCE PROCESS FOR E32 BUS FAILS (2-144X-17 RELAY)

ESQ14418DWI2 - LOAD SEQUENCE PROCESS FOR E42 BUS FAILS (2-144X-18 RELAY) g- Low Pressure Coolant Injection Pump Yes I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels Discharge Flow Low (Bypass) this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the however the Min flow valves are scope of the applicable T.S. can be represented by LPCI pump associated with modeled each signal.

DMV--16ADPI2 - MOTOR VALVE 10-16A FAILS TO OPEN DMV--16BDPI2 - MOTOR VALVE 10-16B FAILS TO OPEN DMV--16CDPI2 - MOTOR VALVE 10-16C FAILS TO OPEN DMV--16DDPI2 - MOTOR VALVE 10-16D FAILS TO OPEN 3- High Pressure Coolant a- Reactor Vessel Water Level Low Low Yes The I&C modeling for HPCI system This I&C equipment function is modeled consistent with T.S. scope for Injection System (Level 2) Low Low Level initiation Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NTL-072ADWI2- LEVEL TRANSMITTER LT-2-3-72A FAILS TO OPERATE NTL-072BDWI2- LEVEL TRANSMITTER LT-2-3-72B FAILS TO OPERATE NTL-072CDWI2- LEVEL TRANSMITTER LT-2-3-72C FAILS TO OPERATE NTL-072DDWI2- LEVEL TRANSMITTER LT-2-3-72D FAILS TO OPERATE

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 24 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

b- Drywell Pressure High Yes The I&C modeling for HPCI system This I&C equipment function is modeled consistent with T.S. scope for DW Press. initiation Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NRE-158ADWI2- RELAY 10A-158A FAILS (HI DWP)

NRE-158BDWI2- RELAY 10A-158B FAILS (HI DWP)

NRE-158CDWI2- RELAY 10A-158C FAILS (HI DWP)

NRE-158DDWI2- RELAY 10A-158D FAILS (HI DWP) c- Reactor Vessel Water Level High (Level 8) Yes The I&C modeling for HPCI system This I&C equipment function is modeled consistent with T.S. scope for RX Level High Trip Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches While the Associated Channels appear in the PRA model, the HPCI Divisional Logic can be used to represent this TS Function HLC--DV1HWI2- HPCI DIV I RELAY LOGIC FAILS HLC--DV2HWI2- HPCI DIV II RELAY LOGIC FAILS d- Condensate Storage Tank Level Low Yes The I&C modeling for HPCI system This I&C equipment function is modeled consistent with T.S. scope for Suction Source Transfer Function required channels. The impact of the loss of channel within the scope of the includes the Power, Relays, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Transmitters, switches HLV---74DWI2- FAILURE OF LEVEL SWITCH LS-74 HLV---75DWI2- FAILURE OF LEVEL SWITCH LS-75 e- Suppression Pool Water Level High Yes The I&C modeling for HPCI system This I&C equipment function is modeled consistent with T.S. scope for Suction Source Transfer Function required channels. The impact of the loss of channel within the scope of the includes the Power, Relay, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Transmitters, switches HLV--91ADWI2- FAILURE OF LEVEL SWITCH LS-91A HLV--91BDWI2- FAILURE OF LEVEL SWITCH LS-91B f- High Pressure Coolant Injection Pump Yes I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels Discharge Flow Low (Bypass) this Function is Not Modeled; within the scope of the T.S. The impact of the loss of channel within the however, the Min flow valve is scope of the applicable T.S. can be represented by the HPCI pump.

modeled HMV---25DPI2 - MOTOR OPERATED VLV.23-25 FAILS TO OPEN HMV---25SPI2 - MIN-FLOW VLV 23-25 FAILS TO OPEN STANDBY FAILURE

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 25 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

4- Automatic a- Reactor Vessel Water Level Low Low Low Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for Depressurization System (Level 1) system Low Level Function required channels. The impact of the loss of channel within the scope of the Trip System A includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NTL-072ADWI2- LEVEL TRANSMITTER LT-2-3-72A FAILS TO OPERATE NTL-072CDWI2- LEVEL TRANSMITTER LT-2-3-72C FAILS TO OPERATE b- Drywell Pressure High Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for system DW Pressure Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NRE-158ADWI2- RELAY 10A-158A FAILS (HI DWP)

NRE-158CDWI2- RELAY 10A-158C FAILS (HI DWP) c- Automatic Depressurization System Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for Initiation Timer system lnitiation timer Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches ART---K4DWI2- 105 SECOND TIME DELAY RELAY 2E-K4 FAILS TO OPERATE d- Reactor Vessel Water Level-Low Low Low Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for (Level 1), (Permissive) system Low Level permissive required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, NTL-072ADWI2- LEVEL TRANSMITTER LT-2-3-72A FAILS TO OPERATE Transmitters, switches NTL-072CDWI2- LEVEL TRANSMITTER LT-2-3-72C FAILS TO OPERATE e- Reactor Vessel Water Confirmatory Level Yes The I&C modeling for the ADS This I&C equipment function is modeled consistent with T.S. scope for Low (Level 4) system Low Level confirmatory required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, NRE-152ADWI2- RELAY 10A-152A FAILS (LO RX LEVEL)

Transmitters, switches f- Core Spray Pump Discharge Pressure High Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for system LPCS Pump running required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, APS-044ADWI2- LPCS PUMP A PRESSURE SWITCH NOT OPERATING Transmitters, switches APS-044BDWI2- LPCS PUMP B PRESSURE SWITCH NOT OPERATING APS-044CDWI2- LPCS PUMP C PRESSURE SWITCH NOT OPERATING APS-044DDWI2- LPCS PUMP D PRESSURE SWITCH NOT OPERATING

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 26 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

g- Low Pressure Coolant Injection Pump Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for Discharge Pressure High system LPCI Pump running required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, APS-120ADWI2- LPCI PUMP A PRESS. SWITCH 10-120A FAILS Transmitters, switches APS-120BDWI2- - LPCI PUMP A PRESS. SWITCH 10-120B FAILS APS-120CDWI2- - LPCI PUMP A PRESS. SWITCH 10-120C FAILS APS-120DDWI2- - LPCI PUMP A PRESS. SWITCH 10-120D FAILS APS-120EDWI2- - LPCI PUMP A PRESS. SWITCH 10-120E FAILS APS-120FDWI2- - LPCI PUMP A PRESS. SWITCH 10-120F FAILS APS-120GDWI2- - LPCI PUMP A PRESS. SWITCH 10-120G FAILS APS-120HDWI2- - LPCI PUMP A PRESS. SWITCH 10-120H FAILS h- Automatic Depressurization System Low Yes The I&C modeling for the ADS A This I&C equipment function is modeled consistent with T.S. scope for Water Level Actuation Timer system Low Level timer Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches ART-K33ADWI2- 9 MINUTE TIME DELAYRELAY 2E-K33A FAILS TO OPERATE 5- Automatic a- Reactor Vessel Water Level Low Low Low Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for Depressurization System (Level 1) system Low Level Function required channels. The impact of the loss of channel within the scope of the Trip System B includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NTL-072BDWI2- LEVEL TRANSMITTER LT-2-3-72B FAILS TO OPERATE NTL-072DDWI2- LEVEL TRANSMITTER LT-2-3-72D FAILS TO OPERATE b- Drywell Pressure High Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for system DW Pressure Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NRE-158BDWI2- RELAY 10A-158B FAILS (HI DWP)

NRE-158DDWI2- RELAY 10A-158D FAILS (HI DWP) c- Automatic Depressurization System Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for Initiation Timer system l timer Function includes required channels. The impact of the loss of channel within the scope of the the Power, Division Logic, Relays, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Transmitters, switches ART--K11DWI2- 105 SECOND TIME DELAY RELAY 2E-K11 FAILS TO OPERATE

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 27 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

d- Reactor Vessel Water Level-Low Low Low Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for (Level 1), (Permissive) system Low Level permissive required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, NTL-072BDWI2- LEVEL TRANSMITTER LT-2-3-72B FAILS TO OPERATE Transmitters, switches NTL-072DDWI2- LEVEL TRANSMITTER LT-2-3-72D FAILS TO OPERATE e- Reactor Vessel Water Confirmatory Level Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for Low (Level 4) system Low Level confirmatory required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, NRE-152BDWI2- RELAY 10A-152B FAILS (LO RX LEVEL)

Transmitters, switches f- Core Spray Pump Discharge Pressure High Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for system LPCS Pump running required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, APS-044ADWI2- LPCS PUMP A PRESSURE SWITCH NOT OPERATING Transmitters, switches APS-044BDWI2- LPCS PUMP B PRESSURE SWITCH NOT OPERATING APS-044CDWI2- LPCS PUMP C PRESSURE SWITCH NOT OPERATING APS-044DDWI2- LPCS PUMP D PRESSURE SWITCH NOT OPERATING g- Low Pressure Coolant Injection Pump Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for Discharge Pressure High system LPCI Pump running required channels. The impact of the loss of channel within the scope of the Function includes the Power, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Division Logic, Relays, APS-120ADWI2 - LPCI PUMP A PRESS. SWITCH 10-120A FAILS Transmitters, switches APS-120BDWI2 - LPCI PUMP A PRESS. SWITCH 10-120B FAILS APS-120CDWI2 - LPCI PUMP A PRESS. SWITCH 10-120C FAILS APS-120DDWI2 - LPCI PUMP A PRESS. SWITCH 10-120D FAILS APS-120EDWI2 - LPCI PUMP A PRESS. SWITCH 10-120E FAILS APS-120FDWI2 - LPCI PUMP A PRESS. SWITCH 10-120F FAILS APS-120GDWI2 - LPCI PUMP A PRESS. SWITCH 10-120G FAILS APS-120HDWI2 - LPCI PUMP A PRESS. SWITCH 10-120H FAILS h- Automatic Depressurization System Low Yes The I&C modeling for the ADS B This I&C equipment function is modeled consistent with T.S. scope for Water Level Actuation Timer system Low Level timer Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches ART-K33BDWI2- 9 MINUTE TIME DELAYRELAY 2E-K33B FAILS TO OPERATE

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 28 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

3.3.5.2 3.3.5.2-1 Reactor Core Isolation 1- Reactor Vessel Water Level Low Low Yes The I&C modeling for RCIC system This I&C equipment function is modeled consistent with T.S. scope for Cooling System (Level 2) Low Low Level Initiation Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches NTL-072ADWI2- LEVEL TRANSMITTER LT-2-3-72A FAILS TO OPERATE NTL-072BDWI2- LEVEL TRANSMITTER LT-2-3-72B FAILS TO OPERATE NTL-072CDWI2- LEVEL TRANSMITTER LT-2-3-72C FAILS TO OPERATE NTL-072DDWI2- LEVEL TRANSMITTER LT-2-3-72D FAILS TO OPERATE 2- Reactor Vessel Water Level High (Level 8) Yes The I&C modeling for RCIC system This I&C equipment function is modeled consistent with T.S. scope for RX Level High trip Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches While the Associated Channels appear in the PRA model, the RCIC Divisional Logic can be used to represent this TS Function RLC--DV1HWI2- RCIC DIV I RELAY LOGIC FAILS RLC--DV2HWI2- RCIC DIV I RELAY LOGIC FAILS 3- Condensate Storage Tank Level Low Yes The I&C modeling for RCIC system This I&C equipment function is modeled consistent with T.S. scope for Suction Source Transfer Function required channels. The impact of the loss of channel within the scope of the includes the Power, Division Logic, applicable T.S. can be directly evaluated in the RTR tool for the RICT Program.

Relays, Transmitters, switches RUT--170DWI2- CST LEVEL TRIP UNIT 13- 170 FAILS RUT--171DWI2- CST LEVEL TRIP UNIT 13- 171 FAILS 3.3.6.1 3.3.6.1-1 1- Main Steam Line a- Reactor Vessel Water Level Low Low Low Yes Not all I&C Equipment Associated This I&C equipment function is NOT modeled to include all required channels Isolation (Level 1) with this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the b- Main Steam Line Pressure Low No scope of the applicable T.S. can be represented by MSIV Break-Outside-Containment initiating events and MSIV Failure to Close Basic Events to represent the failure of the I&C function within the scope of the T.S. LCO.

c- Main Steam Line Flow High No %VMSL- V SEQUENCE THRU MAIN STEAM LINES FAV--80ADNI2- MSIV 80A FAILS TO CLOSE ON DEMAND d- Deleted No FAV--86ADNI2- MSIV 86A FAILS TO REMAIN CLOSED e- Turbine Building Main Steam Tunnel No Temperature High f- Reactor Building Main Steam Tunnel No Temperature-High

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 29 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

2- Primary Containment a- Reactor Vessel Water Level Low (Level 3) Yes Not all I&C Equipment Associated This I&C equipment function is NOT modeled to include all required channels Isolation with this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the scope of the applicable T.S. can be represented by Containment Valves b- Drywell PressureHigh Yes Failure to close basic events to represent the I&C function within the scope of the T.S. LCO.

VAV-2507HOI2- AO-2507 FAILS TO REMAIN CLOSED (NCFO) c- Main Stack Monitor Radiation High No VAV-2506HOI2- AO-2506 FAILS TO REMAIN CLOSED (NCFO) 2ISPH-LG-LEAKF-- LARGE PRE-EXISTING FAILURE (ESTIMATED BY PNL d- Reactor Building Ventilation Exhaust No Radiation High e- Refueling Floor Ventilation Exhaust No Radiation High 3- High Pressure Coolant a- HPCI Steam Line FlowHigh No I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels Injection System Isolation this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the scope of the applicable T.S. can be represented by HPCI Break-Outside-b- HPCI Steam Line FlowTime Delay Relays No Containment initiating events to represent the failure of the I&C function within the scope of the T.S. LCO.

%VHPCI- V SEQUENCE THRU HPCI STEAM LINES c- HPCI Steam Supply Line PressureLow No

%VHPCI-W- V SEQUENCE THRU HPCI STEAM LINES (PUMP DISCHARGE) d- Drywell PressureHigh (Vacuum Breakers) No e- HPCI Compartment and Steam Line Area No TemperatureHigh 4- Reactor Core Isolation a- RCIC Steam Line FlowHigh No I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels Cooling System Isolation this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the scope of the applicable T.S. can be represented by RCIC Break-Outside-b- RCIC Steam Line FlowTime Delay Relays No Containment initiating events to represent the failure of the I&C function within the scope of the T.S. LCO.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 30 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

c- RCIC Steam Supply Line PressureLow No %VRCIC- V SEQUENCE THRU RCIC STEAM LINES

%VRCIC-W- V SEQUENCE THRU RCIC STEAM LINES (PUMP DISCHARGE) d- Drywell PressureHigh (Vacuum Breakers) No e- RCIC Compartment and Steam Line Area No TemperatureHigh 5- Reactor Water Cleanup a- RWCU Flow-High No I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels System Isolation this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the scope of the applicable T.S. can be represented by RWCU Break-Outside-b- SLC System Initiation Yes Containment initiating events and Fails Open Basic events (for RWCU isolation following SLC initiation) to represent the failure of the I&C function within the scope of the T.S. LCO.

c- Reactor Vessel Water Level-Low (Level 3) No %VRWCU- V SEQUENCE THRU RWCU LINES SMV---15DNI2- MOTOR OPERATED VLV.RWCU 2-12-015 NO - FAILS OPEN SMV---18DNI2- MOTOR OPERATED VLV.RWCU 2-12-018 NO - FAILS OPEN SMV---68DNI2- MOTOR OPERATED VLV.RWCU 2-12-068 NO - FAILS OPEN 6- RHR Shutdown Cooling a- Reactor Pressure-High No I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels System Isolation this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the scope of the applicable T.S. can be represented by RHR SDC Break-Outside-Containment initiating events to represent the failure of the I&C function b- Reactor Vessel Water Level-Low (Level 3) No within the scope of the T.S. LCO.

%VRHR- V SEQUENCE THRU RHR SUCTION LINES 7- Feedwater Recirculation a- Reactor Pressure-High No I&C Equipment Associated with This I&C equipment function is NOT modeled to include all required channels Isolation this Function is Not Modeled within the scope of the T.S. The impact of the loss of channel within the scope of the applicable T.S. can be represented by FW Break-Outside-Containment initiating events to represent the failure of the I&C function within the scope of the T.S. LCO.

%VFW- V SEQUENCE THRU FW LINES

%VFW-W- V SEQUENCE THRU FW LINES (PUMP DISCHARGE)

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 31 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.a-1: Overview of I&C Equipment Functions Represented in the PRA Tech Spec Table System Function Function Scope of Modeling Level of detail which Modeling Supports Explicitly Modeled?

8- Traversing Incore Probe a- Reactor Vessel Water Level-Low (Level 3) No I&C Equipment Associated with Not modeled explicitly or Implicitly. Line diameter less than 1" for fluid.

Isolation this Function is Not Modeled SMALL CONTAINMENT FAILURE will be used as a surrogate for TIP isolation as b- Drywell Pressure-High No failure of the TIP to isolate would be most likely to serve as a path for the Containment atmosphere to vent. TIP Isolation is actuated by LOCA signals which also actuate the PCIS Function.

3.3.8.1 3.3.8.1-1 1- 4 kV Emergency Bus a- Bus Undervoltage Yes The I&C modeling for the 4kV This I&C equipment function is NOT modeled to include all required channels Undervoltage (Loss of source transfer and EDG initiation within the scope of the T.S. The impact of the loss of channel within the Voltage) Function includes the Power, scope of the applicable T.S. can be represented by the Undervoltage Relay 2- 4kV Emergency Bus a- Bus Undervoltage Yes Relays, Transmitters, switches Failure to Operate basic events to represent the I&C function within the Undervoltage (Degraded scope of the T.S. LCO.

Voltage Low Setting) b- Time Delay No ESQ12715DWI2- LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 3-127*-

15 FAIL TO OPERATE 3- 4 kV Emergency Bus a- Bus Undervoltage Yes ESQ12715DWI3- LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 2-127*-

Undervoltage (Degraded 15 FAIL TO OPERATE Voltage High Setting) b- Time Delay No ESQ12716DWI2- LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 2-127*-

16 FAIL TO OPERATE 4- 4 kV Emergency Bus a- Bus Undervoltage Yes ESQ12716DWI3-LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 3-127*-

Undervoltage (Degraded 17 FAIL TO OPERATE Voltage LOCA) b- Time Delay No ESQ12717DWI2- LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 2-127*-

17 FAIL TO OPERATE 5- 4 kV Emergency Bus a- Bus Undervoltage Yes ESQ12717DWI3-LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 3-127*-

Undervoltage (Degraded 17 FAIL TO OPERATE Voltage non-LOCA) b- Time Delay No ESQ12718DWI2- LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 2-127*-

18 FAIL TO OPERATE ESQ12718DWI3- LOSS OF VOLTAGE OR DEGRADED VOLTAGE RELAYS 3-127*-

18 FAIL TO OPERATE LPCS - Low Pressure Core Spray LPCI - Low Pressure Coolant Injection ADS - Automatic Depressurization System LOCA - Loss of Coolant Accident SLC - Standby Liquid Control

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 32 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 b) Regarding digital I&C, the NRC staff notes the lack of consensus industry guidance for modeling these systems in plant PRAs to be used to support risk-informed regulatory applications. In addition, known modeling challenges exist such as lack of industry data for digital I&C components, differences between digital and analog system failure modes, and the complexities associated with modeling software failures, including common cause software failures. Also, although reliability data from vendor tests may be available, this source of data is not a substitute for in-the-field operational data.

Given these challenges, the uncertainty associated with modeling a digital I&C system could impact the RICT program.

Attachment 4 of the LAR identifies digital feedwater control system is employed at the plant. However, the modeling of this digital system is not identified in Enclosure 9 as a source of uncertainty. Therefore, it is not clear to the NRC staff whether the digital feedwater control system is the only digital system credited in the PRA and whether there are other digital systems credited in the PRA that could potentially impact RICT calculations. In light of these observations, provide the following information:

i. Describe and provide the results of a sensitivity study performed for each digital system modeled in the PRA demonstrating that the uncertainty associated with PRA modeling the digital I&C systems has inconsequential impact on the RICT calculations.

Response

The Digital Feedwater level control malfunctions could cause a loss of feedwater flow to the RPV. The PRA includes feedwater level control failure modes which could cause a failure of feedwater. This modeling includes failures of the digital feedwater computers and respective power supplies.

For the digital feedwater level control sensitivity, the digital feedwater basic events were increased by a factor of 100 as shown in Table 6.b-1 below. These values apply to both FPIE and FPRA models.

Table 6.b-1: Overview of Digital Feedwater BE Values BE Description Value Base Sens FLC-DCCXHWI2 DIGITAL FEEDWATER COMPUTER 7.20E-05 7.20E-03 DCC-X FAILURE FLC-DCCYHWI2 DIGITAL FEEDWATER COMPUTER 7.20E-05 7.20E-03 DCC-Y FAILURE The results of this sensitivity show that the system has a negligible impact on RICT results and are provided in Table 6.b-2 below.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 33 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Table 6.b-2: Example RICT Calculations Original RICT Sensitivity RICT Tech Spec TS Condition Estimate (days) Estimate (days) 3-5-1-C HPCI system inoperable 29.7 29.7 3-5-1-D HPCI system inoperable and One LP ECCS subsystem 26.9 26.9 inoperable 3-5-3-A RCIC system inoperable 30 30 ii. As an alternative to item (b)i above, identify which LCOs are determined to be impacted by digital I&C system modeling for which RMAs will be applied during an RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation requires additional RMAs.

Response

Because a response to audit question 6.b.i is provided above, no response to this question is required.

APLA QUESTION 07 - PRA Update Process Section 2.3.4 of NEI 06-09 specifies, criteria shall exist in PRA configuration risk management to require PRA model updates concurrent with implementation of facility changes that significantly impact RICT calculations.

LAR Enclosure 7 states that if a plant change or a discovered condition is identified and can have significant impact on the RICT 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. The LAR does not explain under what conditions an unscheduled update of the PRA model will be performed or the criteria defined in the plant procedures that will be used to initiate the update.

In light of these observations, 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, define what is meant by significant impact to the RICT Program calculations.

Response

The Exelon Risk Management FPIE & FPRA Model Update procedures require an evaluation of plant changes or discovered conditions (tracked as Updating Requirement Evaluations (UREs))

against an extensive list of criteria including change in core damage frequency (CDF)/large early release frequency (LERF). A Risk Management Engineer will evaluate each URE to determine whether the model of record (MOR) should be updated expeditiously or the update

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 34 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 can be delayed to the next periodic update. This determination will be made based on whether the PRA model fidelity (representation of the as-built, as operated plant) without the update is adequate to support PRA applications that are currently in effect. This is determined either by qualitative screening or Working model updates for potentially significant changes. Some of the PRA Unscheduled Update Criteria are listed below:

• CDF or LERF change >25%.

• CDF>1E-5.

• LERF>1E-6.

• Significant change in accident class or sequence (greater than factor of 2 increase in an accident class that contributes >5% risk).

• Configuration risk increase factors that could breach the color thresholds used in Maintenance Rule a(4).

These evaluations, particularly the check on significant sequences and configuration risk, ensure changes that could significantly impact RICT calculations initiate an unscheduled PRA model update or result in administrative limits on the RICT program per Exelon procedures (for example, limiting the use of RICT to LCOs where the impact of the condition is not significant).

APLA QUESTION 09 - Impact of Seasonal Variations on the Real-Time Risk Model Regulatory Position 2.3.3 of RG 1.174 states that the level of detail in the PRA should be sufficient to model the impact of the proposed licensing basis change. The characterization of the problem should include establishing a cause-effect relationship to identify portions of the PRA affected by the issue being evaluated. Full-scale applications of the PRA should reflect this cause-effect relationship in a quantification of the impact of the proposed licensing basis change on the PRA elements. Additionally, NEI 06-09 states the following:

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

Section 2 of LAR Enclosure 8 states, The impact of outside temperatures on system requirements are addressed in the RTR model. As part of its audit (ADAMS Accession No. ML20217L346), the NRC staff noted that PRA Notebook PB-MISC-043 states that two EDG fans are required when ambient temperature is above 80 degrees Fahrenheit (°F); however, the PRA model uses a split fraction to represent the percentage of the year assumed to be over 80 °F in modifying the success criteria. The analysis states that RICT will necessitate identifying specific time periods when two fans are required.

Provide further explanation supporting the statement above by summarizing the plant equipment subject to seasonal variations and how it is modeled in the PRA to remove the seasonal dependency.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 35 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

Response

The only seasonal dependency modeled in the PRA is EDG ventilation cooling. When the outside air temperature is greater than 80oF the EDGs require two fans to be operational. When the temperature is below 80oF either fan will provide enough cooling. The PRA contains flags within the PRA model logic to choose the correct success criteria for EDG ventilation based on outside air temperature. PARAGON has been modified to address this seasonal dependency.

In PARAGON, when the outside air temperature is less than 80oF, the Operators will input a schedule code which sets the flags so that the model quantifies requiring only one fan running.

As a default, the PARAGON model conservatively assumes two fans are required until the less than 80oF code is input into PARAGON.

APLA QUESTION 10 - Probabilistic Risk Assessment Model Uncertainty Analysis Process The NRC staff SE to NEI 06-09 specifies that the LAR should identify key assumptions and sources of uncertainty and assess and disposition each as to its impact on the RMTS application. Section 5.3 of NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decisionmaking, Main Report, dated March 2017 (ADAMS Accession No. ML17062A466), presents guidance on the process of identifying, characterizing, and qualitatively screening model uncertainties.

LAR Enclosure 9 states that the process for identifying key assumptions and sources of uncertainty for the IEPRA (includes internal floods) and FPRA was performed using the guidance in NUREG-1855, Revision 1. The LAR indicates that in addition to reviewing generic industry sources of uncertainty for applicability, the IEPRA and FPRA models and notebooks were reviewed for plant-specific assumptions and sources of uncertainty.

However, for the IEPRA (includes internal floods), it is not clear to the NRC staff what specific process and criteria were used to screen uncertainties from an initial comprehensive list of assumptions and sources of PRA modeling uncertainty (including those associated with plant-specific features, modeling choices, and generic industry concerns) in order to conclude that no uncertainty issues could impact the RICT calculations. The NRC staff notes from review of Enclosure 9 of the LAR that the dispositions to many identified sources of uncertainty highlight the phrase not significantly impact the RICT values. It is not clear to the NRC staff what this phrase means in all cases. Also, for certain sources of uncertainty, the disposition states that a sensitivity study was performed to evaluate the impact of the uncertainty, but it is not clear what criteria was used to determine when a sensitivity study was performed or when additional RMAs should be considered.

Therefore, address the following regarding the IEPRA (includes internal floods) uncertainties:

a) Describe the process used to screen uncertainties from the initial comprehensive lists of PRA uncertainties (including those associated with plant-specific features, modeling choices, and generic industry concerns) in order to eventually conclude that the uncertainty issues could not impact the RICT calculations. Include a description of the criteria that was used to screen down from a comprehensive listing of sources of uncertainty to a smaller set of key candidate assumptions and sources of uncertainty.

Also, describe the criteria used to justify that none of the key candidate assumptions and

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 36 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 sources of uncertainty could have an impact on the RICT calculations. As part of this description, explain the criteria used to determine when the results of sensitivity studies do not significantly impact RICT values.

Response

PB-MISC-043 [1] documents the systematic review of PRA sources of uncertainty and assumptions with respect to the TSTF-505 RICT LAR. The PBAPS FPIE and FPRA models and documentation were reviewed for generic (using the applicable lists of EPRI-identified generic sources of uncertainty per EPRI 1016737 and EPRI 1026511) and plant-specific modeling assumptions and related sources of uncertainty. Each PRA model includes an evaluation of the potential sources of uncertainty for the base case models using the approach that is consistent with the ASME/ANS PRA Standard requirements for identification and characterization of uncertainties and assumptions.

The base FPIE PRA model uncertainty discussions are documented in Appendix A of the Summary Notebook (PB-PRA-013) [2] and the base FPRA model uncertainty discussions are documented in the Uncertainty & Sensitivity Analysis Notebook (PB-PRA-021.62) [3]. The results of the base PRA qualitative uncertainty evaluations were systematically reviewed to determine which potential sources of uncertainty could impact the risk-informed applications.

The systematic review of PRA sources of uncertainty used the process outlined in Stage E ("Assessing Model Uncertainty") of NUREG-1855 (as endorsed by Reg. Guide 1.200 Rev. 2 guidance), as summarized below:

• Step E-1: Identify any potential model uncertainties and determine their significance.

o Step E-1.1: Identify sources of model uncertainty and related assumptions Tables A-1 and A-2 of EPRI 1016737 were used to identify potential sources of model uncertainty from the Internal Events and Internal Flooding PRA models.

Appendix B of EPRI 1026511 was used to identify potential sources of model uncertainty from the FPRA model.

Appendix E of EPRI 1026511 was used to identify potential sources of model uncertainty from the Level 2 PRA models used for Internal Events, Internal Flooding, and Fire PRA hazards.

Finally, plant-specific sources of model uncertainty were also considered in the identification process of the uncertainty evaluation.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 37 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 o Step E-1.2: Identify relevant sources of model uncertainty and related assumptions This step allows for screening of potential sources of model uncertainty based on the parts of the model used for the applications. Since the RICT evaluations involve complete model re-quantification for each analyzed case, no specific potential sources of model uncertainty were screened out during the performance of this step.

o Step E-1.3: Characterize sources of model uncertainty and related assumptions Per the guidance in NUREG-1855 and the associated EPRI reports, the characterization process involves identifying:

1) The part of the PRA model affected,

2) The modeling approach or assumptions utilized in the model,

3) The impact on the PRA model, and

4) Representation of conservative bias (if applicable).

These considerations were included in the evaluation of potential sources of model uncertainty o Step E-1.4: Qualitative screening of source of model uncertainty and related assumptions This step allows for screening out potential sources of model uncertainty by referencing consensus model approaches. The evaluation process included identifying the approach utilized (e.g., consensus approach or other applicable guidance), and the level of detailed included in the PRA. These two considerations were used as a means of qualitatively screening potential impacts on the risk-informed applications.

o Step E-1.5: Identify and characterize relevant sources of model uncertainty and related assumptions associated with model changes The implementation of the TSTF-505 RICT program utilizes the base PRA models. As such, no new sources of model uncertainty have been introduced for the application.

b) Concerning the evaluation criteria used to evaluate and screen uncertainties addressed in item (a) above:

i. Discuss the criteria used to consistently determine when a sensitivity study was used to address the identified source of uncertainty.

ii. Discuss the criteria used to consistently determine when additional RMAs should be implemented because of modeling uncertainty.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 38 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

Response

For those topics identified as potential sources of uncertainty for the TSTF-505 RICT application (i.e., not screened out using the methodology discussed in part (a) of this response), sensitivity analyses were performed and the results of the sensitivity analyses are documented in Section 8.0 of PB-MISC-043 [1].

If the results of the sensitivity analysis demonstrated a significant impact on the TSTF-505 RICT calculations (e.g., >10% change), then the importance measures of the base case and the sensitivity case were reviewed in order to determine if the top risk-significant operator actions change, thus requiring a different set of RMAs for the configuration. If the top risk-significant operator actions remain unchanged, then the proposed RMAs for the configuration would also remain unchanged, thus supporting the conclusion that the source of uncertainty doesnt impact the TSTF-505 RICT calculations.

Only the identified potential sources of uncertainty that are discussed in Section 8.0 of PB-MISC-043 are discussed in Enclosure 9 of the TSTF-505 RICT LAR.

The base FPIE PRA model uncertainty discussions are documented in Appendix A of the Summary Notebook (PB-PRA-013) [2] and the base FPRA model uncertainty discussions are documented in the Uncertainty & Sensitivity Analysis Notebook (PB-PRA-021.62) [3]. These documents are updated after every MOR update. RICT system guidelines which contain RMAs to be used during the application of the RICT program are updated after every MOR update in accordance with Exelon procedures. Exelon procedures describe the process for system guideline creation and specifically mention the consideration of PRA model key assumptions and sources of uncertainty when developing RMAs for the system guidelines.

APLA 10 References

[1] Peach Bottom Atomic Power Station, Assessment of Key Assumptions and Sources of Uncertainty for the Peach Bottom Atomic Power Station PRA, PB-MISC-043, Rev. 1, April 2020.

[2] Peach Bottom Atomic Power Station Probabilistic Risk Assessment, Summary Notebook, PB-PRA-013, Rev. 6, February 2020.

[3] Peach Bottom Atomic Power Station Probabilistic Risk Assessment, Uncertainty and Sensitivity Notebook, PB-PRA-021.62, Rev. 2, February 2020.

Probabilistic Risk Assessment Licensing Branch B (APLB) Audit Questions APLB QUESTION 04 - Systems Not Credited in the Fire PRA As part of its audit (ADAMS Accession No. ML20217L346), the NRC staff reviewed PRA Notebook PB-PRA-021.62, which noted that several systems were identified as not being modeled in the FPRA. The NRC staff notes that some conservative PRA modeling assumptions could have a nonconservative impact on the RICT calculations. If an SSC is part of a system

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 39 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 not credited in the FPRA or is supported by a system that is assumed to always fail, then the risk increases due to taking that SSC out of service are masked. Therefore, provide the following information:

a) Identify the systems or components that are assumed to be always failed in the FPRA or not included in the FPRA (e.g., due to lack of cable tracing or other reasons). Justify that these assumptions have an inconsequential impact on the RICT calculations and no RMAs are required to address these items.

Response

As part of the FPRA, some components were assumed to be failed given the cable data detail required for the FPRA was not included. Components in this category are referred to as Unknown Location (UNL) components because specific cables were not identified for the components. This included cable related to SDC, containment isolation signals, RCIC restart capabilities, and SLC.

A sensitivity analysis was performed to measure the risk associated with the assumption that these components fail in all fire scenarios. The sensitivity removed all UNL components from every fire scenario such that the unknown cables were assumed to never fail. This represents a bounding sensitivity case because the cables would be damaged by some fire scenarios. Table APLB-04-01 below shows the results of this study. Based on the results, the inclusion of the UNL components introduces small change in risk to both Fire CDF and LERF.

Table APLB-04-01 Unknown Cable Locations for FPRA Sensitivity Case Case CDF (/yr) Delta CDF (/yr) LERF (/yr) Delta LERF (/yr)

Base FPRA Model 2.21E-05 -- 2.31E-06 --

UNL Sensitivity 2.05E-05 -1.60E-06 1.85E-06 -4.65E-07 Additionally, specific sensitivity cases for this issue were explored for select LCO conditions. Table APLB-04-02 below shows the results of the selected LCO conditions.

Table APLB-04-02: Unknown Cable Location for FPRA RICT Sensitivity Cases TS/LCO Condition Base Model RICT Sensitivity RICT Estimate (days)

Estimate (days) 3.6.2.3.A - One RHR suppression 30.0 30.0 pool cooling subsystem inoperable 3.3.6.1.A - One or more required PCIS instrument channels 30.0 30.0 inoperable 3.5.1.C - HPCI system inoperable 29.7 30.0 3.1.7.B - One SLC subsystem inoperable for reasons other than 30.0 30.0 Condition A

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 40 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 The results of this study indicate that although there could be a slight shift in the calculated RICT values (i.e., each by a few percent in these cases), the identified compensatory measures would remain the same. That is, the identification of important operator actions, important fire areas, and protected equipment priorities do not change given the results of these sensitivity cases.

However, it is recognized that the SLC out-of-service cases may not represent the true delta risk without modeling that system. For this reason, the SLC cable data was obtained and the FPRA model is currently being modified to remove the SLC system from the list of UNL components. For additional information see the response to APLB Question 05 below.

b) As an alternative to item (a) above, propose a mechanism to ensure that a sensitivity study is performed for the RICT calculations for applicable SSCs that accounts for the impact on the RICT of the 1) conservative FPRA assumption of failed SSCs or 2) SSCs not included in the FPRA model. The proposed mechanism should also ensure that any additional risk from correcting these assumptions is either accounted for in the RICT calculations or is compensated for by applying additional RMAs during the RICT.

Response

Not Applicable - See response to Part a) above.

APLB QUESTION 05 - Implementation Item for Cable Data for Standby Liquid Control LAR Attachment 6 lists the following implementation item that must be completed prior to implementation of the RICT program to satisfy the guidance in NEI 06-09 that the PRA reflect the as-built, as-operated plant and that the PRA technical adequacy is acceptable:

• Exelon will ensure that the updated standby liquid control cable data will be incorporated in the Peach Bottom PRA with sufficient detail to accurately calculate the RICT.

LAR Attachment 6 also states that if implementation of this change constitutes a PRA upgrade as defined in the PRA standard, as endorsed by RG 1.200, then a focused-scope peer review will be performed on this change, and any findings will be resolved and incorporated in the PRA prior to the implementation of the RICT program. However, it is unclear to the NRC staff how the addition of this system model will meet CC-II of the PRA standard, as endorsed by RG 1.200.

In light of this observation, describe how this system will be adequately modeled in the FPRA and in accordance with the PRA standards CC-II.

Response

The SLC system is modeled in the Internal Events PRA in accordance with the PRA standards Capability Category II. However, as some of the cable data required for FPRA modeling associated with the SLC was previously unknown, the PBAPS FPRA did not credit (assumed failed with all fire scenarios) the system.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 41 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Prior to implementation of the RICT program at PBAPS the FPRA will be revised to incorporate the SLC system. Equipment Selection and Cable Selection tasks will be performed consistent with the methodologies used for other systems (documented in PB-PRA-021.52_Rev2 and PB-PRA-021.53_Rev1 respectively) and the PRA standards Capability Category II.

Additionally, fire scenario target sets will be revised such that any cable associated with the SLC system that is located within an already identified scenario target (i.e., cable tray or conduit) will be failed in accordance with target failures associated with the relevant scenarios the target is identified to fail. For targets that become relevant due to the incorporation of the SLC system, they will be reviewed and appropriately included in the target failures for the fire scenarios that could result in the failure of the new SLC system target.

APLB QUESTION 08 - 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 Frequently Asked Question (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).

Response

The treatment of sensitive electronics for the FPRA is consistent with the guidance in FAQ 13-0004. Sensitive electronics mounted inside a control panel are considered qualified up to the heat flux damage threshold for thermoset cables provided that the component is not mounted on the surface of the cabinet and the presence of louvers or other typical ventilation means do not invalidate the guidance provided in the FAQ.

Sensitive electronics not shielded by a robust enclosure consider the lower damage thresholds for sensitive electronics per NUREG/CR-6850.

Electrical equipment was reviewed to determine if sensitive electronics may be subject to the lower damage thresholds per FAQ 13-0004. The identified sensitive electronics did not impact the modeling of the existing fire scenarios postulated in the PRA model. In most cases the exposed sensitive electronics identified were not associated with FPRA credited equipment. In other cases, it was identified that the impact of reducing the damage criteria for the sensitive electronics was already bounded by the existing fire scenarios postulated in the plant location (e.g., full room damage scenarios).

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.

Response

Not applicable - see response to Part a) above.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 42 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 APLB QUESTION 10 - Probabilistic Risk Assessment Model Uncertainty Analysis Results The NRC staff SE to NEI 06-09 specifies that the LAR should identify key assumptions and sources of uncertainty and should assess/disposition each as to its impact on the RMTS application. LAR Enclosure 9, Table E9-3, identifies the key assumptions and sources of uncertainty for the FPRA and provides dispositions for each source of uncertainty for this TSTF-505 application. The NRC staff reviewed the dispositions provided in LAR Table 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.

LAR Enclosure 9, Table E9-3, identifies post-fire HRA as a source of FPRA 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 impact is expected to be small with apparently no sensitivities being performed. To address this source of uncertainty, the LAR states that appropriate RMAs would be required - for example, pre-job briefs. It is unclear to the NRC staff how the RMAs will adequately address the impact on RICT values. Therefore, address the following items:

a) 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.

OR b) Explain what RMAs will be considered to compensate for this uncertainty.

Response

a) The parametric uncertainty evaluation for the FPRA model is documented in Section 3.3 and Appendix A of the Uncertainty & Sensitivity Analysis Notebook. The FPRA parametric uncertainty analysis evaluated the following fire-specific parameters (in addition to the uncertainty parameters from the FPIE analysis):

• Fire Ignition Frequencies - Uses NUREG-2169 uncertainty distributions.

• Non-Suppression Probabilities - Uses NUREG/CR-1278 uncertainty distributions.

• Severity Factors - Uses generic FPIE lognormal uncertainty distributions.

• Spurious Probabilities - Uses NUREG/CR-7150 uncertainty distributions.

• Fire Human Error Probabilities - Uses EPRI [human reliability analysis] HRA Calculator uncertainty distributions and NUREG/CR-2178 uncertainty distributions, for Fire HFE Combinations.

Using the UNCERT software, a Monte Carlo simulation was performed for both Fire CDF and Fire LERF to calculate the mean risk metrics that reflect state-of-knowledge correlation (SOKC) considerations. Table APLB-10-1 below summarizes the results of the parametric uncertainty evaluation performed for the base FPRA model.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 43 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 TABLE APLB-10-1 PARAMETRIC UNCERTAINTY ANALYSIS RESULTS POINT-ESTIMATE UNCERTAINTY CASE DELTA RISK  % DIFFERENCE RISK (/YR) MEAN (/YR)

Unit 2 CDF 2.82E-05 2.95E-05 1.28E-06 5%

Unit 3 CDF 3.97E-05 4.19E-05 2.17E-06 5%

Unit 2 LERF 2.28E-06 3.00E-06 7.15E-07 31%

Unit 3 LERF 2.26E-06 3.17E-06 9.13E-07 40%

Unit 2 LERF 1.75E-06 1.92E-06 1.67E-07 10%

(Excluding 20C032)

Unit 3 LERF 1.50E-06 1.57E-06 6.72E-08 4%

(Excluding 30C032)

The resulting propagated mean values for CDF demonstrate that the point-estimate risk is a good representation of the mean risk. The propagated mean values for LERF are noticeably higher than the point estimate value. Upon further investigation it was determined that this is driven by fire scenarios in two specific relay panels, 20C032 and 30C032. These panels contain cables associated with the safety relief valves (SRVs) and low pressure injection line valves. As a result, the related cutsets involve fire induced hot shorts that cause spurious depressurization (spurious opening of two SRVs) and spurious valve closures that prevent low pressure injection. Rerunning the parametric uncertainty with these fire scenarios excluded yields propagated mean results much closer to the point estimate results.

Based on the results presented in Table APLB-10-1 above, the difference between the point-estimate and parametric mean values for Fire CDF and LERF is small. It is concluded that the point-estimate values are good representations of the mean Fire CDF and Fire LERF values and, therefore, the resulting RICT calculations would be expected to be minimally impacted.

Additionally, Exelon Risk Management procedures contain guidance for development of RMAs in support of the RICT program. Development of RMAs considers those developed for other processes, such as the RMAs developed under the 10 CFR 50.65(a)(4) program and the protected equipment program, as well as RMAs based on the Real-Time Risk tool to identify configuration-specific RMA candidates to manage the risk associated with internal events, internal flooding, and fire events. These actions include, but are not limited to, identification of important Operator Actions, including post-fire actions, for briefings. Based on the results of the uncertainty evaluation shown in Table APLB-10-1 above which includes uncertainty distributions associated with post-fire operator actions and the identification of important operator actions during the RMA development process, it is expected that the potential uncertainty related to post-fire HRA modeling will have a minimal impact on the RICT program.

b) Not applicable - see response to Part a) above.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 44 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Probabilistic Risk Assessment Licensing Branch C (APLC) Audit Questions APLC QUESTION 01 - Impacts from Seismic Hazard Frequencies Section 2.3.1, Item 7 of NEI 06-09, 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 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 SE for NEI 06-09 states that Where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

In Section 3 of Enclosure 4 to the LAR, the licensee stated that the site-specific seismic PRA (SPRA) completed in response to the 10 CFR 50.54(f) request for information associated with the Fukushima Near-Term Task Force (NTTF) activities is not directly used in the RICT program but provides input into the calculation for seismic core damage frequency (SCDF) and seismic large early release frequency (SLERF). The licensee selected the seismic hazard curve that was used in the development of NTTF SPRA model, which is based on the peak ground acceleration (PGA). In the same section of the LAR, the licensee mentioned its seismic hazard and screening report (ADAMS Accession No. ML14090A247), which provided seismic hazard curves at various frequencies at 100 (PGA), 25, 10, 5, 2.5, 1, and 0.5 hertz (Hz). The NRC staff compared the seismic hazard curves between these two documents and found that the PGA hazard curve used in the LAR is different than that in the seismic hazard and screening report.

a) Explain the difference between the two PGA hazard curves cited above and justify the selection of the PGA hazard curve for use in the estimation of the SCDF penalty in the LAR.

Response

Question APLC 1a has two parts: the first is to discuss differences between the two hazard curve studies and the second aspect is to justify the selection of the hazard curve used in the PBAPS LAR seismic penalty calculation. Each of these two aspects of the question is discussed below in order.

The two PBAPS Probabilistic Seismic Hazard Analyses (PSHAs) in discussion are: 1) the PSHA used in the PBAPS 2014 Seismic Hazard Screening Report (SHSR) [ref.

APLC1-1] and 2) the PSHA used in the PBAPS 2018 NTTF 2.1 SPRA submittal [ref.

APLC1-2]. The PSHA performed for the PBAPS 2018 NTTF 2.1 SPRA incorporated refinements and new data compared to the PSHA used in the PBAPS 2014 SHSR. The key differences are summarized below:

• The PBAPS SHSR PSHA used the January 2012 version of the CEUS-SSC seismic sources catalog [ref. APLC1-9]. The PBAPS SPRA PSHA seismic sources analysis used a 2015 analysis [ref. APLC1-10] of an updated version of the CEUS-SSC catalog, as well as a review of seismic data post-2008 (the data gathering end date of the CEUS-SSC study) and investigation of induced seismicity (reservoir-induced or otherwise).

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 45 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

• The PBAPS SHSR used the EPRI 3002000717 (June 2013) ground motion attenuation models [ref. APLC1-11] whereas the PBAPS SPRA PSHA incorporated both the EPRI 3002000717 ground motion attenuation models [ref.

APLC1-11] and the updates to EPRI 3002000717 in August 2013 [ref. APLC1-12].

• The PBAPS SHSR used 30 randomization/profiles for the soil hazard whereas the SPRA PSHA used 60.

• The PBAPS SHSR produced a ground motion response spectra whereas the PBAPS SPRA PSHA in addition produced Foundation Input Response Spectra (FIRS) to be used as input into structural response models:

- FIRS1 - Reference Rock Hazard.

- FIRS2 - Soil Column Outcrop Response@ EL 105 ft with 20 ft of Compacted Backfill Above.

- FIRS3 - Surface Response @ EL 136 ft with Moderately Weathered Rock over Hard Bedrock.

- FIRS4 - Surface Response @ EL 117 ft with 40 ft of Compacted Backfill Below.

- GMRS profile defined by EPRI and calculated in the PBAPS SHSR corresponds to FIRS3 in the PBAPS SPRA PSHA.

The FIRS1 PGA hazard curve is used to define the seismic initiating event frequencies in the SPRA accident sequence quantifications.

• Whereas the PBAPS SHSR focused on the horizontal ground motion, the PBAPS SPRA PSHA developed both horizontal and vertical ground motion response spectra.

The PGA seismic hazard curve from the PBAPS 2018 SPRA was used in the seismic penalty calculation included in the PBAPS LAR because it is the latest and most refined PSHA existing at the time (as well as at the time of this question response).

b) Justify that the consideration of seismic hazard curves at frequencies other than the PGA does not significantly change the SCDF penalty proposed in the LAR.

Response

The PGA seismic hazard curve was used in the LAR seismic penalty calculations for two basic reasons: 1) past industry seismic penalty calculations have used the PGA hazard curve; and 2) the majority of SPRAs (past and present, including the PBAPS 2018 NTTF 2.1 Seismic SPRA) use PGA. Further discussion is provided below as well as sensitivity cases using the various Hz hazard curves.

The PGA hazard curve is most often used in industry and regulatory programs. For example, seismic CDF estimation using PGA is used in extended power uprate LARs (refer to Attachment 3 of Matrix 13 of NRC EPU RS-001 [ref. APLC1-3]). In the case of

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 46 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 RICT assessments, the PGA seismic hazard has been customarily used in TSTF-505 LAR seismic penalty calculations (including the Exelon TSTF-505 LARs).

PGA has been used as the ground motion metric in most industry SPRAs performed to date. Although it has been asserted by some seismic fragility experts [ref. APLC1-5] that 5-10 Hz may be a better predictor of seismic-induced damage at a nuclear power plant than is the PGA ground motion metric, the EPRI SPRA guidelines [ref. APLC1-6] and ASME/ANS PRA Standard [ref. APLC1-7] appropriately allow use of either PGA or other Hz to characterize the seismic hazard input to the SPRA models. Whichever ground motion parameter is used, the hazard curve and the component fragilities need to be in the same units (e.g., both based on PGA, or both based on 10 Hz, etc.) to result in a coherently calculated risk result.

In addition, the seismic penalty calculation of the PBAPS LAR has been re-performed in support of this question response using the different available hazard curves from the PBAPS NTTF 2.1 SPRA PSHA [ref. APLC1-8] and adjusting the plant-level fragility value (median, Am, and keeping Bc=04 for each case) per the shape of the Ground Motion Response Spectra (GMRS) used in Reference APLC1-8. The SCDF from each case is summarized below:

PGA 25 Hz 10 Hz 5 Hz 2.5 Hz 1 Hz 0.5 Hz Plant-Level Fragility 0.51 1.02 0.80 0.48 0.25 0.10 0.05 (Am):

Convolved SCDF: 2.3E-5 2.4E-5 2.3E-5 2.4E-5 2.3E-5 2.1E-5 2.9E-5 Note: These sensitivity calculations include dividing the hazard curve into many more intervals than the eight intervals used in the LAR calculation: as such, the PGA based SCDF shown in this sensitivity differs slightly (~4% difference) from the value documented in the LAR. This point is noted here as detailed information but has no impact on the conclusion of this sensitivity nor the reasonableness of the SCDF value documented in the LAR.

As can be seen, from 1 Hz to PGA (100 Hz) the resulting convolved SCDF is very similar (differing by less than 10%). This is a small difference in such a calculation and considering the uncertainties in seismic hazard and response. The 0.5 Hz convolved SCDF is 25% higher than the PGA-based SCDF used in the PBAPS LAR but the 0.5 Hz range is a very low Hz frequency (applicable to such plant effects as sloshing of water pools) and not the Hz range that would be used in an SPRA that selects to use a non-PGA hazard curve. As such, the seismic penalty SCDF (2.2E-5/yr) documented in Enclosure 4 of the LAR is reasonable and will be used in RICT calculations.

Refer to response to question APLC 03 below discussing the conservatism in the Exelon approach of using the entire seismic penalty SCDF and SLERF in RICT calculations.

APLC 01

References:

1. Exelon, "PBAPS Seismic Hazard and GMRS submittal RS-14-071, Exelon Generation Company, LLC, Seismic Hazard and Screening Report (Central and Eastern United States (CEUS) Sites), Response to NRC Request for Information Pursuant to 10 CFR

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 47 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 50.54(f) Regarding Recommendation 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," dated March 31, 2014 (ADAMS Accession No. ML14090A247).

2. Exelon, "Peach Bottom Atomic Power Station (PBAPS) Units 2 and 3 Seismic Probabilistic Risk Assessment in Response to 50.54(f) Letter with Regard to NTTF 2.1 Seismic," Rev. 0, dated August 28, 2018 (ADAMS Accession No. ML18240A065).

3. NRC, "Review Standard for Extended Power Uprates," RS-001, Rev. 0, December 2003 (ADAMS Accession No. ML033640024).

4. EPRI 1025287, Seismic Evaluation Guidance: Screening, Prioritization and Implementation Details (SPID) for the Resolution of Fukushima Near-Term Task Force Recommendation 2.1: Seismic. Electric Power Research Institute, Palo Alto, CA:

February 2013.

5. Kennedy, R.P., Overview of Methods for Seismic PRA and Margin Analysis Including Recent Innovations, Proceedings of the OECD-NEA Workshop on Seismic Risk, Tokyo, Japan, August 1999. Available from OECD Nuclear Energy Agency, LaSeine St.-

Germain, 12 Boulevard des Iles, F-92130 Issy-les-Moulineaus, France.

6. Electric Power Research Institute (EPRI), Seismic PRA Implementation Guide, Report 3002000709, December 2013.

7. American Society of Mechanical Engineers (ASME), "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, Addenda to ASME/ANS RA-S-2008," ASME/ANS RA-Sb-2013, September 30, 2013.

8. Fugro Consultants, Inc., Probabilistic Seismic Hazards Analysis for Peach Bottom Atomic Power Station PSHA Results Report, Fugro Project Report No. 150001-PR-01, Rev. 3, December 29, 2017.

9. Electric Power Research Institute (EPRI), Central and Eastern United States Seismic Source Characterization for Nuclear Facilities, Report 1021097, January 2012.

10. Electric Power Research Institute (EPRI), Central and Eastern United States Seismic Source Characterization for Nuclear Facilities: Maximum Magnitude Distribution Evaluation, Report 3002005684, June 2015.

11. Electric Power Research Institute (EPRI), EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, Report 3002000717, June 2013.

12. Electric Power Research Institute (EPRI), Errata Sheet for EPRI (2004, 2006) Ground-Motion Model (GMM) Review Project, Report 3002000717, August 2013.

APLC QUESTION 02 - Representativeness of Discretization of Seismic Hazard Curve The licensee provided the PGA seismic hazard curve data from 0.005 gram (g) to 7.5 g in Table E4-1 of Enclosure 4 to the LAR. The seismic hazard interval frequencies are represented

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 48 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 by discretizing the hazard curve into eight bins as shown in Table E4-2 of Enclosure 4 to the LAR. The representative PGA for the last bin is selected to be 0.99 g for representing the entire hazard from 0.9 g to 7.5 g. This approach results in a mean fragility probability of 0.95 instead of 1.0 as shown in Table E4-3 of Enclosure 4 to the LAR. As explained in Enclosure 4 to the LAR, this change has a minor impact on the estimated SCDF value. However, the NRC staff notes that sensitivity analysis 1d in the licensees SPRA report (ADAMS Accession No. ML18240A065) shows a 17 percent increase in SLERF due to refinement in the discretization of the last bin. This is likely to increase the seismic conditional large early release probability (SCLERP) estimate, and therefore, the SLERF penalty estimate. The LAR does not discuss the impact of the refinement of the discretization for the last bin on the estimated SLERF penalty.

Justify that the selected representative PGA of 0.99 g for the last bin is reasonable and conservative for the estimated SLERF penalty or provide an updated SLERF penalty.

Response

The discretized representation of the PBAPS seismic hazard curve in the LAR seismic penalty calculation could have used a higher selected representative magnitude than 0.99 g PGA for the last seismic hazard interval but the decision at the time of the calculation was to maintain consistency with the hazard intervals used in the SPRA. Sensitivity Case 1d documented in the PBAPS SPRA report (ADAMS Accession No. ML18240A065) replaced the base case final hazard interval (>0.9 g PGA) with six hazard intervals with the final interval as >2.3 g PGA. For Sensitivity Case 1d, the last seismic hazard interval was modeled with a representative magnitude of 2.53 g PGA (i.e., 1.1x the starting g value of the final interval) and resulted in a seismic conditional large early release probability (SCLERP) of 1.0 for the final interval. The results of Sensitivity Case 1d showed that the seismic large early release frequency (SLERF) increased by 17% from the base SPRA SLERF value of 4.1E-6/yr to 4.8E-6/yr.

The SCLERP value of 0.2 provided in Enclosure 4 to the PBAPS LAR is calculated from the base SLERF divided by the base SCDF (i.e., 4.1E-6/yr / 2.1E-5/yr = ~0.2). If the SCLERP were calculated using the SLERF from Sensitivity Case 1d, the SCLERP would increase to 0.23 (i.e.,

4.8E-6/yr / 2.1E-5/yr = ~0.23). The SPRA based seismic core damage frequency (SCDF) of 2.1E-5/yr is negligibly affected by Sensitivity Case 1d because the conditional core damage probability (CCDP) is already close to 1.0 in the final hazard interval in the base SPRA model.

Given the potential conservatisms identified in Section 5.3.2 of the SPRA submittal report (ADAMS Accession No. ML18240A065) for the PBAPS base SPRA model (e.g., assumptions for treatment of seismic HRA, seismic correlation of fragilities), use of the SCLERP of 0.2 calculated from the base SPRA results is reasonable for the application of the SLERF penalty in the RICT calculations. As such, the seismic penalty SLERF (4.4E-6/yr) documented in of the LAR is reasonable and will be used in RICT calculations.

Refer to the response to question APLC 03 below discussing the conservatism in the Exelon approach of using the entire seismic penalty SCDF and SLERF in RICT calculations.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 49 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 APLC QUESTION 03 - Seismic Core Damage Frequency and Large Early Release Frequency Penalty Estimate Section 2.3.1, Item 7 of NEI 06-09, 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 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 SE for NEI 06-09 states that Where PRA models are not available, conservative or bounding analyses may be performed to quantify the risk impact and support the calculation of the RICT.

The seismic penalty approach is used to quantify the risk impact and to support the RICT evaluation. The staff notes that there is a site-specific seismic PRA that could be used for this analysis. Section 3 of Enclosure 4 to the LAR states that the site-specific SPRA was not directly used in the RICT program but provided input into the calculation for SCDF and SLERF. The licensee compared the estimated SCDF penalty for the proposed RICT calculations against the point-estimate SCDF from the site-specific SPRA. In addition, the licensee used the SLERF to SCDF ratio from the site-specific SPRA to determine the SLERF penalty for use in the proposed RICT calculations.

The comparison of the estimated SCDF and SLERF penalties against the corresponding point-estimate mean values from the site-specific SPRA does not provide justification that the SCDF and SLERF penalty estimates are conservative, as stated in the NEI 06-09 guidance.

There is no upper bound on the change-in-risk calculation, and the change in risk can exceed the base SCDF and SLERF. However, it appears to the NRC staff that the SPRA could provide the means to justify that the proposed SCDF and SLERF penalty estimates are conservative, and therefore, consistent with the staffs SE for NEI 06-09.

Justify that the SCDF and SLERF penalty estimates are conservative based on the results and insights from change-in-risk calculations for the proposed RICTs using the recent site-specific SPRA.

Response

Throughout NEI 06-09 Rev 0-A and the NRC SE for that document, reference is made to either a "bounding" or "conservative" analysis, or sometimes to a "reasonable bounding analysis," as being acceptable to account for risk for external hazards when a PRA model is not available.

The estimation of seismic risk results for the PBAPS RICT program are more accurately characterized as a "conservative" analysis that uses the following:

• An estimated SCDF,

• An estimated average SCLERP to determine a SLERF, and

• A conservative implementation of the SCDF and SLERF used in RICT assessments.

A truly "bounding" analysis would assume characteristics such as SCDF equal to the seismic hazard frequency of the safe shutdown earthquake (SSE) and an estimated averaged SCLERP of 1.0, both of which are neither reasonable nor useful estimates. The PBAPS RICT evaluation estimates a nominal SCDF, estimates an averaged SCLERP based on the PBAPS SPRA model, and then employs these estimates in a conservative manner in the RICT process by

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 50 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 applying the total SCDF and total SLERF as delta SCDF and delta SLERF in each RICT calculation.

As identified by the NRC in the introductory text above, it can be postulated that there is no upper bound on the change-in-risk calculation, and the change in risk can exceed the base SCDF and SLERF. However, consistent with recent industry SPRA results, the PBAPS SPRA results are dominated by the risk contribution from seismic fragility failure modes and seismic operator actions, which are not impacted by changes due to the RICT program. Non-seismic failure modes (e.g., pump fails to run or start, valve fails to change position) are non-significant contributors to the PBAPS SPRA quantitative results (Fussell-Vesely risk importance values

<5E-3; the majority are E-4 and much lower, as well as most Risk Achievement Worth (RAW) importance values are ~1.0) such that changes due to the RICT program are very unlikely to result in an increase in SCDF or SLERF that is equal to or greater than the base estimated SCDF and SLERF used for the PBAPS RICT seismic penalty calculations. This is typical of SPRAs (i.e., non-seismic failures are non-significant contributors to calculated seismic risk).

Therefore, applying the total SCDF and total SLERF as delta SCDF and delta SLERF in each RICT calculation is judged to invoke sufficient conservativism regarding the incorporation of seismic risk insights into the RICT calculations.

In response to this question and for added illustration, the PBAPS SPRA was quantified for selected key SSCs and the delta risk results were compared to the LAR seismic penalty results.

This information is summarized below in Table APLC3-1. As can be seen from the table, use of the PBAPS SPRA would produce seismic risk contributions to RICT calculations that are very low (e.g., one to two orders of magnitude lower) in comparison to the LAR approach of using seismic penalty SCDF (i.e., 2.2E-5/yr) and SLERF (i.e., 4.4E-6/yr) annual values as the delta seismic risk contributions in all RICT calculations.

Table APLC3-1 Examples of PB SPRA-Based Delta Seismic Risk Input to RICT vs Seismic Penalty Ratio Ratio Compared Compared to LAR to LAR Delta SCDF Delta SLERF Component(1) SCDF(2) Penalty SLERF(2) Penalty U2 RCIC Turbine/Pump 5.5E-07 0.02 2.8E-07 0.06 U2 HPCI Turbine/Pump 4.4E-07 0.02 2.9E-07 0.07 EDG E3 2.2E-07 0.01 6.8E-08 0.02 125VDC Battery 2BD01 3.3E-06 0.15 2.2E-07 0.05 125VDC Battery 3CD01 2.5E-06 0.12 2.2E-07 0.05 4kV AC Bus 20A16 6.8E-07 0.03 7.2E-09 0.002 125VDC Bus 3CD17 1.0E-06 0.05 1.3E-07 0.03 Notes to Table APLC3-1:

(1) Components selected as examples to illustrate the conservative effect of using the seismic penalty approach compared to use of the SPRA directly.

(2) Delta SCDF and Delta SLERF calculated using PBAPS NTTF 2.1 SPRA and setting the component in each case to Failed (i.e., logical TRUE) for the SPRA quantification.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 51 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Technical Specifications Branch (STSB) Audit Questions STSB QUESTION 01 - Technical Specification 3.5.1.E, One ADS [Automatic Depressurization System] Valve Inoperable LAR Enclosure 1, Table E1-1 lists in the column of TS 3.5.1.E a condition with one ADS valve inoperable. The corresponding column of the SSCs Covered by TS LCO Condition indicates that ADS (five safety relief valves) are required to be operable, and the column of Design Success Criteria indicates that five ADS valves are available.

Clarify for TS 3.5.1.E condition with one of five required ADS valves inoperable, that the Design Success Criteria need 3 or 4 available ADS valves. Discuss the Analyses of Record (AOR) that demonstrated adequacy of 3 or 4 ADS valves for reactor pressure vessel rapid depressurization to mitigate the loss-of-coolant accident consequences and reference the NRC documents approving the AOR of the concern or address the acceptability of the AOR if it was not previously approved by the NRC.

Response

Reference STSB-01-1, the Loss of Coolant Accident (LOCA) report, is the current LOCA analysis of record (AOR) as documented in the current PBAPS Core Operating Limits Report (COLR), via Reference 2 of the COLR, for both Units 2 and 3 (References STSB-01-2 and STSB-01-3), and per the most recent PBAPS 10 CFR 50.46 report submitted to the NRC (Reference STSB-01-4). The limiting small break scenario is documented in the Reference STSB-01-1 LOCA AOR. A more detailed discussion on the limiting scenario definition can also be found in Reference 15 of the LOCA AOR (Reference STSB-01-5).

Both the Reference 1 LOCA AOR and Reference 5 LOCA report document that the most limiting small break scenario is the small recirculation-line-break with battery failure, which takes out, in addition to a number of low pressure emergency core cooling systems, the HPCI system.

Per these reports, this scenario is the most limiting break location, break size, and single failure combination for PBAPS, and bounds all other credible single failure conditions for small recirculation line breaks. Therefore, a small break scenario with a single failure that takes one ADS valve out of service is bounded by the most limiting small break scenario described in the LOCA AOR. Therefore, TS 3.5.1.E (one ADS valve inoperable) is supported by the existing LOCA analysis.

For a postulated small break scenario with a single failure of one ADS valve (four out of five valves available), the HPCI system would be available. The HPCI system is able to deliver inventory at high pressure without the need for depressurizing the vessel, and hence, would significantly delay fuel uncovery (if not completely mitigate inventory loss through the break) compared to the limiting scenario, which in turn would help reduce the heat up rate (if any), and results in very limited cladding temperature increase (if any).

The condition that assumes unavailability of both the HPCI system and one ADS valve (multiple coincidental failures) is a separate condition that is addressed in TS 3.5.1.i, which requires immediate entry to TS 3.0.3.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 52 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 STSB 01 References

1. LOCA AOR: Project Task Report, Exelon Generation Company LLC, Peach Bottom Atomic Power Station, Units 2 & 3 MELLLA+ Task T0407: ECCS-LOCA Performance, 0000-0162-2354-R0, Revision 0, (PLM OOON0296 Revision 0), December 2013.

2. COLR Unit 2: COLR PEACH BOTTOM 2 Rev. 16.

3. COLR Unit 3: COLR PEACH BOTTOM 3 Rev. 16.

4. Letter from David P. Helker (Exelon Generation Company, LLC) to U.S. Nuclear Regulatory Commission, "Emergency Core Cooling System (ECCS) Loss of Coolant Accident (LOCA) 10 CFR 50.46 30-Day Report," dated November 5, 2020.

5. GNF2 LOCA Report: Peach Bottom Atomic Power Station Units 2 & 3, GNF2 ECCS-LOCA Evaluation, 0000-0100-8531-R1, March 2011.

STSB QUESTION 02 - Technical Specification 3.5.1.F, One Automatic Depressurization System valve inoperable and One Low Pressure Emergency Core Cooling System Subsystem Inoperable LAR Enclosure 1, Table E1-1 lists in the column of TS 3.5.1.F a condition with one ADS valve inoperable and one low pressure Emergency Core Cooling System (ECCS) injection/spray subsystem inoperable. Clarify the same for 3.5.1.F. The corresponding column of the SSCs Covered by TS LCO Condition states, See LCO Condition 3.5.1.A and 3.5.1.E, which indicates that ADS (five safety relief valves) are required to be operable, and the column of Design Success Criteria indicates that five ADS valves are available.

Clarify for TS 3.5.1.F Condition with one of 5 required ADS valves inoperable, that the DSC need three or four available ADS valves. Discuss the AOR that demonstrated adequacy of three or four ADS valves for reactor pressure vessel rapid depressurization to mitigate the loss-of-coolant accident consequences and reference the NRC documents approving the AOR of the concern, or address the acceptability of the AOR if it was not previously approved by the NRC.

Response

For TS 3.5.1.F, for the one ADS valve inoperable aspect of this Condition, the response is the same as the response for TS 3.5.1.E. When in this Condition, the TS Bases for 3.5.1.F states:

"If any one low pressure ECCS injection/spray subsystem is inoperable in addition to one inoperable ADS valve, adequate core cooling is ensured by the OPERABILITY of HPCI and the remaining low pressure ECCS injection/spray subsystem."

Electrical Engineering Branch (EEEB) Audit Questions EEEB QUESTION 01 - Technical Specification 3.8.1.D, Two or More Offsite Alternating Current Power Circuits Inoperable

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 53 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Peach Bottoms DSC is derived from the current licensing basis of the plant, as documented in the Updated Safety Analysis Report, and should include a minimum set of required equipment that has the capacity and capability to safely shut down the reactor in case of an accident and maintain it in a safe condition. In Table E1-1 of Enclosure 1 of the LAR, the DSC for TS LCO 3.8.1.D (two or more offsite AC power circuits inoperable) is one of two offsite AC power sources. The NRC staff notes that if both offsite circuits are inoperable, one offsite AC power source as listed in the DSC is not available to provide the necessary power to safely shut down the reactor and maintain it in safe condition. Therefore, it is not clear how one offsite circuit can be the DSC for TS 3.8.1.D during the RICT program entry.

Explain this apparent discrepancy in Table E1-1 of Enclosure 1 of the LAR. Additionally, describe any effect the discrepancy may have on the PRA success criteria for TS LCO 3.8.1.D.

Response

The Design Success Criteria provided in LAR Table E1-1 for Technical Specification 3.8.1.D, Two or More Offsite Alternating Current Power Circuits Inoperable, would apply under two discrete scenarios: 1) a loss of Offsite Circuits due to onsite equipment conditions, and 2) a loss of Offsite Circuits due to a loss of offsite power supplies resulting from a Loss of GRID.

Under scenario 1: Loss of Qualified Offsite Circuits, the offsite power supplies remain available; however, the onsite portion of the offsite circuit is degraded.

For example:

With an initial condition of the 2 Emergency Auxiliary (EA) transformer out of service for planned maintenance, all eight 4 kV buses would be powered from 3EA transformer via the second preferred offsite source, 343 Startup (SU) transformer (220-34 Line). Assume 343 SU transformer Load Tap Changer malfunctions and is declared TS inoperable resulting in both offsite power circuits being declared inoperable. The robust design of the PBAPS offsite power supply configuration would allow the Design Success criteria to be met by transferring the source of power to 3EA transformer from 343 SU transformer to 3SU transformer, maintaining 3EA powered from the Grid. In this scenario RICT would apply.

Under scenario 2, Loss of Grid, would result in a complete loss of the 500 kV Grid and a complete loss of the 230 kV Grid. In this scenario both PBAPS units would proceed to a normal shutdown using Onsite AC Sources (i.e., EDGs). RICT would not apply as the units would no longer be in Mode 1 or 2.

Background:

PBAPS UFSAR Section 8.3.2 provides a description of the off-site power supplies:

"Startup auxiliary power is provided from any of the three offsite sources:

1. The tap on the 230 kV Nottingham-Cooper line feeds the 230/13 kV regulating transformer (startup and emergency auxiliary transformer no. 2) at the station.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 54 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278

2. At the North Substation, thirteen kilovolts (13 kV) from the tertiary winding on the 500/230 kV auto-transformer feeds the 13/13 kV regulating transformer (startup and emergency auxiliary regulating transformer no. 3) which connects to the 13 kV switchgear at the station.

3. At the North Substation, 13 kV can be supplied from the 230/13 kV regulating transformer (startup transformer no. 343) which is supplied by the 230 kV Peach Bottom-Newlinville line and connects to the 13 kV switchgear.

The PBAPS design includes two "qualified offsite circuits consisting of all breakers, transformers, switches, interrupting devices, cabling, and controls required to transmit power from the offsite transmission network to the onsite Class 1E emergency bus or buses."

The TS 3.8.1 Bases describe the alternate power source as follows: "the alternate offsite source is the auto-transformer (500/230 kV) at North Substation which feeds a 230/13.8 kV regulating transformer (startup and emergency auxiliary transformer no. 3), the 3SU regulating transformer switchgear, and the 2SUA switchgear." The Bases further states: "The alternate source can only be used to meet the requirements of one offsite circuit."

The power sources for the station auxiliary power system are sufficient in number and independent electrically and physically such that no single event would be likely to cause a simultaneous outage of all sources.

The source for startup and emergency auxiliary regulating transformer no. 3 provides an added degree of reliability. It is sourced from the tertiary winding of a 500 kV/ 230 kV/ 13 kV transformer (no. 1 Transformer in the PB North 500 kV Switchyard). This configuration allows no. 3 transformer to be powered from a combination of the 500 kV transmission system and the 230 kV transmission system. In the event the 230 kV transmission system is lost, no. 3 transformer can remain powered from the 500 kV transmission system.

A bounding example of scenario 2 above would be a Loss of Offsite Power followed by a LOCA as discussed in TS Bases 3.8.1:

"With two or more of the offsite circuits inoperable, sufficient onsite AC sources are available to maintain the unit in a safe shutdown condition in the event of a DBA or transient. In fact, a simultaneous loss of offsite AC sources, a LOCA, and a worst-case single failure were postulated as a part of the design basis in the safety analysis. Thus, the 24-hour Completion Time provides a period of time to effect restoration of all but one of the offsite circuits commensurate with the importance of maintaining an AC electrical power system capable of meeting its design criteria."

"This level of degradation means that the offsite electrical power system may not have the capability to effect a safe shutdown and to mitigate the effects of an accident; however, the onsite AC sources have not been degraded. This level of degradation generally corresponds to a total loss of the immediately accessible offsite power sources."

EEEB QUESTION 05 - RMA Examples As part of its evaluation, the NRC staff reviews the proposed RMA examples for reasonable assurance that the RMAs are considered to monitor and control risk and to ensure adequate defense in depth. Enclosure 12 of the LAR describes the RMAs examples for TS 3.8.1.A,

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 55 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 TS 3.8.1.B, TS 3.8.1.D, and TS 3.8.4.A. However, the LAR does not include the RMA examples for TS 3.8.7 conditions related to the power distribution system. Provide the RMA example(s) for TS 3.8.7.

Response

TS 3.8.7 addresses conditions related to the AC and DC electrical distribution systems.

Candidate RMAs for an AC distribution system would include:

• Protect remaining AC power distribution subsystems, per OP-AA-108-117.

• Prohibit any elective maintenance on ALL safety related AC distribution subsystems.

• Take the required actions per procedure for loss of AC distribution subsystem.

• Pre-stage material for work activity and ensure parts availability for any contingent work.

• Minimize activities on equipment powered by remaining AC vital buses.

• Prohibit any elective maintenance on ALL vital AC and DC distribution subsystems.

• Prohibit trip sensitive activities and activities that could result in a plant transient.

• Take any Maintenance Rule a4 fire risk RMAs.

• Brief shift operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of AC power.

Candidate RMAs for a DC power distribution subsystem would include:

• Protect remaining DC power distribution subsystems, per OP-AA-108-117.

• Prohibit any elective maintenance on ALL DC distribution subsystems.

• Take the required actions per procedures for loss of 125/250VDC safety related bus.

• Prohibit trip sensitive activities and activities that could result in a plant transient.

• Take any Maintenance Rule a4 fire risk RMAs.

• Pre-stage materials for work activity.

• Brief shift operations crew concerning the unit activities, including any compensatory measures established, and review of the appropriate emergency operating procedures for a Loss of DC power and station blackout.

• Brief shift operations crew concerning the impact the DC division has on the potential response to plant events such as reduced control systems.

NOTE: The first two CTs are for opposite unit buses, the last two are for this unit. PBAPS has common EDGs and unitized 4KV buses and unitized DC buses.

Instrumentation & Controls Branch B (EICB) Audit Questions EICB QUESTION 01 - Instrumentation & Controls Redundancy and Diversity RG 1.174, Revision 3, states the licensee should assess whether the proposed licensing basis change meets the defense-in-depth principle by not overrelying on programmatic activities as compensatory measures associated with the change in the licensing basis. RG 1.174 further elaborates that human actions (e.g., manual system actuation) are considered as one type of compensatory measure.

Supplement to License Amendment Request Attachment Adopt Risk Informed Completion Times TSTF-505 Page 56 of 56 Response to NRC Audit Questions Docket Nos. 50-277 and 50-278 Therefore, in LAR Attachment 5, if the diverse means identified are the manual actuations, demonstrate by one example that these manual actuations identified as the diverse means are modeled in the plant PRA defined in plant operation procedures to which operators are trained, and confirm the completion times associated with these actions are evaluated as adequate.

Response

TS 3.3.5.2 for a failure of an automatic shutdown of the RCIC system with high reactor water level was examined. This would require manual shutdown of the RCIC system. The procedure for the Operator Response Time Program at PBAPS provides the list of time critical / time sensitive operator actions, including an action for controlling HPCI/RCIC injection to prevent high level trip. The time required for performance is 6.1 minutes. The expected performance time is 4.28 minutes. Another example would be Loss of Offsite Power requiring RCIC start.

Per the response time procedure, RCIC is required to be started within 15.7 minutes. The expected performance time is 1 minute, 45 seconds. These actions are periodically validated through simulator performance and governed by the operator response time procedure.

Operations is trained regularly on this procedure.

Based on question clarification during the audit, the following additional information is provided.

The LAR Attachment 5 referenced in the above question (EICB 01), was provided to indicate the design of the PBAPS logic systems and how it addresses the listed transients/

accidents. For the column "Diverse Instrumentation," the logic design was tabulated to provide all the means by which the logic addresses the column "Transient/Accident" as listed in the Updated Final Safety Analysis Report (UFSAR). The list presented in Attachment 5 does not represent any change in the way defense in depth is achieved; it only reflects those functions which can affect the associated logic (e.g., automatic or manual relays within the reactor protection system (RPS)). This Attachment 5 table represents the current licensing basis and no proposed changes to system redundancy, independence, and diversity are proposed in this LAR. Additionally, since the proposed change does not alter the logic in the system in , there is no increase in programmatic activities and no change to safety margins and defenses incorporated into the current plant design and operation.

While the manual actions listed in the Attachment 5 are provided by design, they are not the primary means of mitigating the accident/transient listed. Each instrumentation logic is provided with redundant channels (e.g., RPS A1/A2 and B1/B2) which are credited in the licensing basis for successfully performing the function. If a sufficient number of the redundant channels fail, such that a loss of function occurs, then a RICT cannot be applied as described in the LAR and the TS markups. Since a RICT would not be calculated under the loss of automatic function circumstance, the PRA modeling of the manual actions listed in Attachment 5, either directly or through the use of surrogates, has no impact.