RS-24-019, Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Completion Times TSTF

Response to Request for Additional Information Related to License Amendment Request to Revise Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Completion Times TSTF

ML24079A122
Person / Time
Site:	Quad Cities
Issue date:	03/19/2024
From:	Humphrey M Constellation Energy Generation
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
RS-24-019
Download: ML24079A122 (1)

Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office 10 CFR 50.90 RS-24-019 March 19, 2024 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

Response to Request for Additional Information Related 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 P.R. Simpson (Constellation Energy Generation LLC) to U.S.

NRC,

Subject:

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 June 8, 2023 (ADAMS Access No. ML23159A249)

2. Email from R. Kuntz (U.S. NRC) to R. Steinman (Constellation Energy Generation, LLC),

Subject:

Request for Additional Information RE:

TSTF-505 and 10 CFR 50.69 license amendments, dated March 6, 2024 (ADAMS Accession No. ML24066A153)

3. Letter from A. Russell (U.S. NRC) to Technical Specifications Task Force,

Subject:

Final Model Safety Evaluation of Technical Specifications Task Force Traveler TSTF-591, "Revise Risk-Informed Completion Time (RICT)

Program" (EPID L-2022-PMP-0003), dated December 18, 2023 (ADAMS Accession Nos. ML23325A213 and ML23325A214)

In the Reference 1 letter, Constellation Energy Generation, LLC (CEG) requested an amendment to Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS) Units 1 and 2. The proposed amendment would modify the Technical Specifications (TS) requirements to permit the use of Risk Informed Completion Times (RICTs) in accordance with TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b," (ADAMS Accession No. ML18183A493).

Reference 2 identified additional information needed to support the NRC review of Reference 1.

March 19, 2024 U.S. Nuclear Regulatory Commission Page 2 This letter provides the requested additional information in Attachment 1, except for one item which will be addressed in a separate submittal.

Since the submittal of the original license amendment request in June 2023, a subsequent Technical Task Force Traveler, TSTF-591, "Revised Risk-Informed Completion Time (RICT)

Program," has been reviewed and approved as a change to the Standard Technical Specifications (STS). QCNPS has determined that is it desirable to incorporate the additional changes to the RICT Program described in TSTF-591 as part of their initial implementation of RICT as opposed to subsequently requesting adoption under the Consolidated Line Item Improvement Process (CLIIP). As a result, this letter requests adoption of TSTF-591 in conjunction with the current request to adopt TSTF-505. TSTF-591 revises the TSTF-505 proposed TS Section 5.5 "Risk Informed Completion Time Program," to reference Regulatory Guide (RG) 1.200, Revision 3, instead of Revision 2, and to make other changes to the program description wording. Additionally, TSTF-591 adds a new requirement to TS Section 5.6, "Reporting Requirements," to inform the NRC of future newly developed methods used to calculate a RICT. Additional supporting information related to the proposed adoption of TSTF-591 is provided in Attachment 2. provides replacement mark-ups for select Technical Specification (TS) pages that support the responses provided in Attachment 1. The revised TS pages included in supersede the previously transmitted Reference 3 mark-up for the respective page. Pages not included in Attachment 3 remain as originally submitted. Attachment 4 provides a revised copy of the program implementation items table, which supersedes the originally submitted Table A5-1 in its entirety. Similarly, Attachment 5 provides a complete updated copy of Table E1-1. Attachment 6 provides an excerpt of a change to the wording in for a single item that is being removed from the RICT program. In this case, only the affected rows of the table are provided, all other rows of the originally submitted table are unchanged.

CEG has reviewed the information supporting the finding of no significant hazards consideration, and the environmental consideration that were previously provided to the NRC in Reference 1.

The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.

CEG is notifying the State of Illinois of this supplement to a previous application for a change to the operating license by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b).

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Ms. Rebecca L. Steinman at (779) 231-6162.

March 19, 2024 U.S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19th day of March 2024.

Respectfully, Mark Humphrey Sr. Manager Licensing Constellation Energy Generation, LLC Attachments:

1.

Response to Request for Additional Information

2.

Evaluation of the TSTF-591 Proposed Change

3.

Revised Proposed Technical Specification Changes (Mark-Ups)

4.

Revised Table A5-1 RICT Program Implementation Items

5.

Revised Table E1-1 In-Scope TS/LCO Conditions to Corresponding PRA Functions

6.

Excerpts of Revised TSTF-505/QCNPS Cross-References from Attachment 4 cc:

Regional Administrator - NRC Region III NRC Senior Resident Inspector - QCNPS NRC Project Manager, NRR - QCNPS Illinois Emergency Management Agency - Division of Nuclear Safety Humphrey, Mark D.

Digitally signed by Humphrey, Mark D.

Date: 2024.03.19 10:46:20 -05'00'

ATTACHMENT 1 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Response to Request for Additional Information Related 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" Response to Request for Additional Information

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 1 of 72 Docket Nos. 50-254 and 50-265 REQUEST FOR ADDITIONAL INFORMATION LICENSE AMENDMENT REQUESTS TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2 AND IMPLEMENT 10 CFR 50.69 CONSTELLATION ENERGY GENERATION, LLC QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254 AND 50-265 By letters dated June 8, 2023 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML23159A249 and ML23159A253, respectively), Constellation Energy Generation, LLC (Constellation, the licensee) submitted two license amendment requests (LARs) for Quad Cities Nuclear Power Station (Quad Cities), Units 1 and 2. The proposed amendments would modify Renewed License Nos. DPR-29 and DPR-30, and the Technical Specifications (TSs) to adopt Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, "Provide Risk-Informed Extended Completion Times, RITSTF

[Risk-Informed Technical Specification Task Force] Initiative 4b" (ML18183A493), and to allow for the implementation of the provisions of Title 10 of the Code of Federal Regulations, Part 50 (10 CFR 50), section 50.69, "Risk-informed categorization and treatment of structures, systems, and components [SSCs] for nuclear power reactors."

The NRC staff has determined that additional information is needed to support its review. The following is the NRC staff's draft request for additional information.

APLA RAI-01 Concerning the quality of the PRA model, Nuclear Energy Institute (NEI) 06-09-A, "Risk-Informed Technical Specifications Initiative 4b Risk-Managed Technical Specifications (RMTS) Guidelines," Revision 0-A (ML12286A322), states that Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis" (ML17317A256) and RG 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities" (ML090410014) define the quality of the PRA in terms of its scope, level of detail, and technical adequacy. The quality must be compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change.

Regarding digital instrument and control (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 the lack of industry data for digital I&C components, the difference 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 risk informed completion time (RICT) and 50.69 programs.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 2 of 72 Docket Nos. 50-254 and 50-265 Table E9-1 of Enclosure 9 to the RICT (TSTF-505) LAR and the table in Attachment 6 of the 10 CFR 50.69 LAR identifies the digital feedwater control (DFWC) failure probabilities as a potential key uncertainty and performed a sensitivity of increasing the likelihood that the DFWC would result in overfilling the reactor pressure vessel (RPV). The NRC staff notes that another failure mode consideration is loss of feedwater to the RPV. It is unclear to the NRC staff if the sensitivity study failure mode is bounding for this uncertainty. Therefore, address the following:

a) Provide justification that the DFWC failure mode addressed in the LAR sensitivity is the bounding failure mode of this control system.

b) If another failure mode is determined to be bounding, provide justification (e.g., describe and provide the results of a sensitivity study) that demonstrates the modeling uncertainty associated with crediting digital I&C systems has an inconsequential impact on the RICT calculations and 50.69 categorization.

Clarify whether digital I&C systems, other than DFWC, are credited in the PRA models that will be used in the RICT and 50.69 programs.

c) If other digital I&C systems are credited in the PRA models and will be used in the RICT or 50.69 programs, provide justification (e.g., describe and provide the results of a sensitivity study) that demonstrates the modeling uncertainty associated with crediting digital I&C systems has an inconsequential impact on the RICT calculations and 50.69 categorizations.

Alternatively, for RICT, if a justification is not provided, identify which LCOs are determined to be impacted by digital I&C systems modeling for which risk management actions (RMAs) will be applied during a RICT. Explain and justify the criteria used to determine what level of impact to the RICT calculation requires additional RMAs.

Constellation Response to APLA RAI-01 a) The PRA contains one basic event representing failure of the digital feedwater controller (DFWC), namely, event 1FWCNTRLR----F--, FAILURE OF FRV DUE TO FW CONTROLLER MALFUNCTION.

This event is input to two gates:

i.

FRV-CONTROL, FEEDWATER REG. VALVE CONTROL FAILURE Failure of the DFWC will contribute to the possibility of failure of the FW reg valves. A manual action to place the master controller in manual mode must also fail.

The selection of the DFWC basic event for use in the sensitivity analysis is thus appropriate for considerations of loss of feedwater to the reactor pressure vessel (RPV).

ii.

GMSO-2AI-3, FEEDWATER CONTROL FAILURE TO STOP FEEDWATER Failure of the DFWC basic event will cause a failure to stop feedwater as modeled in this part of the PRA logic.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 3 of 72 Docket Nos. 50-254 and 50-265 Thus, both modes of FW control are addressed by the sensitivity, using a single basic event to impact both modes: fail to continue FW (loss of flow to RPV) and fail to stop FW (RPV overfills).

b) Both failure modes listed in part a) are encompassed by the sensitivity study described in TSTF-505 LAR Enclosure 9.

c) Not applicable. No other digital instrumentation and control (DI&C) systems are credited.

APLA RAI-02 The NRC staff safety evaluation to NEI 06-09, Revision 0, specifies that the LAR should identify key assumptions and sources of uncertainty and to assess/disposition each as to their impact on the RMTS application. LAR Enclosure 9, Table E9-1 identifies the key assumptions and sources of uncertainty for the internal events and fire PRA models and provides dispositions for each source of uncertainty for this TSTF-505 application. NRC staff reviewed these dispositions and is unclear that some uncertainty dispositions fully addressed the potential impact to RICT calculations. Therefore, address the following:

TSTF-505 LAR Enclosure 9, Table E9-1 regarding core cooling success following containment failure, states that the sensitivity study demonstrates that all risk metrics are sensitive to this uncertainty. NRC review of the TSTF-505 LAR, Section 8.1 of the Assessment of Key Assumptions and Sources of Uncertainty Notebook demonstrates impacts on CDF and LERF ranging from thirty-three to eighty-one percent. It appears that this source of uncertainty could plausibly impact this application. However, the TSTSF-505 LAR reasons that the increased factor value is not considered credible. The NRC staff notes that it requires insights from credible sensitivity studies for its review process and determination. It is unclear to the NRC staff that this assumption has no impact on the RICT or 50.69 programs. Therefore, address the following:

a) Clarify what increased factor value should be used for this sensitivity. Include in this discussion justification that the selected increased factor value appropriately bounds the increase in risk associated with this uncertainty.

b) Based on the response to part (a) above, provide justification that the uncertainty associated with core cooling success following containment failure does not significantly impact any RICT calculation or 50.69 categorizations.

c) Alternatively to part (b) above, for RICT, provide how this source of uncertainty, such as additional risk management actions (RMAs), would be addressed in the RICT program.

d) Additionally, the Quad Cities, Units 1 and 2, PRA does not model post-containment failure injection. Modeling of post-containment failure injection may impact the conservatism of the stated probability of 6E-02 for large drywell failure. Explain how post-containment failure injection was specifically analyzed, and confirm whether this failure mode is bounded by the probability of drywell failure.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 4 of 72 Docket Nos. 50-254 and 50-265 Constellation Response to APLA RAI-02 a) An increase in the conditional probability of basic event 1CNPVDWRUPT--R-- (large drywell (DW) containment failure causes loss of injection) by a factor of two (i.e., increase from 6.0E-2 to 1.2E-1) is evaluated to be more appropriate as a realistic upper bound sensitivity study than the factor of 10 increase discussed in LAR Enclosure 9, Table E9-1. The factor of 10 increase for the probability is not considered credible for this sensitivity study because the base probability of 6.0E-2 for large DW containment failure is already conservative.

Background Information The containment failure probabilities used in the QCNPS PRA are modeled for a range of containment pressure and containment temperature scenarios that use information from Appendix C.8 of NUREG-1150 for the documented Mark I BWR plant (i.e., Peach Bottom)

[Reference 1]. The introductory section of NUREG-1150 Appendix C.8 identifies the following other critical factors in the characterization of containment performance, in addition to the likelihood of failure, considered in the study:

x Failure size x

Location of failure In NUREG-1150, the failure sizes were categorized as leakage or rupture. The failure locations were defined based on the type of containment (e.g., drywell head, drywell body, wetwell above the water line, wetwell below the water line for the Peach Bottom Mark I containment).

NUREG-1150 identifies that for low containment temperature conditions (e.g., < 400°F), the range of possible failure pressures for the Peach Bottom primary containment was evaluated to be in the range of 120-174 psig. For the low temperature conditions, the containment failure probabilities were dominated by DW leakage, wetwell (WW) leakage, or WW rupture. The last full paragraph of p. C-88 of NUREG-1150 provides the following results:

Conditional upon containment failure in the lower part of this range

[120-174 psig], 50 percent of probability was associated with leakage at the drywell head while the remaining failure probability was dominated by wetwell leakage above the suppression pool. At the top edge of the failure pressure range [120-174 psig], wetwell rupture above and below the suppression pool were each assessed to account for 25 percent of the total probability, with catastrophic wetwell rupture accounting for a further 10 percent. Leakage in the drywell (principally at the head) accounted for approximately 25 percent of the conditional probability, with wetwell leakage accounting for the remaining 15 percent.

Large DW failure (i.e., rupture) was not identified as a contributor to the containment failure probabilities at low temperature. Therefore, based on the results in NUREG-1150 for Peach

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 5 of 72 Docket Nos. 50-254 and 50-265 Bottom, the probability of a large DW failure at low temperature was estimated to be small (e.g., large DW failure probability <1E-2).

The QCNPS Mark I primary containment is designed with features similar to the Mark I BWR plant documented in NUREG-1150 and is evaluated to result in similar containment failure probabilities for similar severe accident conditions. In addition, Section 1.3.6 "Containment Analysis" of the QCNPS Individual Plant Examination (IPE) [Reference 2] identifies that if the containment fails at relatively high pressures, the likely location will be the wetwell vent line bellows because the eight (8) vent line bellows have the lowest mean failure pressure in the analysis.

Although the NUREG-1150 analysis did not explicitly identify a probability for a large DW failure (i.e., rupture) for low temperature conditions, the QCNPS PRA conservatively estimated a large DW failure probability of 6.0E-2. The large DW failure basic event is conservatively modeled to result in guaranteed failure of RPV injection following containment failure (e.g., for a Loss of DHR scenario). Given that the 6.0E-2 probability assigned for large DW failure is estimated to be conservative based on the information in NUREG-1150, increasing the value by a factor of 2 (i.e., increase from 6.0E-2 to 1.2E-1) is evaluated to be more appropriate as a realistic upper bound sensitivity study than the factor of 10 increase discussed in LAR Enclosure 9, Table E9-1.

b) Results of the sensitivity analysis using a factor of two increase and the impact on selected RICT cases are provided in two tables. Table APLA RAI-02.1 shows the change in CDF and LERF due to increasing the basic event probability by a factor of two in the FPIE and fire PRAs, based on a zero maintenance model. Table ALPA RAI-02.2 shows the change in RICT for a selected group of Technical Specifications (TS); these TS were selected because the related systems are associated with post-containment challenge injection scenarios.

Both tables demonstrate the relative insensitivity of the results to uncertainty in the parameter value, even given a 100% increase in the parameter value (compared to the nominal value, which is already conservative).

c) No additional risk management actions (RMAs) are considered at this time.

Table APLA RAI-02.1 - CDF and LERF Sensitivity Results -

Factor of Two Increase for 1CNPVDWRUPT--R--

Model / Case CDF (/yr)

%Delta CDF(3)

LERF (/yr)

%Delta LERF(3)

FPIE(1)

Base 3.45E-06 2.22E-07 Sensitivity 3.57E-06 3.5%

2.280E-07 2.7%

Fire(2)

Base 3.43E-05 2.99E-06 Sensitivity 3.74E-05 9.0%

3.16E-06 5.7%

Total Base 3.78E-05 3.21E-06 Sensitivity 4.09E-05 8.5%

3.38E-06 5.5%

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 6 of 72 Docket Nos. 50-254 and 50-265 Notes to Table APLA RAI-02.1:

(1) FPIE truncation limit is 1.00E-11/yr for CDF and 1.00E-12/yr for LERF.

(2) Fire truncation limit is 1.00E-11/yr for CDF and 1.00E-12/yr for LERF.

(3) %Delta CDF and LERF calculations were performed on non-rounded CDF and LERF results.

Table APLA RAI-02.2 - Impact on RICT - Factor of Two Increase for 1CNPVDWRUPT--R--

Tech Spec TS/LCO Condition Base Model RICT Estimate (days)(1)

Sensitivity RICT Estimate (days)

% Decrease 3.5.1.B One LPCI subsystem inoperable for reasons other than Condition A OR one Core Spray subsystem inoperable.

16.1 10.99 31.7%

3.5.1.G HPCI System inoperable 30 30 0%

3.5.3.A RCIC System inoperable 30 30 0%

3.7.9.A SSMP System inoperable 30 30 0%

3.8.4.E Opposite unit 125 VDC electrical power subsystem inoperable 5.5 5.41 1.6%

3.8.7.C One or more required opposite unit AC or DC electrical power distribution subsystems inoperable 5.5 5.41 1.6%

Note to Table APLA RAI-02.2:

(1) Includes seismic/high winds penalties and uses unfactored results from PRAQuant.

d) The Quad Cities, Unit 1 and Unit 2 PRA model does model failure of RPV injection post-containment failure. FW / Condensate / SBCS and CRD are credited for success after containment failure, but an additional basic event (1CNPVDWRUPT--R-- large DW containment failure causes loss of injection) is included that represents the likelihood that the containment failure size and location disrupts the capability of all RPV injection (e.g.,

FW/Condensate/SBCS and CRD) post-containment failure. The value of 1CNPVDWRUPTR-- is based on the estimated large DW failure probability of 6.0E-2 used in the model.

Although the NUREG-1150 analysis did not explicitly identify a probability for a large DW failure (i.e., rupture) for low temperature conditions, the QCNPS PRA conservatively estimated a large DW failure probability of 6.0E-2. The large DW failure basic event is conservatively modeled to result in guaranteed failure of all RPV injection following containment failure (e.g., for a Loss of DHR scenario).

APLA RAI-03 RG 1.200, Revision 2, states, in part: "The base PRA serves as the foundational representation of the as-built and as-operated plant necessary to support an application."

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 7 of 72 Docket Nos. 50-254 and 50-265 The TSTF-505 LAR does mention the existence of interconnected auxiliary systems between units. The NRC staff notes that for some of these systems, it appears the sharing of a system is not consistent between units. Enclosure 8 to the TSTF-505 LAR states that the Real Time Risk (RTR) model can represent the availability of these shared components. However, it is unclear to the NRC staff if all operational aspects, such as alternate alignments, were excluded from the PRA models. In addition, multi-unit events (e.g., loss of offsite power and seismic events), credit for a shared system may be limited to one unit.

Clarify what systems are shared, how they are shared, and whether they can support the other unit in an accident. Explain how the shared systems are credited for each unit in the PRA models. This discussion should also address the following:

a) Explain whether shared systems are credited in the internal events, including flood and fire PRA models for both units and, if so, identify those systems.

b) If shared systems are credited in the RTR model that supports the RICT calculations, then explain how the shared system is modelled for each unit in a dual unit event demonstrating that shared systems are not over-credited in the PRA models.

c) If a shared system is credited in the RTR model that supports the RICT calculations and the impact of events that can create a concurrent demand for the system shared by both units is not addressed in the PRA models, then justify that this exclusion has an inconsequential impact the RICT calculations.

d) Explain how the PARAGON CRMP model displays the availability of shared systems on the operator screen for both units. Confirm that the PARAGON tool considers the unavailability of a shared system for both units.

Constellation Response to APLA RAI-03 a) The Quad Cities Nuclear Power Station (QCNPS) PRA models, including internal events, internal flooding, and fire PRA, credit shared systems between Units 1 and 2. Specific credited systems in the PRA models are:

a. Service Water (SW) system

b. Fire Protection (FP) system

c. Safe Shutdown Makeup Pump (SSMP) system

d. Residual Heat Removal Service Water (RHRSW)

e. Instrument Air (IA) and Service Air (SA) systems

f.

Diesel Generator (DG) system

g. AC Power system

h. 125 VDC system

i.

250 VDC system b) As noted in a) above, shared systems are included in the Real Time Risk (RTR) models. As the RTR models communicate across units, any changes to alignments or common systems are reflected in both units' models.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 8 of 72 Docket Nos. 50-254 and 50-265 Service Water (SW)

The SW system has five shared pumps that feed a common header, and anywhere from one to four pumps are usually operating based on seasonal configurations. The Unit 1 and Unit 2 QCNPS RTR models both credit all five SW pumps as well as the service water configuration alignments associated with seasonal success criteria requirements. The PRA considers a complete loss of the service water system as a dual unit initiating event (loss offsite power to both units simultaneously causes a complete loss of service water).

The PRA does not evaluate partial losses of SW which may be dealt with by power derates.

Because the SW system is sized to supply both units simultaneously while those units are at full power, and because both units are served by a common header, the SW system is capable of (and credited with) supplying both units when one unit may have experienced an initiating event included in the PRA while the other unit is still operating at 100% power.

This is because the unit experiencing the shutdown will have SW requirements (i.e., heat loads) no greater than the requirements associated with 100% power, and in most cases the heat loads will be much lower in the unit experiencing the off-normal conditions.

Other than dual unit loss of power, or complete loss of service water (which results in initiating events at both units), a simultaneous loss of 125 VDC Buses 1 and 2 will also cause dual unit initiating events. The simultaneous losses of these buses do not by themselves impact the operation of the SW system to result in a complete loss of SW to both units.

Fire Protection (FP)

The FP system consists of two shared diesel-driven pumps that feed a common header.

The FP header can act as a redundant suction source to the SSMP if the Contaminated Condensate Storage Tank (CCST) becomes unavailable and can be directed via a spoolpiece to the RHR system piping for low pressure injection. The FP system is shared between both units and is credited only as an alternate injection source in the QCNPS RTR internal events models in extreme cases due to the system supplying low-quality river water.

Additionally, the FP system pumps are not affected by dual unit events as they are started and controlled by dedicated 24V batteries. The PRA does not evaluate the presence of a fire in one unit and a separate initiating event at the second unit (such scenarios are screened probabilistically), and therefore the potential for FP water sources being diverted from one unit to the other is not assessed.

Safe Shutdown Makeup Pump (SSMP)

The SSMP is a shared high pressure injection system that can be used as an alternate water injection source during accident conditions. Although the SSMP can be aligned to either unit, it can only supply injection to a single unit at a time. For dual unit events with no other makeup sources, the SSMP may need to be shared between units, which would require operators to swap the SSMP injection alignment between units. This operator action exists in the QCNPS PRA models through the inclusion of human failure events representing the failure of the action. A dual unit loss of onsite and offsite power prevents operation of the system.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 9 of 72 Docket Nos. 50-254 and 50-265 Residual Heat Removal Service Water (RHRSW)

RHRSW is not an explicitly shared system, however the capability to crosstie the opposite unit RHRSW system is modeled in the PRA. There is a cross connection via a single locked closed manual valve that is included in the PRA models through the inclusion of human failure events for the operator to implement the necessary actions to align the crosstie.

Instrument Air (IA) and Service Air (SA)

The IA system has a total of four IA compressors, two unit-specific air compressors and two shared compressors. The two shared IA compressors are normally in standby and can supply both units. A crosstie line between the Units 1 and 2 IA systems is normally closed and locked. A dual unit LOOP would cause all IA compressors to trip until power is restored or adequate diesel generator capacity is available. The SA system has three SA compressors and is modeled in the QCNPS RTR model as a whole system rather than by individual compressor. This is conservative modeling since the SA system is modeled in the PRA only for backup to the IA system and failure of any one of the SA compressors is assumed to fail the entire SA system as a backup source.

Diesel Generators (DG) and AC Power There are three emergency DGs, each capable of powering the largest postulated vital loads under postulated accident conditions: two dedicated unit DGs and a swing DG. These loads are automatically sequenced onto the requisite buses upon restoration of bus voltage by the DG system. A single DG (or a single station blackout DG) is required given a LOOP or dual LOOP event to supply necessary loads on a single emergency bus (for a single unit).

The swing diesel can be aligned to supply either, but not both, units. There is an interlock in place to prevent the swing DG from simultaneously feeding both emergency buses and potentially overloading the DG. A 50% probability is assumed for the probability that the swing DG aligns to the opposite unit and an operator action is ANDed to allow for the manual alignment to opposite unit.

There are also two station blackout (SBO) DGs, one for each unit, which are larger than the emergency DGs. For each unit, the PRA models the dedicated DG and dedicated SBO DG as the primary power supplies for a given unit, with the potential to cross-tie from one unit to another should the need arise. The PRA includes logic assessing the alignment of the swing diesel between the two units, with a human failure event representing failure of an operator action to align the swing DG to the unit requiring power.

The PRA models a station blackout as a complete loss of all DGs (three) and SBO DGs (two). This condition is the entry condition to initiating FLEX alignments.

125 and 250 VDC The 125 and 250 VDC systems are normally shared between Units 1 and 2. The Unit 1 125 VDC system provides power to its own turbine building main bus 125 VDC loads and reactor building 125 VDC loads, and also supplies 125 VDC power to the turbine building

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 10 of 72 Docket Nos. 50-254 and 50-265 reserve bus loads for Unit 2. Similarly, the Unit 2 125 VDC system provides power to its own turbine building main bus 125 VDC loads and reactor building 125 VDC loads, as well as the 125 VDC loads for the turbine building reserve buses for Unit 1. The Unit 1 250 VDC system provides power to its own turbine building 250 VDC MCC and reactor building 250 VDC MCC 1A loads and provides power to the Unit 2 reactor building 250 VDC MCC 2B loads. Similarly, the Unit 2 250 VDC system provides power to its own turbine building 250 VDC MCC and reactor building 250 VDC MCC 2A loads and provides power to the Unit 1 reactor building 250 VDC MCC 1B loads.

In the event of loss of an operating battery charger supplying a 125 VDC battery bus, the dedicated standby battery charger for that unit can be manually aligned to the battery bus in accordance with QCNPS operating procedures. In the event of loss of one of the dedicated battery chargers supplying a 250 VDC battery bus, the swing battery charger for that unit can be manually aligned to either of the 250 VDC battery buses in accordance with QCNPS operating procedures. In an SBO event, the only sources of power would be DC power and power from the FLEX DGs.

As described above, simultaneous losses of 125 VDC Buses 1 and 2 cause dual unit initiating events. The loss of 125 VDC Bus 1 by itself, and the loss of Bus 2 by itself, are included in the PRA as separate initiating events affecting only a single unit.

c) The impacts of events that can create a concurrent demand for the system shared by both systems are addressed in the PRA models.

d) While the RTR model applies separate models for QCNPS Units 1 and 2, each result group will use a shared, configuration-specific schedule. As mentioned in response b), the RTR models communicate across units which means that any changes to alignments or common systems are reflected in both units' models. All systems mentioned in response b) have been modeled in the RTR model as available to either unit via a shared schedule, and systems that require specific alignment to a single unit are also modeled in the RTR tool as such. When Operations toggles the unavailability of a system in the RTR tool, it automatically will apply to the shared Units 1 and 2 schedule. As such, if a system is modeled as shared this unavailability will be reflected in both units configuration-specific schedule.

APLA RAI-04 The Tier 3 assessment in RG 1.177, "An Approach for Plant-specific, Risk-informed Decision-making: Technical Specifications," Revision 2 (ML20164A034), stipulates that a licensee should develop a program that ensures the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06-09-A and its associated NRC SE (ML071200238) state that, for the impact of seasonal changes, either conservative assumptions should be made, or the PRA should be "adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration." of the TSTF-505 LAR mentions modifications related to the impact of outside temperatures for one structures, systems, and component (SSC). However, it does not appear

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 11 of 72 Docket Nos. 50-254 and 50-265 to state if other modeling adjustments are needed to account for seasonal and time of cycle dependencies and what kind of adjustments will be made. Therefore, address the following to clarify the treatment of seasonal and time of cycle variations:

a) Explain how the RICT calculations address changes in PRA data points, basic events, and SSC operability constraints as a result of extreme weather conditions, seasonal variations, other environmental factors, or time of cycle. Also, explain how these adjustments are made in the configuration risk management program (CRMP) model and how this approach is consistent with the guidance in NEI 06-09-A and its associated NRC final SE.

b) Describe the criteria used to determine when PRA adjustments due to extreme weather conditions, seasonal variations, other environmental factors, or time of cycle variations need to be made in the CRMP model and what mechanism initiates these changes.

c) Explain how the Quad Cities, Units 1 and 2, PRA accounts for seasonal variations in the modeling of affected components, and how these variations, if any, are reflected in the PARAGON CRMP (RTR) tool.

Constellation Response to APLA RAI-04 a) See response b) below for how extreme weather conditions are addressed procedurally at QCNPS. For the QCNPS model, the impact of outside temperatures on system requirements like seasonal service water pumps are evaluated in the Configuration Risk Management Program (CRMP) model with seasonal configuration alignments that apply a configuration probability based on operating experience. In the Real Time Risk (RTR) tool, Operations sets the actual seasonal configuration that is occurring based on the amount of service water pumps in operation. When Operations sets a seasonal configuration in the RTR model, the specific seasonal configuration is set to True in the PRA fault tree while the remaining seasonal configurations are set to False. This will ensure the proper success criteria is met for service water requirements in the CRMP model which is consistent with guidance in NEI 06-09-A that requires that the model either conservatively assess the seasonal variations or be adjusted to reflect current seasonal configuration.

b) Extreme temperature condition modeling is consistent with the plant design basis. There are no circumstances in which extreme weather conditions are addressed differently than the plant TS would allow.

Severe weather (high wind, severe thunderstorm warning, tornado watch/warning, solar magnetic grid disturbance) conditions that are a potential High Risk Evolution (HRE) for a loss of offsite power (LOOP) are proceduralized the Constellation work control procedures.

Appropriate risk management actions, such as deferring any AC power maintenance, may be taken as a result of triggering the HRE in the CRMP. In addition, a separate Operations procedure provides guidance for severe weather mitigation, such as restoring Emergency Core Cooling Systems (ECCS) or other system equipment to cope with severe weather, holding shift briefings for potential severe weather impacts, and reviewing LOOP procedures as well as guidance for specific threats such as removing potential on site missiles for high wind conditions or additional intake monitoring for blizzard conditions.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 12 of 72 Docket Nos. 50-254 and 50-265 c) The PRA model is representative of an average condition, and therefore adjustments to account for potential seasonal variations are not applied. The generic and plant-specific data used to calculate failure rates and probabilities are averages applicable throughout a reactor-year, and thus the PRA model does not adjust failure probabilities based on seasonal variations. Inputs to the LOOP initiating event frequency include impacts due to severe weather and, separately, extreme weather.

Seasonally adjusted success criteria are applied to the service water system. As mentioned in response a) above, the RTR tool applies a flag to the PRA model that sets the given seasonal configuration to True and the remaining seasonal configurations False in the PRA fault tree. The flag is determined by a toggle that Operations can set based on the number of service water pumps in operation.

APLA RAI-05 The NRC SE for NEI 06-09-A, states in part: "The impact of the proposed change should be monitored using performance measurement strategies." NEI 06-09-A considers the use of NUMARC 93-01, Revision 4F, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants" (ML18120A069), as endorsed by RG 1.160, "Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," Revision 4 (ML18220B281), for the implementation of the Maintenance Rule. NUMARC 93-01, Section 9.0, contains guidance for the establishment of performance criteria.

In addition, the NEI 06-09-A methodology satisfies the five key safety principles specified in RG 1.177, Revision 2 relative to the risk impact due to the application of a RICT. Moreover, NRC staff position C.3.2 provided in RG 1.177, Revision 2, for meeting the fifth key safety principle acknowledges the use of performance criteria to assess degradation of operational safety over a period. It is unclear how the licensee's RICT program captures performance monitoring for the SSCs within the scope of the RMTS program. Therefore:

a) Confirm that the Quad Cities, Units 1 and 2, Maintenance Rule program incorporates the use of performance criteria to evaluate SSC performance as described in NUMARC 93-01, as endorsed by RG 1.160.

b) Alternatively, describe the approach or method used for SSC performance monitoring, as described in NRC staff position C.3.2 of RG 1.177, Revision 2, for meeting the fifth key safety principle. In the description, include criteria (e.g., qualitative, or quantitative) along with the appropriate risk metrics and explain how the approach and criteria demonstrate the intent to monitor the potential degradation of SSCs in accordance with the NRC SE for NEI 06-09-A.

Constellation Response to APLA RAI-05 a) QCNPS, like all operating Constellation sites, follows NEI 18-10, "Monitoring the Effectiveness of Nuclear Power Plant Maintenance," guidance for meeting the requirements of 10 CFR 50.65 (Maintenance Rule). This is an alternative to the NRC-endorsed guidance

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 13 of 72 Docket Nos. 50-254 and 50-265 in NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." Therefore, the response will address option (b) above. NEI 18-10 differs from NUMARC 93-01 guidance primarily regarding how (a)(2) structures, systems, and component (SSC) functions are managed.

Regarding the Risk-Informed Completion Time (RICT) program, NEI 06-09, "Risk-Informed Technical Specifications Initiative 4b: Risk-Managed Technical Specifications (RMTS)

Guidelines Industry Guidance Document," outlines the requirements that must be followed to allow extension of allowable completion times. This includes evaluating the acceptability of a single completion time extension and the cumulative risk contribution based on the extension of all RICT windows throughout a 24-month period. The RICT program manages risk through use of several risk metrics and also through the use of Risk Management Actions (RMAs). The risk metrics are described in NEI 06-09 and include limits while in a RICT, limits to prevent entry into potential high risk configurations, and cumulative tracking limits imposed to assure that the guidance of RG 1.174, Revision 1, is met. These limits aid in minimizing the impact on plant safety.

In the NRC Safety Evaluation (SE) for NEI Topical Report (TR) 06-09 (ADAMS Accession No. ML071200238, dated May 17, 2007), the five key safety principles of risk-informed decision making presented in RG 1.174, Revision 1, for risk-informed applications are addressed including the fifth key safety principle:

"The impact of the proposed change should be monitored using performance measurement strategies."

As stated in the SE, the cumulative impact of implementation of a RMTS is periodically assessed and must be shown to result in a total risk impact below certain values (on an annual basis). The SE concludes that these criteria are consistent with the guidance of RG 1.174, Revision 1, for acceptable small changes in risk. The SE also acknowledges that "the NRC staff anticipates that the use of extended CTs [Completion Times] within an RMTS program is unlikely to be a routine practice, since licensees already accomplish planned maintenance activities within the existing TS CTs." Furthermore, the SE states:

Although the RMTS are permitted to be applied to planned maintenance activities, other requirements, such as 10 CFR 50.65 performance monitoring, and regulatory oversight of equipment performance, are disincentives to a licensee for incurring significant additional unavailability of plant equipment, even when allowed by an RMTS program. This provides a further control on the use of the RMTS which could result in a significant increase in equipment unavailability and the commensurate risk.

The SE then considers a single CT extension which could (alone) approach the risk limits of NEI 06-09, but acknowledges that while allowable, such configurations are not routinely encountered during plant maintenance activities and are not the anticipated application of the RMTS. The SE concludes that:

"...the performance monitoring and feedback specified in the TR, is sufficient to reasonably assure changes in risk due to the implementation of the RMTS are small

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 14 of 72 Docket Nos. 50-254 and 50-265 and are consistent with Section 3.2 of RG 1.177. Thus, the fifth key safety principle of RG 1.177 is satisfied."

Demonstration that the SSCs within-scope of the RICT application remain capable of performing their intended functions is addressed by NEI 18-10 guidance, which includes measures to prevent incurring significant additional unavailability of plant equipment and analyzes equipment failures in the context of maintenance program effectiveness. The approach/method used by QCNPS (and all operating Constellation plants) for demonstrating that SSCs remain capable of performing their intended functions includes an examination of Core Damage Frequency (CDF) trends. NEI 18-10, Section 9.1.3, and Constellation procedure, ER-AA-320-1007, "Maintenance Rule 18 Periodic (a)(3) Assessment,"

require review of CDF trends over the assessment period for the purpose of ensuring a proper balance of SSC availability and reliability, as required by 10 CFR 50.65 paragraph (a)(3). While cumulative risk tracking (specifically intended for RICT) examines the incremental risk (above the front stop) for RICT entries on a 24-month basis, CDF trending (performed for broader Maintenance Rule purposes) examines the aggregate risk of all online work - not just RICT window entries and is performed as a 12-month rolling average for Maintenance Rule purposes.

The results of CDF trending are addressed in the periodic (a)(3) assessment of the effectiveness of maintenance actions, performed once per fuel cycle. This assessment is required by 10 CFR 50.65, paragraph (a)(3) which states (in part):

"... ensure that the objective of preventing failures of structures, systems, and components through maintenance is appropriately balanced against the objective of minimizing unavailability of structures, systems, and components due to monitoring or preventive maintenance."

CDF trending (for Maintenance Rule purposes) examines the risk impact associated with both planned and unplanned maintenance and considers the impact of failures through the associated unplanned maintenance. CDF trending also provides an aggregate assessment of maintenance planning and execution. The Internal Events PRA model (including internal flooding) is used for CDF trending. External events such as fire, seismic and external flooding are excluded because they are not explicitly quantified in the (a)(4) process. The calculated aggregate risk is compared to the annual average base CDF. The CDF trend evaluation is then used to perform the required periodic assessment in accordance with Engineering procedure, "Maintenance Rule 18 Periodic (a)(3) Assessment." The process requires:

obtaining the CDF trends for the assessment period from the PARAGON configuration risk management tool, and then evaluating fluctuations in the trend.

CDF trends are reviewed during the periodic (a)(3) assessment for a minimum of:

long unavailability durations, peak periods of risk increase, need to update PRA, and

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 15 of 72 Docket Nos. 50-254 and 50-265 multiple occurrences of the same configuration due to ineffective maintenance.

Any such fluctuations found are then examined for purposes of identifying multiple occurrences of the same unplanned configuration due to ineffective maintenance, or an imbalance in planned maintenance activities per the maintenance strategy to unplanned events requiring corrective maintenance activities. Excessive instances of long unavailability windows and/or frequent extension of completion times are indicative of an ineffective maintenance strategy. If any concerns are identified, an Issue Report (IR) is generated in the Corrective Action Program (CAP) to evaluate cause and an (a)(1) determination is performed. This will lead to SSC functions moving to (a)(1) monitoring requirements and goal setting.

NEI 18-10 guidance handles reliability as follows. If an event or failure occurs and an IR is generated in CAP associated with a scoped in SSC with High Safety Significant (HSS) function(s), the IR will be reviewed for HSS Maintenance Rule Functional Failures (MRFF).

Any HSS MRFF will result in an immediate (a)(1) determination (i.e., every HSS function has an equivalent of a reliability performance criterion value of 0). All IRs that represent a Plant Level Event (PLE) will result in an immediate (a)(1) determination. For Low Safety Significant (LSS) functions the reliability is monitored by evaluation of system performance trends. When a trend in system/function performance is observed, this would drive an immediate (a)(1) determination. Trends are identified on an ongoing/continuous basis by identification through engineer review, through operating experience review, or during the (a)(3) assessment. LSS trending is taking system health inputs (e.g., IRs, degraded conditions, preventive and predictive maintenance results, etc.) and determining if issues are occurring repeatedly such that the maintenance program is ineffective. There is no set limit or number to constitute a trend. While engineers perform real-time trend reviews, an additional review is performed by the Maintenance Rule Coordinator as part of the (a)(3) assessment.

In this fashion, 10 CFR 50.65 performance monitoring complements the RICT program, and ensures that significant additional unavailability of plant equipment leading to a degradation of plant safety will not be incurred and, therefore, meeting the fifth key safety principle of RG 1.177.

b) Not applicable. See a) above.

APLA RAI-06 The NRC's safety evaluation for NEI 06-09-A specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis for when there is a difference. Table E1-1 of the TSTF-505 LAR, Enclosure 1, identifies each Limiting Condition for Operation (LCO) in the TSs proposed for inclusion in the RICT program. The table also describes whether the systems and components covered by the LCO are modeled in the PRA and, if so, presents both the design success criteria and PRA success criteria. For certain LCOs, the table explains that the associated SSCs are not modeled in the PRAs but will be represented using a surrogate event that fails the function performed by the SSC. For some

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 16 of 72 Docket Nos. 50-254 and 50-265 LCOs, the LAR did not provide an adequate description for the NRC staff to conclude that the PRA modeling will be sufficient.

a) Regarding TS LCO 3.3.5.1.B, Table E1-1 states that, for emergency core cooling system (ECCS) actuation instrumentation for core spray (CS), low pressure coolant injection (LPCI), high pressure coolant injection (HPCI), and diesel generators (DGs) that SSCs associated with Functions 3.a and 3.b are not explicitly modeled in the PRA. However, the associated Table E1-1 comments entry states that all of the SSCs can be explicitly modeled in the RTR tool. The NRC staff notes that the Functions 3.a and 3.b appear to be associated with HPCI instrumentation. It is unclear to the staff what system is associated with Functions 3.a and 3.b and if this associated instrumentation is incorporated into the RTR tool.

i.

Clarify the system(s) associated with TS LCO 3.3.5.1.B Functions 3.a and 3.b.

Include in this response if the actuation instrumentation related to these two functions are explicitly modeled in the RTR tool.

ii.

If the actuation instrumentation related to Functions 3.a and 3.b of TS LCO 3.3.5.1.B are not explicitly modeled in the RTR tool, then provide the following information:

a. Identify the RTR model surrogates to be used for Functions 3.a and 3.b of TS LCO 3.3.5.1.B

b. Provide justification that the surrogate(s) is related and bounds both Functions 3.a and 3.b of TS LCO 3.3.5.1.B.

b) Regarding TS LCO 3.3.5.1.C, Table E1-1 states that, for ECCS actuation instrumentation for CS, LPCI, and DGs that SSCs associated with Functions 3.c and 3.g are not explicitly modeled in the PRA. However, the associated Table E1-1 comments entry states that all of the SSCs can be explicitly modeled in the RTR tool. It is unclear to the staff what system is associated with Functions 3.c and 3.g and if the associated instrumentation is in the RTR tool.

i.

Clarify the system(s) associated with TS LCO 3.3.5.1.C Functions 3.c and 3.g.

Include in this response if the actuation instrumentation related to these two functions are explicitly modeled in the RTR tool.

ii.

If the actuation instrumentation related to Functions 3.c and 3.g of TS LCO 3.3.5.1.C are not explicitly modeled in the RTR tool, then provide the following information:

a. Identify the RTR model surrogates to be used for Functions 3.c and 3.g of TS LCO 3.3.5.1.C

b. Provide justification that the surrogate(s) is related and bounds both Functions 3.c and 3.g of TS LCO 3.3.5.1.C.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 17 of 72 Docket Nos. 50-254 and 50-265 c) Regarding TS LCO 3.6.1.2.C, Table E1-1 states that, for primary containment air locks not modeled, that a large pre-existing containment isolation failure that is modeled will be used as a surrogate. It is unclear to the staff how pre-existing containment isolation failure is either conservative or bounding.

Provide justification that the surrogate conservatively bounds TS LCO 3.6.1.2.C. Explain whether this justification was specifically modeled or included as an assumption.

d) Regarding TS LCO 3.6.1.3.A, Table E1-1 states that, for primary containment isolation valves not modeled, that a large pre-existing containment isolation failure that is modeled will be used as a surrogate. It is unclear to the staff how pre-existing containment isolation failure is either conservative or bounding.

Provide justification that the surrogate conservatively bounds TS LCO 3.6.1.3.A.

Constellation Response to APLA RAI-06 a) Functions 3.a and 3.b are associated with High Pressure Coolant Injection (HPCI) system initiation due to Reactor Vessel Water Level Low Low signals and Drywell Pressure High signals. The actuation instrumentation for these functions is not explicitly modeled in the Real Time Risk (RTR) model. These functions are represented instead by conservatively failing the HPCI turbine-driven pump, which would prevent automatic initiation of the system.

b) Functions 3.c and 3.g are associated with tripping the HPCI system initiation due to Reactor Vessel Water Level High signals and preventing manual initiation. The instrumentation for these functions is not explicitly modeled in the Real Time Risk (RTR) model. These functions are represented instead by conservatively failing the HPCI turbine-driven pump, which would prevent availability of the system for any accident scenarios and prevent manual initiation of the system.

c) The Quad Cities PRA basic event for a large pre-existing containment isolation failure is modeled in the Level 2 PRA as a direct bypass of the primary containment and is modeled to lead directly to a Large Early Release Frequency (LERF) scenario. Unavailability of a primary containment air lock would be modeled with the same consequence as the large pre-existing containment isolation failure in the Level 2 PRA model. Therefore, modeling a large pre-existing containment isolation failure is an appropriate surrogate for modeling unavailability of a primary containment air lock for TS LCO 3.6.1.2.C.

d) See response to (c) above. Therefore, modeling a large pre-existing containment isolation failure is an appropriate surrogate for modeling unavailability of primary containment isolation valves for TS LCO 3.6.1.3.A.

APLA RAI-07 Section 2 of Enclosure 9 of the RICT LAR states that FLEX is credited in the Quad Cities internal events PRA (FPIE), which includes internal flooding, and the FPRA.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 18 of 72 Docket Nos. 50-254 and 50-265 NRC memorandum dated May 6, 20221 provides the NRC's staff updated assessment of identified challenges and strategies for incorporating Diverse and Flexible Mitigation Capability (FLEX) equipment into a PRA model in support of risk-informed decisionmaking in accordance with the guidance of RG 1.2002. The NRC staff states in Conclusion 4 of the memo: "Licensees that choose not to use the generic failure probabilities in PWROG-18042 to develop plant-specific failure probabilities for portable FLEX equipment modeled in PRAs used for risk-informed applications should submit a justification for the methods and probabilities used to the NRC for review and approval."

It appears that NUREG-6928 fixed equipment failure rates, including a 2x increase, were used as probabilities for FLEX portable equipment. It is unclear to the NRC staff how the Quad Cities, Units 1 and 2, approach satisfies the concerns of Conclusion 4.

The enclosure explains how NRC's "Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Probabilistic Risk Assessments" (ML22014A084) is addressed in the modeling of FLEX.

Address the following:

a) Propose a mechanism to incorporate updated FLEX parameter values in accordance with PWROG-18042-N into the Quad Cities PRA models used for RICT calculations prior to implementing the RMTS program.

-OR-Alternatively, identify the LCO conditions impacted by the treatment of this modelling uncertainty for which RMAs will be applied during a RICT. Include discussion of the kinds of RMAs that would be applied and justification that the RMAs will be sufficient to address the modeling uncertainty.

b) Provide a discussion detailing the methodology used to assess operator actions related to installation and operation of FLEX equipment. The discussion should include:

i.

A list of the FLEX-related operator actions and a summary description of the plant-specific HRA used as the basis to develop the HEPs for each operator action.

Include an evaluation of the HRA associated with declaration of Extended Loss of AC Power (ELAP).

ii. If the FLEX-related HRA is not in accordance with the NRC memorandum dated May 6, 2022, justification that the HRA assumptions have an inconsequential impact on the RICT calculations.

1 U.S. NRC memorandum, "Updated Assessment of Industry Guidance for Crediting Mitigating Strategies in Risk Assessments," dated May 6, 2022 (ML22014A084).

2 U.S. NRC, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," RG 1.200, Revision 3, December 2020 (ML20238B871).

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 19 of 72 Docket Nos. 50-254 and 50-265 iii. If, in response to part iii) above, it cannot be determined that the cited assumptions have an inconsequential impact on the estimated RICTs, then identify the LCO conditions impacted by the treatment of this modelling uncertainty for which RMAs will be applied during a RICT. Include a discussion of the programmatic changes that the licensee will consider in order to compensate for this uncertainty and the basis for their consideration (e.g., identification of additional RMAs and justification that they are sufficient to address the modeling uncertainty).

c) If the PRA modeling of FLEX equipment and/or operator actions is revised or updated to be in accordance with the NRC memorandum dated May 6, 2022, provide justification that the revisions do not meet the definition of an PRA upgrade as defined by RG 1.200.

-OR-Alternatively, if justification cannot be provided, propose a mechanism to conduct a focused-scope peer review (FSPR) regarding incorporation of the PWR Owners Group FLEX equipment reliability modeling and/or EPRI FLEX HRA methodology for the ANO-2 PRA models. Include in the mechanism to close out all Facts and Observations (F&Os) that result from the FSPR prior to implementing the RMTS program.

Constellation Response to APLA RAI-07 a) At the time of preparation of the PRA used in support of the license amendment request (LAR) submittal the PWROG data did not exist. The multiplier methodology was used in several PRAs to represent the possible uncertainty associated with FLEX equipment reliability data.

The sensitivity performed for the RICT evaluation credit for FLEX operation in response to postulated full power internal events was removed completely through the use of "FLAG" events embedded in the PRA logic model. These settings have the effect of "failing" FLEX with a probability of 1.0. As documented in the LAR, the impact on the PRA results is very small in terms of any change in the CDF or LERF as compared to the baseline. This change would be small regardless of the FLEX parameter values used for the baseline model and quantification.

The tables which follow present the results of the sensitivity analysis and the impact on selected RICT cases. Table APLA RAI-07.1 shows the change in CDF and LERF due to removing credit for FLEX in the FPIE and fire PRAs, based on a zero maintenance model. Table APLA RAI-07.2 shows the change in RICT for a selected group of TS; these TS were selected because the related systems are associated with PRA station blackout scenarios. Both tables demonstrate the insensitivity of the results to uncertainty in the parameter values and HEPs used for FLEX.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 20 of 72 Docket Nos. 50-254 and 50-265 Table APLA RAI-07.1 - CDF and LERF Sensitivity Results - No FLEX Credit Model / Case CDF (/yr)

%Delta CDF(3)

LERF (/yr)

%Delta LERF(3)

FPIE(1)

Base 3.45E-06 2.22E-07 Sensitivity 3.46E-06 0.3%

2.22E-07 0.0%

Fire(2)

Base 3.43E-05 2.99E-06 Sensitivity 3.43E-05 0.0%

3.00E-06 0.3%

Total Base 3.78E-05 3.21E-06 Sensitivity 3.78E-05 0.0%

3.22E-06 0.3%

Notes to Table APLA RAI-07.1:

(1) FPIE truncation limit is 1.00E-11/yr for CDF and 1.00E-12/yr for LERF.

(2) Fire truncation limit is 1.00E-11/yr for CDF and 1.00E-12/yr for LERF.

(3) %Delta CDF and LERF calculations were performed on non-rounded CDF and LERF results.

Table APLA RAI-07.2 - Impact on RICT - No FLEX Credit Tech Spec TS/LCO Condition Base Model RICT Estimate (days)

Sensitivity RICT Estimate (days) 3-5-1-G_2 HPCI System inoperable 30 30 3-5-3-A_1 RCIC System inoperable 30 30 3-6-2-3-A_1 HPCI System inoperable 30 30 3-7-9-A_

SSMP System inoperable 30 30 3-8-1-C_1 Two required offsite circuits inoperable.

15.34 15.28 3-8-7-C_4 One or more required opposite unit AC or DC electrical power distribution subsystems inoperable 27.49 27.41 NOTE:

TS for a single DG out of service are not shown in Table APLA RAI-07.2; there is no change in RICT for those TS.)

Therefore, the conclusions regarding insensitivity of the results to the FLEX parameter values remain valid with or without use of the PWROG values. Because there is very minimal impact on the results there are no LCO conditions impacted by this modeling uncertainty and no additional RMAs are required.

b) i)

The following table (Table APLA RAI-07.3) lists the FLEX-related FPIE PRA and Fire PRA human failure events (HFEs) included in the PRA models used in support of the RICT calculations. The table contains the human error probabilities (HEPs) used in the analyses.

ii) The screening values used are not in accordance with the memorandum. However, the HRA assumptions (i.e., the screening values) have an inconsequential impact on RICT calculations, as described in the response to 7 a).

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 21 of 72 Docket Nos. 50-254 and 50-265 iii) As described above, the sensitivity performed for FLEX removed all credit for FLEX in response to any postulated full power internal event or fire event. Thus, the FLEX related HFEs and their HEPs have no impact on the sensitivity results; the sensitivity has the same effect as setting all HEPs to 1.0 in addition to setting failure probabilities of FLEX equipment to 1.0. The estimated RICTs are insensitive to uncertainty in FLEX parameter values and HEPs.

c) There are no plans to update the FLEX data or HEPs prior to RICT implementation.

Table APLA RAI-07.3 - FLEX-Related Human Failure Events, FPIE and Fire HFE Basic Event Identifier HFE Description HEP Comments FPIE 1FXOP-BATCHRGH--

OP ACT: FAILURE TO PROVIDE FLEX PWR TO 125 AND 250 VDC BAT CHRGS 1.00E-01 All HEPs are Screening Values 1FXOPFLEXALT-H--

OP ACT: FAILURE TO ALIGN ALT FLEX PMP WITH PRIMARY FAILED 5.00E-01 1FXOPFLEX-PMPH--

OP ACT: FAILURE TO INITIATE FLEX PUMP STRATEGY W/IN 1 HOUR 1.00E-01 1FXOPFLXRMCLGH--

OP ACT: FAILURE TO INITIATE ALTERNATE FLEX RCIC ROOM COOLING 1.00E-01 1FXOP-HCVSPNLH--

OP ACT: FAILURE TO VENT USING HVCS PNL 1.00E-03 1FXOP-LPI----H--

OP ACT: FAILURE TO INITIATE LOW PRESSURE INJECTION 1.00E-01 1FXOP-LPI-L2-H--

OP ACT: FAILURE TO INITIATE LOW PRESSURE INJECTION (LEVEL 2) 1.00E-01 1FXOPVNTLOCALH--

OP ACT: OPERATORS TAKE LOCAL ACTIONS GIVEN SUPPORT FAILURE 1.00E+00 1FXOPWELLMU--H--

OP ACT: FAILURE TO PROVIDE MU FROM WELL PUMPS 1.00E-01 FIRE 1FXOP-BATCHRGH-F OP ACT: FAILURE TO PROVIDE FLEX PWR TO 125 AND 250 VDC BAT CHRGS (Fire) 1.00E-01 All HEPs are Screening Values 1FXOPFLXRMCLGH-F OP ACT: FAILURE TO INITIATE ALTERNATE FLEX RCIC ROOM COOLING (Fire) 1.00E-01 1FXOP-HCVSPNLH-F OP ACT: FAILURE TO VENT USING HVCS PNL (Fire) 1.00E-01 1FXOP-LPI----H-F OP ACT: FAILURE TO INITIATE LOW PRESSURE INJECTION (Fire) 1.00E-01

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 22 of 72 Docket Nos. 50-254 and 50-265 Table APLA RAI-07.3 - FLEX-Related Human Failure Events, FPIE and Fire HFE Basic Event Identifier HFE Description HEP Comments 1FXOP-LPI-L2-H-F OP ACT: FAILURE TO INITIATE LOW PRESSURE INJECTION (LEVEL 2)

(Fire) 1.00E-01 c) There are no plans to update the FLEX data or HEPs prior to RICT implementation.

APLB RAI-01 The TSTF-505 LAR, Enclosure 2, provides the history of the Fire PRA (FPRA) peer review but does not appear to discuss methods used in the FPRA. Methods may have been used in the FPRA that deviate from guidance in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities," (ML052580075, ML052580118, and ML103090242), or other acceptable guidance (e.g., frequently asked questions (FAQs), NUREGs, or interim guidance documents).

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

b) If such deviations exist, then justify their use in the FPRA and impact on the RICT program.

c) As an alternative to item b above, add an implementation item to replace those methods with a method acceptable to NRC prior to the implementation of the RICT program.

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

Constellation Response to APLB RAI-01 a) The fire PRA does not use methods that deviate from the guidance in NUREG/CR-6850 or other acceptable guidance.

b) Not applicable per the response to a) above.

c) Not applicable per the response to a) above.

APLB RAI-02 The key factors used to justify using transient fire reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, letter from Joseph Giitter, U.S. Nuclear Regulatory Commission, to Biff Bradley, NEI, "Recent Fire PRA Methods

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 23 of 72 Docket Nos. 50-254 and 50-265 Review Panel Decisions and Electrical Power Research Institute (EPRI) 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires'." (ML12172A406).

If any reduced transient HRRs below the bounding 98% HRR of 317 kW from NUREG/CR-6850 were used, discuss the key factors used to justify the reduced HRRs. Include in this discussion:

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

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

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

Constellation Response to APLB RAI-02 a) The FPRA does not credit a reduced transient fire HRR as discussed in the referenced letter.

The QCNPS FPRA uses the Table 8-1 NUREG-2233, "Methodology for Modeling Transient Fires in Nuclear Power Plant Fire Probabilistic Risk Assessment," generic transient fire distribution with a 98% heat release rate (HRR) of 278 kW for postulated transient fires. The transient fire HRR distributions in Table 8-2 for transient combustible control location (TCCL) transient fires were not used.

b) Not applicable per the response to a) above.

c) Not applicable per the response to a) above.

APLB RAI-03 FAQ 13-0004, "Clarifications on Treatment of Sensitive Electronics" (ML13322A085) provides supplemental guidance for application of the damage criteria provided in Sections 8.5.1.2 and H.2 of NUREG/CR-6850, Volume 2, for solid-state and sensitive electronics.

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

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

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 24 of 72 Docket Nos. 50-254 and 50-265 c) As an alternative to item b above, add an implementation item to replace the current approach with an acceptable approach prior to the implementation of the RICT program.

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

Constellation Response to APLB RAI-03 a) The treatment of sensitive electronics for the fire PRA (FPRA) is consistent with the guidance in frequently asked question (FAQ) 13-0004.

The damage thresholds considered for sensitive electronics were:

x Exterior surface mounted - Heat flux of 3 kW/m2 and temperature of 65°C x

Enclosed - heat flux of 11 kW/m2 x

Enclosed - Temperature of 65°C Plant walkdowns were performed and the fire areas were reviewed to ensure the FPRA treatment bounded the potential for sensitive electronics (e.g., the assumed initiator bounds feedwater digital controls) or the fire modeling included sensitive electronics using the damage thresholds above including consideration for surface mounted sensitive electronics and the presence of louver or vents.

b) Not applicable per the response to a) above.

c) Not applicable per the response to a) above.

APLB RAI-04 NUREG-1921, "EPRI/NRC-RES Fire Human Reliability Analysis Guidelines-Final Report,"

(ML12216A104), discusses the need to consider a minimum value for the joint probability of multiple human failure events (HFEs) in human reliability analyses (HRAs).

NUREG-1921 refers to Table 2-1 of NUREG-1792, "Good Practices for Implementing Human Reliability Analysis (HRA)," (ML051160213), which recommends that joint human error probability (HEP) values should not be below 1E-5. Table 4-4 of EPRI 1021081, "Establishing Minimum Acceptable Values for Probabilities of Human Failure Events," provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. Therefore, the guidance in NUREG-1921 allows for assigning joint HEPs that are less than 1E-5, but only through assigning proper levels of dependency.

The TSTF-505 LAR, Enclosure 9, Table E9-1 in regarding joint human error probabilities (JHEPs) in the FPRA, states that a sensitivity study was performed since the analysis implements a JHEP floor of 1E-06. The sensitivity study used the JHEP floor value of 1E-05, which is consistent with industry guidance. The LAR states that the sensitivity demonstrated that this source of uncertainty had a slight impact on FPRA results. However, during the staff's portal review of the Quad Cities, Units 1 and 2, FPRA Sensitivity Analysis Notebook, it was

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 25 of 72 Docket Nos. 50-254 and 50-265 noted that the sensitivity results provided in Section 4.2.3 demonstrates a 9 percent impact in overall increase in core damage frequency (CDF) and large early release frequency (LERF) for Quad Cities, Unit 1. In addition, Section 8.8 of the Assessment of Key Assumptions and Sources of Uncertainty Notebook demonstrates a fourteen percent impact on TS LCO 3.8.7.C.4 RICT calculation and provides results for only seven other TS LCO proposed for the RICT program. It is unclear to the NRC staff that this assumption has no impact on the RICT program. Therefore, address the following:

a) Provide justification, such as a RICT sensitivity study that the that the minimum joint HEP value has no impact on the remaining TS LCOs proposed for the RICT application.

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

i.

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

ii.

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

Constellation Response to APLB RAI-04 a)

An application specific sensitivity study was performed. The sensitivity study presented in Section 8.8 of the Assessment of Key Assumptions and Sources of Uncertainty Notebook focus was the fire probabilistic risk assessment (FPRA) and not performed consistent with the sample calculations presented in Enclosure 1 of the license amendment request (LAR). Table APLB RAI-04.1 presents sensitivity results consistent with the Enclosure 1 sample calculations.

Table APLB RAI-04.1 shows that using a FPRA joint human error probability (JHEP) floor value of 1E-5 results in no more than a 1 day change in the calculated risk-informed completion time (RICT) for a Technical Specification (TS).

The only TS 3.8.7.A case 4 and 3.8.7.C case 4 sensitivity cases (both estimated RICT when an opposite unit AC power division is unavailable) resulted in more than a 1 day change in the estimated RICT. Each of these cases resulted in a 3.2 day reduction in RICT (24.7 days to 21.5 days), which is considered inconsequential given the estimated RICT for these cases and since all of the other cases for TS 3.8.7.A and 3.8.7.C led to no impact or a negligible impact.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 26 of 72 Docket Nos. 50-254 and 50-265 Table APLB RAI-04.1 FPRA JHEP Sensitivity Study Results TS LCO Condition LAR RICT Estimate FPRA JHEP Sensitivity RICT Estimate Change in RICT Estimate 3.1.7.A Standby Liquid Control (SLC) System -

One SLC subsystem inoperable.

30.0 30 0

3.3.1.1.A Reactor Protection System (RPS)

Instrumentation - One or more required channels inoperable.

2.3 2.3

< 0.1 3.3.1.1.B Reactor Protection System (RPS)

Instrumentation - One or more Functions with one or more required channels inoperable in both trip systems.

2.3 2.3

< 0.1 3.3.2.2.A Feedwater System and Main Turbine High Water Level Trip Instrumentation -

One or more Feedwater System and main turbine high water level trip channels inoperable.

30.0 30 0

3.3.4.1.A Anticipated Transient Without Scran Recirculation Pump Trip (ATWS-RPT)

Instrumentation - One or more channels inoperable.

30.0 30 0

3.3.5.1.B Emergency Core Cooling System (ECCS)

Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.1-1.

30.0 30 0

3.3.5.1.C ECCS Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.1-1.

30.0 30 0

3.3.5.1.D ECCS Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.1-1.

30.0 30 0

3.3.5.1.E ECCS Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.1-1.

30.0 30 0

3.3.5.1.F ECCS Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.1-1.

30.0 30 0

3.3.5.1.G ECCS Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.1-1.

30.0 30 0

3.3.5.3.B Reactor Core Isolation Cooling (RCIC)

System Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.3-1.

24.6 24.5

<0.1 3.3.5.3.D RCIC System Instrumentation - As required by Required Action A.1 and referenced in TS Table 3.3.5.3-1.

30.0 30 0

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 27 of 72 Docket Nos. 50-254 and 50-265 Table APLB RAI-04.1 FPRA JHEP Sensitivity Study Results TS LCO Condition LAR RICT Estimate FPRA JHEP Sensitivity RICT Estimate Change in RICT Estimate 3.3.6.1.A Primary Containment Isolation Instrumentation - One or more required channels inoperable 30.0 29.5 0.52 3.3.6.3.A Relief Valve Instrumentation - One relief valve inoperable due to inoperable channel(s )

30.0 30 0

3.3.8.1.A Loss of Power (LOP) Instrumentation -

One or more channels inoperable.

30.0 30 0

3.4.3.A Safety and Relief Valves - One relief valve inoperable.

30.0 30 0

3.5.1.B ECCS-Operating - One LPCI subsystem inoperable for reasons other than Condition A OR one Core Spray subsystem inoperable.

16.1 16.1

< 0.1 3.5.1.C ECCS-Operating - One LPCI pump in each subsystem inoperable.

30.0 30 0

3.5.1.E ECCS-Operating - Two LPCI subsystems inoperable for reasons other than Condition C.

30.0 30 0

3.5.1.G ECCS-Operating - HPCI System inoperable.

30.0 30 0

3.5.1.H ECCS-Operating - One ADS valve inoperable.

30.0 30 0

3.5.3.A RCIC System - RCIC System inoperable.

30.0 30 0

3.6.1.2.C Primary Containment Air Lock - Primary containment air lock inoperable for reasons other than Condition A or B.

6.0 5.8 0.19 3.6.1.3.A Primary Containment Isolation Valves (PCIVs) - One or more penetration flow paths with one PCIV inoperable for reasons other than Condition D.

6.0 5.8 0.19 3.6.1.6.A Low set relief valves - One low set relief valve inoperable.

30.0 30 0

3.6.1.7.A Reactor Building-to-Suppression Chamber Vacuum Breakers - One or more lines with one reactor building-to-suppression chamber vacuum breaker not closed.

N/A N/A N/A

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 28 of 72 Docket Nos. 50-254 and 50-265 Table APLB RAI-04.1 FPRA JHEP Sensitivity Study Results TS LCO Condition LAR RICT Estimate FPRA JHEP Sensitivity RICT Estimate Change in RICT Estimate 3.6.1.7.C Reactor Building-to-Suppression Chamber Vacuum Breakers - One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

N/A N/A N/A 3.6.1.8.A Suppression Chamber-to-Drywell Vacuum Breakers - One required suppression chamber-to-drywell vacuum breaker inoperable for opening.

3.5 3.5

< 0.1 3.6.2.3.A Residual Heat Removal (RHR)

Suppression Pool Cooling - One RHR suppression pool cooling subsystem inoperable.

30.0 30 0

3.6.2.6.A RHR Drywell Spray - One RHR drywell spray subsystem inoperable.

30.0 30 0

3.7.1.B Residual Heat Removal Service Water (RHRSW) System - One RHRSW pump in each subsystem inoperable.

30.0 30 0

3.7.1.C RHRSW System - One RHRSW subsystem inoperable for reasons other than Condition A.

24.8 24.8

<0.1 3.7.9.A Safe Shutdown Makeup Pump (SSMP)

System - SSMP System inoperable 30.0 30 0

3.8.1.A AC Sources-Operating - One required offsite circuit inoperable.

14.5 14.4

<0.1 3.8.1.B AC Sources-Operating - One required DG inoperable.

30.0 30 0

3.8.1.C AC Sources-Operating - Two required offsite circuits inoperable.

14.4 14.3

<0.1 3.8.1.D AC Sources-Operating - One required offsite circuit inoperable AND one required DG inoperable.

4.8 4.8

< 0.1 3.8.4.A DC Sources-Operating - One 250 VDC electrical power subsystem inoperable.

30.0 30 0

3.8.4.B DC Sources-Operating - Division 1 or 2 125 VDC battery inoperable as a result of maintenance or testing.

14.8 14.7

<0.1 3.8.4.C DC Sources-Operating - Division 1 or 2 125 VDC battery inoperable, due to the need to replace the battery, as determined by maintenance or testing.

14.8 14.7

<0.1

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 29 of 72 Docket Nos. 50-254 and 50-265 Table APLB RAI-04.1 FPRA JHEP Sensitivity Study Results TS LCO Condition LAR RICT Estimate FPRA JHEP Sensitivity RICT Estimate Change in RICT Estimate 3.8.4.D DC Sources-Operating - Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Condition B or C.

5.4 5.4

< 0.1 3.8.4.E DC Sources-Operating - Opposite unit 125 VDC electrical power subsystem inoperable.

5.5 5.5

< 0.1 3.8.7.A Distribution Systems-Operating - One or more AC electrical power distribution subsystems inoperable.

1.3 1.3

< 0.1 3.8.7.B Distribution Systems-Operating - One or more DC electrical power distribution subsystems inoperable.

5.4 5.3

< 0.1 3.8.7.C Distribution Systems-Operating - One or more required opposite unit AC or DC electrical power distribution subsystems inoperable.

5.5 5.5

< 0.1 b) i.

The core damage frequency (CDF) and large early release frequency (LERF) fire PRA (FPRA) dependency analyses were performed separately but combined into a single recovery file. There are separate Unit 1 and Unit 2 dependency analyses. For Unit 1, there are 2,780 joint human error probabilities (HEPs) of which 1,681 have values less than 1E-5. The range of these is 9.97E-6 to the floor value of 1E-6 (1,305 have the floor value). For Unit 2, there are 2,545 joint HEPs of which 1,342 have values less than 1E-5. The range of these is 9.97E-6 to the floor value of 1E-6 (946 have the floor value).

The Fire HRA Dependency Analysis assessment process is consistent with NUREG-1921 Section 6.2 which states, "For fire HRA, it is recommended that the application of a lower bound follow the same guidance as was applied to the internal events PRA." The internal events PRA applied a floor value of 1E-6. The FPRA applied the same lower bound floor value of 1E-6.

Additionally, Section 6.2 of NUREG-1921 acknowledges that the floor value of 1.0E-05 stated in NUREG-1792 is a suggestion and that use of the 1.0E-05 floor value can introduce skewing of risk metrics and importance measures as seen in the Significance Determination Process (SDP), which is a "delta" type calculation similar to a RICT.

Prior to the application of the JHEP floor value, dependency analysis was applied using the Electric Power Research Institute (EPRI) Human Reliability Analysis Calculator (HRAC) Dependence Decision Tree. The HRAC Dependence Decision Tree is given in NUREG-1921 Figure 6-1 and includes the following dependency factor considerations as listed in NUREG-1921, Section 6.2:

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 30 of 72 Docket Nos. 50-254 and 50-265 x

The time required to complete all actions in relation to the time available to perform the actions x

The availability of resources (e.g., crew members and other plant personnel to support the performance of actions outside the control room) x Factors that could lead to dependence (e.g., common instrumentation or procedures, an inappropriate understanding or mindset as reflected by the failure of a preceding human failure event (HFE), and increased stress) o Intervening Success o Same Crew o Common cognitive effort o Cue timing o Operations Resources o Common location o Sequential Timing o Stress The results for TS 3.8.7.A case 4 and TS 3.8.7.C case 4 were reviewed. JHEPs that contribute to the decrease in RICT estimate were reviewed. Most of the JHEPs that contribute to the decrease include combinations of 3, 4, 5, or 6 operator actions that include actions to start torus cooling, start an idle service water (SW) pump to support reactor pressure vessel (RPV) injection, or to depressurize the RPV. Therefore, the following two examples were chosen.

The first example, Unit 1 Combination 2, includes the following two actions:

1.

1RHOP-SPCE---H-F, OPERATOR INITIATES TORUS COOLING (NON-ATWS)

- EARLY - FPRA VERSION (4.30E-05)

2.

1ADOP-DEP-ADSH-F, OPERATOR MANUALLY DEPRESSURIZES THE RPV (NON-ATWS) - FPRA VERSION (7.26E-04)

The zero level of dependence is appropriate for this combination of actions because the actions serve two different functions: maintaining RPV level control and providing containment heat removal. They therefore do not have common cognitive cues. The cue to initiate suppression pool cooling (SPC) on high torus temperature occurs early in the accident sequence. The cue for depressurization to support low pressure injection is not encountered until after a heat capacity temperature limit (HCTL) violation. The failure to initiate SPC leads to the need to depressurize the RPV, therefore, the actions are sequential. Both actions are executed in the main control room. Neither action has a high execution stress assignment. These aspects lead to an HRAC dependency endstate of zero dependence.

The resulting JHEP is 3.12E-08, which is increased to the JHEP floor of 1.0E-06.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 31 of 72 Docket Nos. 50-254 and 50-265 A second example, Unit 1 Combination 58, contains the following two actions:

1. 1RHOP-SPCE---H-F, OPERATOR INITIATES TORUS COOLING (NON-ATWS)

EARLY - FPRA VERSION (4.30E-05)

2. BSWOP-SWPMP--H-F, OPERATOR STARTS IDLE SW PUMP - FPRA VERSION (1.08E-02)

As in the first example, the actions serve two different functions: maintaining RPV level control (start service water (SW) pump to support feedwater, safe shutdown make-up pump (SSMP), control rod drive (CRD), or condensate) and providing containment heat removal (which is supported by the separate RHR Service Water System). They therefore do not have common cognitive efforts. The cue for starting a SW pump to support RPV injection is encountered early if needed to support high pressure injection or may be later if needed to support low pressure injection (which is the case for loss of decay heat removal sequences). The cue to initiate SPC on high torus temperature occurs many hours earlier and because failure to initiate SPC leads to the need to depressurize the RPV, the actions are sequential. Both actions are executed in the main control room. Neither action has a high execution stress assignment. These aspects lead to an HRAC dependency endstate of zero dependence.

The resulting JHEP is 4.63E-07, which is increased to the JHEP floor of 1.0E-06.

These two examples are illustrative of the types of floor JHEPs that are driving the

>3 day reduction (~14%) in the RICT estimate for the two cases noted above. Most of the other contributing JHEPs involve similar combinations of these actions and sometimes involve a few other actions. Applying an arbitrary floor value of 1E-5 to all of the floor JHEPs is not warranted.

b) ii Not applicable. See a) and b)i responses above.

APLB RAI-05 NUREG-2178, Volume 1 "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume," (ML16110A14016) contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction.

Additionally, NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet.

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

b) Justify any modelling in which the base of an obstructed plume is located at less than one half of the cabinet's height.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 32 of 72 Docket Nos. 50-254 and 50-265 c) As an alternative to item b above, add an implementation item to remove credit for the obstructed plume model in the FPRA prior to the implementation of the RICT program.

Constellation Response to APLB RAI-05 a) In cases where the obstructed plume model was used the base of the fire was assumed to be 1' below the top of the panel. Therefore, no fires were assumed to be at an elevation less than one-half of the cabinet height.

b) Not applicable per the response to a) above.

c) Not applicable per the response to a) above.

APLB RAI-06 Guidance in FAQ 08-0042 from Supplement 1 of NUREG/CR-6850 applies to electrical cabinets below 440V. With respect to Bin 15 as discussed in Chapter 6, it clarifies the meaning of "robustly or well-sealed." Thus, for cabinets of 440V or less, fires from well-sealed cabinets do not propagate outside the cabinet. For cabinets of 440V and higher, the original guidance in Chapter 6 remains and states that Bin 15 panels which house circuit voltages of 440V or greater are counted because an arcing fault could compromise panel integrity (an arcing fault could burn through the panel sides, but this should not be confused with the high energy arcing fault type fires)." FPRA FAQ 14-0009, "Treatment of Well-Sealed MCC Electrical Panels Greater than 440V" (ADAMS Accession No. ML15119A176) provides the technique for evaluating fire damage from MCC cabinets having a voltage greater than 440V. The NRC guidance recommends that propagation of fire outside the ignition source panel should be evaluated for all MCC cabinets that house circuits of 440V or greater.

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

b) If well-sealed cabinets less than 440V are included in the Bin 15 count of ignition sources, provide justification for using this approach as this is contrary to the guidance.

Constellation Response to APLB RAI-06 a) Well-sealed motor control center (MCC) cabinets greater than 440 V were evaluated consistent with the guidance in frequently asked question (FAQ) 14-0009. Per the guidance, fire propagation outside the MCC from an arcing fault was modeled. The FAQ guidance for fire modeling was used to assign the fire severity factors.

b) Electrical cabinets less than 440 V are included in the Bin 15 count of ignition sources consistent with the guidance in FAQ 08-0042 from Supplement 1 of NUREG/CR-6850. Per the guidance, the electrical cabinet is included if the cabinet penetrations could not be dismissed to readily allow for the passage of air.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 33 of 72 Docket Nos. 50-254 and 50-265 APLB RAI-07 NUREG/CR-6850, Section 6, "Fire Ignition Frequencies," and FAQ 12-0064 "Hot Work/Transient Fire Frequency Influence Factors" (ADAMS Accession No. ML12346A488) describe the process for assigning influence factors for hot work and transient fires. Provide the following regarding application of this guidance:

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

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

c) Indicate whether you have any combustible control violations and discuss your treatment of these violations for the assignment of transient fire frequency influence factors. For those cases where you have violations and have assigned an influence factor of 1 (Low) or less, indicate the value of the influence factors you have assigned and provide your justification.

d) If you have assigned an influencing factor of "0" to Maintenance, Occupancy, or Storage, or Hot Work for any fire physical analysis units (PAUs) provide justification.

e) If a weighting factor of "50" was not used in any fire PAU, provide a sensitivity study that assigns weighting factors of "50" per the guidance in FAQ 12-0064.

Constellation Response to APLB RAI-07 a) NUREG/CR-6850 guidance was used to apply influence factors to calculate transient fire frequencies with the addition of using the hot work influence factor. Each physical analysis unit (PAU) was assigned rankings based on the NUREG/CR-6850 transient influence factor descriptions. Subsequently, the rankings were reviewed and adjusted to ensure MEDIUM represented the average of PAUs such that the intent of FAQ 12-0064 was included in the FPRA. The extremely low and very low factors were not used since they were not included in the NUREG/CR-6850 guidance.

b) Administrative controls are not used to reduced influence factors.

c) Not applicable per the responses to a) and b) above.

d) An influence factor of "0" was not used.

e) A weighting factor of "50" was applied to the following PAUs:

x 19.1 - Service Building - First Floor x

2.0 - Service Building - Main Control Room x

6.3 - Service Building - Aux. Electric Equipment Room x

OUTDOOR - Outdoor Areas

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 34 of 72 Docket Nos. 50-254 and 50-265 APLB RAI-08 Quad Cities, Units 1 and 2 are adjoined and thus have common areas. The risk contribution from fires originating in one unit must be addressed for impacts to the other unit given the physical proximity of the other unit, common areas, and existence of shared systems.

Therefore, address the following:

a) Explain how the risk contribution of fires originating in one unit is addressed for the other unit given impacts due to the physical proximity of equipment and cables in one unit to equipment and cables in the other unit. Include identification of locations where fire in one unit can affect components in the other unit and explain how the risk contributions of such scenarios are allocated in the LAR.

b) Explain how the contributions of fires in common areas are addressed, including the risk contribution of fires that can impact components in both units.

c) Explain the extent to which systems are shared by both units and whether shared systems are credited in the PRA models for both units. If shared systems are credited in the PRA models for each unit, then explain how the PRAs address the possibility that a shared system is demanded in both units in response to a single initiating event or fire initiator.

Constellation Response to APLB RAI-08 a) Fire scenarios are postulated and damage to equipment or routing points are included regardless of unit. A fire is postulated to damage Unit 1 and Unit 2 equipment or cables within the zone of influence for the ignition source, e.g., if a Unit 1 ignition source includes Unit 2 equipment or cables in the zone of influence then those Unit 2 equipment or cables are included. As a result, the ignition source in the example given will include risk contribution to both the Unit 1 and the Unit 2 FPRA. Note, a plant trip is assumed for the unit being analyzed. The equipment or cable failures may result in a dual unit initiator which will be propagated through each unit.

Given the number of shared systems there are numerous physical analysis units (PAUs) that contribute to the fire risk of both units. The results for PAUs for Unit 1 and Unit 2 are included in the FPRA quantification notebook.

Part b provides a list of common area PAUs that contribute to the top 10 PAUs for Unit 1 or Unit 2.

b) Common areas that are included in the top 10 PAUs for either unit include:

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 35 of 72 Docket Nos. 50-254 and 50-265 PAU PAU Description Unit 1 CDF

(% Cont.)

Unit 1 LERF

(% Cont.)

Unit 2 CDF

(% Cont.)

Unit 2 LERF

(% Cont.)

2.0 Service Building - Main Control Room 3%

1%

3%

1%

3.0 Service Building - Cable Spreading Room 8%

22%

9%

10%

6.3 Service Building - Aux.

Electrical Equipment Room 14%

15%

24%

36%

6.1.A Unit 1Turbine Building - DC Panel Room (Small) 3%

1%

< 1%

< 1%

6.1.B Unit 1 Turbine Building - DC Panel Room (Large) 3%

2%

< 1%

< 1%

6.2.A Unit 2Turbine Building - DC Panel Room (Small)

< 1%

< 1%

2%

1%

6.2.B Unit 2 Turbine Building - DC Panel Room (Large)

< 1%

< 1%

2%

1%

8.2.6.C Unit 1/2 Turbine Building Ground Floor - Center Area 12%

8%

10%

4%

8.2.7.C Unit 1/2 Turbine Building Mezzanine Level - Center Area 5%

5%

3%

3%

8.2.8.E Unit 1/2 Turbine Building Operating Floor 3%

6%

3%

6%

c) Refer to the response to APLA RAI-03 for shared systems and the credit in the PRA models.

The FPRA credits the same systems as the FPIE PRA model. For the FPRA, the equipment of shared systems are included in the FPRA data relationships (e.g., equipment, cables, routing points) for each unit such that a fire initiator in either unit will result in fire failures of shared systems, if applicable.

APLB RAI-09 Traditionally, the cabinets on front face of the main control board (MCB) have been referred to as the MCB for purposes of FPRA. Appendix L of NUREG/CR-6850, (ML052580075) provides a refined approach for developing and evaluating those fire scenarios. FPRA FAQ 14-0008, "Main Control Board Treatment" dated July 22, 2014 (ML14190B307) clarifies the definition of the MCB and provides guidance for when to include the cabinets on the back side of the MCB as part of the MCB for FPRA. It is important to distinguish between MCB and non-MCB cabinets because misinterpretation of the configuration of these cabinets can lead to incomplete or incorrect fire scenario development. This FAQ also provides several alternatives to NUREG/CR-6850 for using Appendix L to treat partitions in an MCB enclosure. Therefore, address the following:

a) Describe the main control room MCB configuration, and use the guidance in FAQ 14-0008, to determine whether cabinets on the rear side of the MCB are a part of the MCB. Provide justification using the FAQ guidance.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 36 of 72 Docket Nos. 50-254 and 50-265 b) If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure using the definition in FAQ 14-0008 confirm that guidance in FAQ 14-0008 was used to develop fire scenarios in the MCB and determine the frequency of those scenarios.

c) If the cabinets on the rear side of the MCB are part of a single integral MCB enclosure and the guidance in FAQ 14-0008 was not used to develop fire scenarios involving the MCB provide a description of how the fire scenarios for the backside cabinets are developed and an explanation of how the treatment aligns with NRC accepted guidance.

d) If in response to parts (c) above, the current treatment of the MCB and those cabinets on the rear side of the MCB cannot be justified using NRC accepted guidance, then justify that the treatment has no impact on the RICT calculations. Alternatively, propose a mechanism that ensures that the FPRA is updated to treat the MCB enclosure consistent with the guidance in FAQ 14-0008, prior to implementation of the RICT program.

Constellation Response to APLB RAI-09 a) The main control board (MCB) consists of a horseshoe shape with Unit 1 MCBs on one half and Unit 2 MCBs on the other half. The MCB is a walk through MCB and consists of front and back panels that form an enclosed volume. This MCB configuration is consistent with FAQ 14-0008 guidance for inclusion of the rear side of the MCB panel.

b) The guidance in FAQ 14-0008 was used to develop fire scenarios in the MCB and determine the frequency of the scenarios. Specifically, the rear side panels were included in the fire scenario frequency and severity factors and non-suppression probabilities were recalculated based on the guidance in NUREG/CR-6850 Appendix L.

c) Not applicable per the response to b) above.

d) Not applicable per the response to b) above.

APLB RAI-10 The TSTF-505 LAR states in part that the Internal FPRA model was developed consistent with NUREG/CR-6850 and only utilizes NRC approved methods. As part of the ongoing PRA maintenance and update process described in the LAR, the licensee will address Internal FPRA methods approved by the NRC since the development of the Internal FPRA. Furthermore, the TSTF-505 LAR specifies that a full-scope FPRA model peer review was performed in 2013.

There have been numerous changes to the FPRA methodology since the last full scope peer review of the FPRA. The integration of NRC-accepted FPRA methods and studies described below that are relevant to this submittal could potentially impact the TSTF-505 results and/or the CDF and LERF. NRC has issued updated guidance for aspects of FPRA that supplant earlier guidance issued by NRC.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 37 of 72 Docket Nos. 50-254 and 50-265 x

NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities,"

(DELORES-VEWFIRE)," (ML16343A058) regarding the updated approach to credit incipient fire detections systems.

x NUREG-2169, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database," (ML15016A069) regarding changes in fire ignition frequencies and non-suppression probabilities.

x NUREG/CR-7150, "Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE)," Volume 2, "Expert Elicitation Exercise for Nuclear Power Plant Fire-Induced Electrical Circuit Failure," (ML14141A129) regarding possible increases in spurious operation probabilities.

x NUREG-2230, "Methodology for Modeling Fire Growth and Suppression Response for Electrical Cabinets Fires in Nuclear Power Plants," (ML20157A148) regarding electrical cabinet fires.

x NUREG-2178, "Refining and Characterizing Heat Release Rates from Electrical Enclosures During Fire (RACHELLE-FIRE), (ML20168A655) regarding heat release rates (Volume 2).

Section 2.5.5 of RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides guidance that indicates additional analysis is necessary to ensure that contributions from the above influences would not change the conclusions of the TSTF-505 LAR.

a) Provide a detailed justification for why the integration of the above NRC accepted FPRA methods and studies would not significantly impact the RICT calculation. As part of this justification, identify potential FPRA methodologies used in the FPRA that are no longer consistent with NRC guidance. Provide technical justification for methods in Quad Cities, Units 1 and 2, FPRA not accepted by the staff and evaluate the significance of their use on the RICT estimates.

OR b) Alternatively, if the above guidance has been implemented in Quad Cities, Units 1 and 2, FPRA, provide the following:

i.

Indicate whether the changes to the FPRA are PRA maintenance or a PRA upgrade as defined in ASME/ANS RA-Sa-2009, Section 1-5.4, as qualified by RG 1.200, Revision 2, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," (ML090410014) along with justification for the determination.

ii.

Discuss the focused scope (or full scope) peer review(s) that was performed to evaluate the changes that were determined in Part b.i. above to constitute a PRA upgrade and provide the date for when the peer review(s) was performed and for

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 38 of 72 Docket Nos. 50-254 and 50-265 when the peer review report(s) that evaluated the incorporation of the method(s) was approved.

Constellation Response to APLB RAI-10 a) Not applicable. See b) below.

b) i. Implementation of new guidance in the FPRA is determined to be PRA maintenance or PRA upgrade based on ASME/ANS-RA-Sa-2009 definition of PRA maintenance and PRA upgrade. The Nonmandatory Appendix 1-A, PRA Maintenance, PRA Upgrade, and the Advisability of Peer Review, is used to justify the determination.

x NUREG-2180 - Not implemented - Not applicable - no VEWFDS.

x NUREG-2169 - Implemented - PRA maintenance. The implementation of NUREG-2169 fire ignition frequencies is a data change from NUREG/CR-6850. This is similar to Appendix 1-A Example 3 in which a new data set is used.

x NUREG/CR-7150 - Implemented - PRA maintenance. The implementation of NUREG/CR-7150 spurious operation probabilities is a data change from NUREG/CR-6850. This is similar to Appendix 1-A Example 3 in which a new data set is used.

x NUREG-2230 - Implemented - PRA upgrade. The implementation of NUREG-2230 is a methodology change. The FPRA at the time of the original peer review used an alternate approach to detailed fire modeling. An electrical cabinet factor was previously used. These factors were removed and detailed fire modeling using NUREG-2230 was used.

x NUREG-2178 - Implemented - PRA maintenance/PRA upgrade. The implementation of NUREG-2178 electrical cabinet HRRs is PRA maintenance given this is a data change from NUREG/CR-6850. This is similar to Appendix 1-A Example 3 in which a new data set is used.

The implementation of the NUREG-2178 obstructed plume model is a PRA upgrade given this is a new methodology.

b) ii. The focused-scope peer review (FSPR) was performed February 25 through March 2, 2021, and the report was approved April 27, 2021. The scope of the FSPR included FSS-C1 though C8, FSS-D1 through D4, FSS-D6, FSS-H1 through H10, and FSS-G1.

The FSPR resulted in five finding level F&Os. In May 2021, a follow-on F&O closure was conducted and the F&Os were closed. Refer to Enclosure 2 of the LAR.

APLB RAI-11 RG 1.200, Revision 2 provides guidance for addressing PRA acceptability. RG 1.200, Revision 2, describes a peer review process using the ASME/ANS PRA standard

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 39 of 72 Docket Nos. 50-254 and 50-265 ASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications,"

as one acceptable approach for determining the technical acceptability of the PRA. The primary results of peer review are the F&Os recorded by the peer review team and the subsequent resolution of these F&Os. A process to close finding-level F&Os is documented in Appendix X to the Nuclear Energy Institute (NEI) guidance documents NEI 05-04, NEI 07-12, and NEI 12-13, titled "NEI 05-04/07-12/12-06 Appendix X: Close-out of Facts and Observations (F&Os)" (ML17086A431), which was accepted by the NRC in a letter dated May 3, 2017 (ML17079A427).

Section 4 of Enclosure 2 to the TSTF-505 LAR states than one FPRA F&O, F&O 9-1, remains open and provides a succinct disposition. However, the TSTF-505 LAR does not provide the full description of the finding from the focused-scope peer review (FSPR), the recommendations from the FSPR team to address the finding, and a current disposition of this open F&O. Provide the FSPR full description, comments, recommendations, and licensee disposition related to this application for F&O 9-1.

Constellation Response to APLB RAI-11 The following is a copy of the Facts and Observations (F&O) 9-1 description and recommendations from the focused-scope peer review (FSPR) report:

QU-F3 (backward looking SR supporting FPRA SR FQ-F1) is NOT MET as applicable to FPRA: Tables 4-4 through 4-9 list significant cutsets and nonsignificant cutsets, Tables 4-12 and 4-13 list the significant accident sequences, Tables 4-16 and 4-17 list the significant PAUs, and Tables 4-18 and 4-19 list the significant fire scenarios. However, it does not appear that a detailed description of each of these significant contributors is included, such as detailed descriptions of the sequences and scenarios with an explanation as to why these are expected.

Tables 4-22 through 4-25 provide a list of the significant operator actions and equipment.

However, it does not appear that a detailed description of each of these significant contributors is included, such as a discussion of the important operator actions and/or equipment and why they make sense and/or what could be done to decrease their importance.

To meet Cat II/III, the documentation should include a detailed description of significant contributors.

(This F&O originated from SR FQ-F1)

Associated SR(s) Basis for Significance Description of significant cutsets, accident sequences, etc., are not provided, as required by SR FQ-F1 QU-F3, backward looking SR supporting FPRA SR FQ-F1.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 40 of 72 Docket Nos. 50-254 and 50-265 Possible Resolution Provide a detailed description of significant risk contributors, such as detailed descriptions of the sequences and scenarios with an explanation as to why these are expected. Provide a discussion of the important operator actions and/or equipment and why they make sense and/or what could be done.

F&O 9-1 disposition related to this application:

F&O 9-1 has been addressed in the FPRA for this application. As recommended by the FSPR, a detailed description of the significant risk contributors was added for the referenced tables. A detailed description of the top 10 cutsets, accident sequences, PAUs, fire scenarios, operator actions, and equipment includes an explanation why the contributors make sense and are expected.

APLC RAI-01 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 staff's 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 revised flood analysis refers to a portable Darley pump to be utilized to provide makeup flow during certain external flooding scenarios. QCOA-0010-16, "Flood Emergency Procedure,"

Rev. 29, states that to provide a suction source for the Darley pump, while waters are between 595 ft and 599 ft elevation, a hose should be routed over the 4 ft barrier installed in the Reactor Building (RB) 1/2 trackway personal access. Access to flood water suction provides Operations with an additional method of cooling makeup before flood waters exceed 599 ft.

However, this sequence does not appear to be discussed in the estimate of flood risk in the TSTF-505 LAR. The LAR provides an estimate of 0.3 conditional core damage probability (CCDP) for operators to install flood barriers.

a) Describe how the Darley pump mitigates risk during the external flood scenarios.

b) Provide the risk contribution (importance) of this sequence.

c) Describe the training operators receive regarding this scenario, including the frequency of training.

d) Describe how the timing for taking the action will be validated during training and how deviations between the actual and modeled required time will be addressed.

e) Describe the credit given for this scenario in the licensee's RICT application.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 41 of 72 Docket Nos. 50-254 and 50-265 f) Clarify if the licensee intends for this to be an implementation item for the applications.

Provide, if needed, a description of the implementation item.

Constellation Response to APLC RAI-01 a) The Darley pump is not credited in any scenario used to mitigate external flood risk in the TSTF-505 or 50.69 applications. The current station external flood protection strategy described in USFAR 3.4.1.1 includes shutting down both units, cooling down, and disassembling the vessel prior to the flood waters exceeding the plant grade elevation of 594.5 feet and will remain in place. The external flood screening analysis conservatively assumes no credit for the Darley pump and will instead rely on modified flood barriers to protect the station up to 599 feet.

b) As mentioned above, no credit is given for the Darley pump with respect to screening the external flood hazard.

c) Operator training on circulating water covers the Darley pump's purpose, physical location, function, and operation. This training is provided biennially for end of cycle (EOC) and quadrennially for licensed operator requalification training (LORT) in accordance with the QCNPS training program description (TPD) and prior difficulty-importance-frequency (DIF) surveillances. It also includes a performance objective to start and stop the pump in accordance with QCOA-0010-16 Attachment A that is initial only on-the-job training (OJT).

d) The Darley pump is not used to screen any scenarios in the TSTF-505 or 50.69 applications. Therefore, no time validation is required as it is not considered in any scenarios.

e) No credit is taken for the Darley pump this scenario. The scenario conservatively assumes a CCDP of 1.0 when water tops the flood barriers.

f) As described above, there is no need for a separate Darley Pump related implementation item.

APLC RAI-02 The TSTF-505 LAR Attachment 4, Section 5, states that completion of EC 636914 "Update to LIP Barriers to Assist the Station External Flood Response," which is being developed to modify barriers to protect the plant up to the 599 feet elevation, is tracked as a RICT Program Implementation Item. It appears that EC 636912, "Update to the Station External Flood Response to Support Risk Reduction," as well as changes to QCOA 0010-16 are not complete.

The LAR states that "the addition of the LIP barrier installation to QC 0010-16 with available time will greatly reduce the risk to the station from external river floods."

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 42 of 72 Docket Nos. 50-254 and 50-265 Clarify why the completion of these documents are not proposed as implementation items.

Alternatively, provide implementation items that ensure completion prior to implementing either application.

Constellation Response to APLC RAI-02 Completion of the modification to the LIP barriers and the associated documentation updates is already tracked as the last implementation item in LAR Attachment 5, Table A5-1 (complete EC 636914 modifications). Updates to procedural guidance in QCOA-0010-16 and QCOA-0010-22 are explicitly included in EC 636912 and is also separately tracked by AT 4514225-63. EC 636914 directly cross-referenced EC 636912, so only the one EC was originally listed as an implementation item. The physical plant change and the related documentation revisions described in the engineering change packages (ECs) are required to be completed prior to amendment implementation and become effective at the same time as the amendment.

For clarity, CEG will revise TSTF-505 LAR Attachment 5, Table A5-1 to restate the related implementation item as shown below. The new copy of Attachment 5 Table A5-1 in of this submittal supersedes the previous version of that table in its entirety.

Source Description Implementation Item,

Section 5.2.3.3, LIP Flood Barrier Upgrades and Deployment LIP barriers are modified to protect the plant up to 599.0' Complete EC 636914, Update to LIP Barriers to Assist the Station External Flood, and EC 636912, Update to Station External Flood Response in Support of Risk Reduction, modifications (scope includes both physical plant and documentation /

procedure changes).

APLC RAI-03 The TSTF-505 LAR states that installation of the Local Intense Precipitation (LIP) barriers in the to-be-revised QCOA-0010-16 requires 8 equipment operators working concurrently.

QCOA-0010-22 states that "Fastlogs" weigh 113lbs and installation of each Fastlog requires a minimum of 2 people. The new flooding analysis states that operators will need additional training to support the flood strategy response where barriers are installed to prevent excessive water intrusion for floods up to 599 ft.

a)

Clarify if licensee procedures account for the number of required equipment operators on site to perform this installation.

b)

Provide the frequency of training that the operators receive or will receive on the to-be-revised procedure regarding this installation.

c)

Describe how the timing for taking the action will be validated during training and how deviations between the actual and modeled required time will be addressed.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 43 of 72 Docket Nos. 50-254 and 50-265 d)

It appears that the Human Error Probability (HEP) and Conditional Core Damage Probability (CCDP) for this installation incorporates the failure of the barriers themselves.

i.

Describe the quality control measures for these barriers.

ii.

Discuss the barriers storage location.

iii.

Describe and justify the barriers failure rates.

Constellation Response to APLC RAI-03 a) The validations performed for FLEX utilized minimum staffing levels (8 people used for barrier installation validation activities per TSTF-505 LAR Enclosure 4). In addition, there will be increased warning time associated with the river flood (96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />) compared to the local intense precipitation (LIP) (6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) hazard that was previously validated for FLEX. This increased warning time allows additional time to secure additional resources, if required, even though the previous staffing validation confirmed installation acceptability at the minimum staffing level.

b) Continuing training in a classroom setting for LIP barrier installation is currently performed quadrennially for licensed operator requalification training (LORT) and biennially for end of cycle (EOC). Initial training is conducted in the classroom as well as in a performance setting for QCOA 0010-16 and QCOA-0010-22 during the individual student's on-the-job training/task performance evaluation (OJT/TPE) phase. The engineering change (EC) package for the physical modification of the barriers drives any necessary changes to the training materials to include possible periodicity changes based on operator and supervisor input.

c) Validation Plan 12 in EC-EVAL 404409 documents the completion of reasonable simulation of the installation of the LIP barriers and it was estimated that the longest install time for a single barrier takes approximately 45 minutes. As a result, the time estimate for installing the barriers in parallel with one another is 45 minutes (however, given the large amount of warning time, the installation of the barriers would likely be performed in series). Validation was conducted against the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> LIP warning time which is conservative compared to the river flooding warning and execution time of approximately 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. Per EC 636912, the only stop log barrier that will use more stop logs post-modification is Barrier #8 (final height is eight logs). As described in the EC, new barrier height is estimated to add 10 minutes to the previously validated installation activity for that barrier (25 minutes per QCOA-0010-22).

However, EC 636912 also reduces the two tallest barriers from nine logs (with a 45-minute installation time) to eight logs.

QCOA-0010-16 will be updated to include guidance on when to start the installation of the LIP Barriers with respect to the river flood and will provide adequate warning time to account for any deviations between actual and modeled time. A more detailed review of the actions has been conducted utilizing the NRC developed Integrated Human Event Analysis System for Event and Condition Assessment (IDHEAS-ECA) methodology. In the enclosure to this response, it is estimated that the total human error probability (HEP) for installing all seven

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 44 of 72 Docket Nos. 50-254 and 50-265 barriers is 1.1E-2.

This conclusion was arrived at using the IDHEAS methodology to identify "Understanding" as the main driver of failure probability to the human response of the river flood. Once the operators in the main control room understand they need to enter the procedure, the steps to carry out and follow are very clear. A performance influencing factor on the understanding portion of the action was assigned the degraded value of SF3 for an unfamiliar scenario that has adequate training but is performed infrequently. This assigned an HEP value of 1E-2 to the cognitive portion of the action.

Installation of each of the seven barriers is also a well understood, documented, trained, and executed action for site personnel. Therefore, the execution portion of the HFE definition is judged to have only one degraded Performance Influencing Factor (PIF) of PD1 for the fastlogs and ideal PIFs (i.e., PD0)for the swing gate and door panel. Fastlog gates are individually estimated at 1.5E-4 execution failure probability given the physical demand of each of the fastlogs, the swing gate (Barrier #10) and door panels (Barriers #4 and #11) are estimated at 1.0E-4 execution HEPs.

A sensitivity was performed to understand the impact of assigning higher PIFs to the execution of the barriers. In the sensitivity case, the scenario familiarity PIF remained at SF3 for understanding and the physical demand increased to PD2 for all barriers. This increased the gate execution HEPs to 2E-4, each. Summing the values of the understanding and execution portions of the HEPs in the sensitivity resulted in essentially the same HEP (rounded to 1.1E-2) for installing the barriers. Assigning these alternate more conservative PIFs would not change the conclusions from this assessment.

The analysis shown in Table APLC RAI-03 EA-3 (enclosure to this response) supports the overall conclusion that using a 0.3 CCDP for the scenario is demonstrably conservative, as there is ample time to enter the procedure, diagnose the scenario, install the barriers, and recover, if needed. The time margin is large enough that the deviation in the time estimated and the actual time to perform is negligible in its contribution to the overall failure rate.

d)

i.

The LIP barriers are augmented quality per EC 396297. PM 193497 ensures the deployable LIP barriers are periodically inspected and maintained.

ii.

The barriers are stored at the installation location (right next to the opening or physically attached to the wall for hinged barriers). The barriers are taken off the walls and placed in the opening or swung closed using a hinge.

iii.

The conservative HEP utilized to represent failure to install each barrier is considerably higher than the failure probability associated with the barrier failure fragility. This treatment was determined adequate based on the actual height required to ensure successful screening of the scenario. The barriers will be designed to withstand hydrostatic and hydrodynamic loading up to 599 feet. However, the scenario only requires the barriers perform their function to an elevation of approximately 596 feet (1 foot above finished floor). At elevation 596 feet, the frequency of exceedance of a flood is approximately 1E-6/yr, which would meet the quantitative screening criteria. If

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 45 of 72 Docket Nos. 50-254 and 50-265 the barrier fails, the scenario will still screen despite the conservative assumption that there is no mitigation capability when water breaches the LIP barriers. The frequency of exceedance at the finished floor elevation of 595 feet is approximately 2E-6/yr.

Given that the barriers are designed to withstand four times the amount of pressure required to screen the scenario, the structural failure of the barrier is a much lower probability than the failure to install such that using the conservative value of 0.3 for failure of the barriers (install or structural) is appropriate.

APLC RAI-04 The licensee presentation during audit discussions on external flooding listed seven human failure events (HFEs), six with a human error probabilities (HEP) value of 5E-02 and one with a 3E-01 value. The 5E-02 value was stated as a conservative value assigned to HEP-1A (Tsw = 45 minutes), HEP-1B (Tsw = 45 minutes), HEP-3 (Tsw = 5 minutes), HEP-8 (Tsw = 25 minutes), HEP-9 (Tsw = 40 minutes), and HEP-11 (Tsw = 45 minutes). Tsw represents the amount of time from the initiating event (T = 0) to complete the action. It is unclear to the NRC staff how HFEs with significantly different Tsw values and represent three types of installations (swing gates, panels, and fastlogs) would have the same HEP value. It is unclear what HFE XF-LIP-HEP represent that has a Tsw value of 45 minutes and a HEP of 3E-01.

Section 5.3.3 of NUREG 1792 provides good HRA practices for post-initiator HFEs. Good Practice #2 states no HFE screening value should be lower than 0.1 or lower than the worst-case anticipated value (which appears to be 3E-01). Good Practice #4 states to revisit the use of post-initiator screening values versus detailed assessments for special applications. It is unclear to the NRC staff if the above HFEs are developed or screened and if the assigned HEP value is appropriate and if screened that these HFEs should be adequately assessed. The NRC staff notes that Capability Category I of the 2009 ASME/ANS PRA Standard SR HR-G1 the use of screening values for HECC-I states to use conservative estimate for the HEPs of the HFEs. The presentation represented these values as conservative. It is unclear to the NRC staff if the HRA development of these HFEs are conservative or bounding.

a) Clarify what operator action HFE XF-LIP-HEP represents and why its probability is different than the other six HFEs.

b) Clarify if the seven HFEs listed associated with the barriers are screened or developed operator actions. For those HFEs that are screened, justify that they are not risk significant.

c) For the HFEs that have a screening value, that constitute different actions, and have different Tsw values, justify that the use of the same screening value is appropriate.

d) Provide justification that the HEP value of 5E-02 is consistent with the guidance of NUREG-1792.

e) Provide justification that these HEP values are conservative for this application consistent with the guidance in NEI-06-09-A.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 46 of 72 Docket Nos. 50-254 and 50-265 Good Practice #6 of NUREG-1792 for post-initiator state to account for dependencies among post-initiator HFEs. SR HR-G7 requires, for multiple actions in the same accident sequence, to assess the degree of dependence and to calculate a joint HEP. It is unclear to the NRC staff if a dependency analysis (DA) was performed for the external flood sequence. Based on the material available to the staff during its regulatory audit, it appears that the following actions would also be performed during this sequence:

x Remove decay heat x

Install RHR 6" fire hose crossties to fire water supply x

Fill both torus x

Remove shiel plugs, drywell heads, and reactor vessel head x

Set up portable pump (Darley) for decay heat removal x

Fill Radwaste tanks with fire system water x

Portable makeup demineralizers to the CST x

Fill reactor cavities and separator-dryer pools x

Remove gates between storage pools x

Rack out all main breakers for equipment below 608 feet (lose normal decay heat removal systems) x Open plant doors The NRC staff understands that the licensee's analysis did not credit the portable pump.

However it is unclear if the other actions listed above would be performed in addition to the barrier installations. and if all the relevant operator actions are included in the licensee's analysis. If the licensee's analysis relies on the arrival of offsite personnel, the staff notes that access to the plant (flooded roads and bridges) should be accounted for.

f) Clarify if a DA was included in the external combined effects flood. If not, provide justification for its exclusion.

g) Clarify what other operator actions would be required during the installation of the barriers. If yes, include in this discussion their inclusion in the DA.

h) Clarify if offsite personnel are required for this response. If yes, include in this discussion how plant access was considered.

i)

Based on the responses to Parts (d), (e), and (f) provide justification that sum of these issues does not significantly impact the application.

Constellation Response to APLC RAI-04 The external flooding analysis presented in Enclosure 4 of the 50.69 LAR was intended to highlight the conservative nature of the external flooding analysis that was performed as part of the external hazards progressive screening approach consistent with Part 6 of the ASME/ANS PRA Standard [Reference 8].

The relevant rationale from the Technical Requirements for Screening and Conservative Analysis in Part 6 of the standard is as follows.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 47 of 72 Docket Nos. 50-254 and 50-265 (c) If an external hazard cannot be screened out using these screening criteria, then a demonstrably conservative or bounding analysis, when used together with quantitative screening criteria, can also provide a defensible basis for screening out the event, without the need for detailed analysis.

The relevant quantitative screening criteria is defined in Supporting requirement EXT-C1, Criterion C:

The core damage frequency, calculated using a bounding or demonstrably conservative analysis, has a mean frequency <10-6 / yr.

a) The table from the presentation is reproduced here for completeness. The times provided in the table are taken from the FLEX Validation #12 plan for LIP barrier installation. These times are provided for informational purposes. Note that Barrier 3 should actually have been noted as Barrier #10 in both the presentation and the original LAR submittal. That error has been corrected in Table APLC RAI-04.1 provided below.

Table APLC RAI-04.1 External Flood HEPs for LIP Barrier Installation HEP Name Barrier HEP Value Time Required1 HEP-1A 1A 5E-2 45 HEP-1B 1B 5E-2 45 HEP-3 10 5E-2 5

HEP-8 8

5E-2 25 HEP-9 9

5E-2 40 HEP-11 11 5E-2 5

XP-LIP-HEP All 3E-1 45 1 Installation times shown in this table are from the EC-EVAL 404409 Validation Plan #12.

It was subsequently identified that an additional barrier requires manual closure (Barrier #4).

This barrier was included in the detailed assessment for the installation of the barriers provided in the Appendix to APLC RAIs and its addition does not change the overall conclusion that 0.3 is a bounding value that is appropriate for use in representing the combined failure probability XP-LIP-HEP.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 48 of 72 Docket Nos. 50-254 and 50-265 b) The following paragraphs cover the combined response to subitems b, c, d, and e.

The seven listed HEPs were not developed HEPs. The installation of the seven LIP barriers in response to a potential extreme river flood scenario would be performed in conditions that would provide ample time to perform the actions associated with each barrier (i.e.,

>> 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> available compared to the time needed to install the barriers).

In lieu of providing additional justification for the use those screening values, a more detailed analysis provided in the enclosure to APLA RAI-03c shows that a more reasonable HEP estimate for performing the complete set of LIP barrier installations for this scenario is 1.1E-2. This confirms that the use of the combined failure probability of 0.3 is demonstrably conservative.

f) It is anticipated that the LIP barrier installation would be performed and completed prior to the 11 steps associated with the use of the UFSAR flood mitigation strategy described in UFSAR Section 3.4.1.1. (See the response to APLC RAI-06 for additional details of this UFSAR section.) A specific dependency analysis was not performed but taking no credit for use of the UFSAR flood mitigation strategy bounds any impact that a detailed dependency analysis would reveal.

g) Refer to the response to part f.

h) No offsite personnel are required for the response. Given that the flood procedure is entered prior to any extreme river flooding, the LIP barrier installation would occur prior to any potential threat to plant access.

i)

Refer to the response to part b.

APLC RAI-05 Material available to the staff during its regulatory audit (specifically, document EC 636914) states that the RB crane will be loaded onto the EDG if a LOOP occurs during external flooding.

a) Describe the function of the RB crane in this scenario.

b) Is loading the RB crane onto the EDG proceduralized as part of the EDGs loading following LOOP? If not, how will the loading be ensured if such loading is required.

Constellation Response to APLC RAI-05 a) As described in USFAR 3.4.1.1 the current station external flood protection strategy removes the shield plugs, drywell heads, and reactor vessel head using the Reactor Building (RB) crane prior to the flood waters topping the flood barrier. Loading the RB crane on the EDG is a defense in depth measure to ensure that the RB crane continues to have power if a loss of offsite power (LOOP) occurs early in the river flood scenario, where water is expected to top the flood barriers. No credit is taken for this configuration or the RB crane in the scenarios included in the TSTF-505 or 50.69 LAR, it is conservatively assumed failed.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 49 of 72 Docket Nos. 50-254 and 50-265 b) Loading the RB crane onto the EDG is not proceduralized at QCNPS. During a LOOP, the 1/2 EDG will load to Bus 13-1 or 23-1 and the unit EDGs load to Bus 14-1 and 24-1, respectively. The RB crane is powered from a dual feed from MCC 18-3 or 28-3. Bus 18 and Bus 28, which provide power to MCCs 18-3/28-3, are energized when Buses 13-1 and 23-1 are energized after EDG loads. Therefore, this is an automatic action that does not need to be proceduralized.

APLC RAI-06 The NRC staff reviewed the licensee's integrated assessment (IA; ML18180A033) for the revised flood hazard as part of the agency's post-Fukushima actions. The staff's review is documented at ML19168A196.

a) Identify instances where the revised flooding analysis and actions provided for this application change the information provided to the staff as part of the licensee's IA.

b) If changes to the licensee's IA are identified, for each change, justify why the staff's conclusions on the IA continue to remain valid.

Constellation Response to APLC RAI-06 a) The current external flood response strategy for flood events with flood levels up to 603' elevation is described in UFSAR Section 3.4.1.1. The strategy documented in the UFSAR does not substantially change (e.g., instead of just opening the plant doors to flood the plant at the end, the LIP barriers are installed first protecting the plant for an additional 4 feet and water enters the plant only after it reaches the top of the installed barriers). The table below provides a side-by-side comparison of the current description of flood mitigation steps in UFSAR 3.4.1.1 to the planned revision (italicized text) of these steps as part of the modification package implementation. All the other actions required in the current licensing basis (Steps 1-11) will still be performed, as evaluated in the IA. The addition of the LIP barriers as revised Step 1 is a risk reduction measure that does not impact the completion of any of the other CLB actions or mitigation capabilities it reduces the probability that the event actually gets to the last "allow flood waters to breach flood barrier" step.

The above changes to the flood strategy were carefully considered to ensure that the conclusions in the Integrated Assessment (IA) would not be changed. Increasing the height of the LIP Barriers to protect a full 4' above plant grade will only marginally change the installation time for the "fastlogs" (~15 mins additional time required) and not impact the installation of the panels or swing gates.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 50 of 72 Docket Nos. 50-254 and 50-265 Current Steps in UFSAR 3.4.1.1 Planned Revision of UFSAR 3.4.1.1

1. Shut down both units (normal procedures)

2. Remove decay heat (normal procedures)

3. Install RHR system 6" Firehoses crosstie to Fire Water Supply System (FWSS)

4. Fill both tori with water through 6" Firehoses to RHR system

5. Remove shield plugs, drywell heads and reactor vessel head

6. Set up the portable pumping equipment

7. Fill radwaste tanks with fire system water

8. Place portable makeup demineralizers in service to fill condensate storage tanks

9. Fill reactor cavities and separator-dryer pools using core spray system and RHR system

10. Remove gates between storage pools

11. Rack out all main breakers for equipment below elevation 608 feet

12. Open plant doors

1. Install flood barriers up to 599' 2 Shut down both units (normal procedures)

3. Remove decay heat (normal procedures)

4. Install RHR system 6" Firehoses crosstie to Fire Water Supply System (FWSS)

5. Fill both tori with water through 6" Firehoses to RHR system

6. Remove shield plugs, drywell heads and reactor vessel head

7. Set up the portable pumping equipment

8. Fill radwaste tanks with fire system water

9. Place portable makeup demineralizers in service to fill condensate storage tanks

10. Fill reactor cavities and separator-dryer pools using core spray system and RHR system

11. Remove gates between storage pools

12. Rack out all main breakers for equipment below elevation 608 feet

13. Allow flood waters to breach flood barriers b) As mentioned above, the only change to the strategy includes installing the LIP Barriers prior to a river flood and changing the height of the stoplog barriers to all protect to a height of 4 feet. Given that the LIP Barrier installation was included in the IA and evaluated using a warning time of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for the LIP event, the reliability of these actions remains valid since the river flood scenario has approximately 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> of warning time.

APLC RAI-07 NEI 00-043 Figure 5-6 provides guidance to be used to determine SSC safety significance. The same document states, in part, that if it can be shown that the component either did not participate in any screened scenarios or, even if credit for the component was removed, the screened scenario would not become unscreened, then it is considered a candidate for the LSS category.

Section 3.2.4 of the 50.69 LAR states that "All external hazards, except for seismic, were screened for applicability to QCNPS [Quad Cities Nuclear Power Station] using a plant-specific evaluation in accordance with Generic Letter 88-20 (Reference [27]) and updated to use the criteria in ASME PRA Standard RA-Sa-2009."

Likewise, Attachment 4 of the 50.69 LAR lists all external hazards as screened except for seismic hazard with an alternate approach. The guidance in NEI 00-04, Figure 5-6 regarding SSCs that play a role in screening a hazard is not discussed in Section 3.2.4 nor in 3 NEI 00-04, Revision 0, "10 CFR 50.69 SSC Categorization Guideline", July 2005 (ADAMS Accession No. ML052910035).

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 51 of 72 Docket Nos. 50-254 and 50-265 of the 50.69 LAR. Therefore, it appears to the NRC staff based on this lack of information that at the time an SSC is categorized it will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard because that evaluation has already been made. NRC staff notes that plant changes, plant or industry operational experience, updates to hazard frequency information, and identified errors or limitations in the PRA models could potentially impact the conclusion that an SSC is not needed to screen an external hazard.

a)

Clarify whether an SSC will be evaluated during categorization of the SSC using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard.

b)

If an SSC will not be evaluated using the guidance in NEI 00-04, Figure 5-6 to confirm that the SSC is not credited in screening an external hazard at the time of categorization because that evaluation has already been made, then explain how plant changes, plant or industry operational experience, updated information in hazard frequencies, and identified errors or limitations that could change that decision are addressed.

Table in Attachment 4 of the 50.69 LAR under the external flooding evaluation, states "Table A4-1 includes a list of LIP barriers credited for screening this hazard." Table A4-1 lists 14 barriers credited for screening the external flood hazard (six of which require manual action).

The TSTF-505 LAR the states that "the addition of the LIP barrier installation to QC 0010-16 with available time will greatly reduce the risk to the station from external river floods."

c)

Describe how the SSCs listed in Table A4-1 are to be categorized using the guidance in NEI 00-04, Figure 5-6.

Constellation Response to APLC RAI-07 a) During categorization of structures, systems, and components (SSCs), consistent with the guidance in NEI 00-04, Figure 5-6 (provided at the end of this response for convenience) will be followed.

b) See the response a) above.

c) The "Other External Hazards" table from Attachment 4 of the 50.69 LAR discusses external flooding, and in particular, the "Precipitation, Intense" (i.e., local intense precipitation or LIP) hazard. The LIP hazard was evaluated in the flood hazard reevaluation report, which found that the 14 flood barriers listed in Table A4-1 at key locations around the power block are needed to keep LIP flood waters from entering the QCNPS buildings and impacting safety related equipment.

A copy of 50.59 LAR Table A4-1 is provided here as Table APLC RAI-07.1 for convenience.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 52 of 72 Docket Nos. 50-254 and 50-265 Table APLC RAI-07.1: Barriers Credited for Screening External Flood Hazard Barrier Number Barrier Description Requires Manual Closure 1A/1B U1 & U2 Turbine Building Roll Up Doors (2 Barriers)

Yes 2A/2B U1 & U2 Turbine Building/Reactor Building Interlock to HRSS Personnel Doors No 3

Reactor Building 1/2 Trackway Door No 4

1/2 EDG Interlock Door Yes 5

Steel Plate on North Side of 1/2 EDG Building No 6

Unit 2 Turbine Building Northwest Personnel Access Door No 7

Turbine Building to Radwaste Building Door No 8

LTD Building to Trackway 1 Door Yes 9

Personnel Decon Room Door Yes 10 Service Building to Trackway 1 Door Yes 11 Aux Electric Room Door Yes TB North Siding Turbine Building North Siding No Since the barriers above are credited for screening, per the NEI 00-04 Figure 5-6 guidance (shown below), if the component is credited in the screening evaluation, the component is (high) safety significant.

The final step to "identify safety significant attributes" means that the alternate treatment implementers are informed of the attributes that cause the component (in this case, barriers) to be safety significant (e.g., keeping flood waters from impacting safety related SSCs).

Identification of safety significant attributes involves identifying the performance aspects and failure modes of the SSC that contribute to it being safety-significant. These attributes are to be provided to the IDP.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 53 of 72 Docket Nos. 50-254 and 50-265 APLC RAI-08 NEI 00-04, Rev. 0, Section 5, "Component Safety Significance Assessment" states, "If the plant does not have an external hazards PRA, then it is likely to have an external hazards screening evaluation that was performed to support the requirements of the IPEEE." NEI 00-04, Rev. 0, also states in Section 3.3.2, "Other Risk Information (including other PRAs and screening methods)," that the characterization of the adequacy of risk information should include "a basis for why the other risk information adequately reflects the as-built, as-operated plant."

The TSTF-505 and 10 CFR 50.69 LARs, Table E4-16 and Attachment 4 respectively, state under the Industrial or Military Facility section that "none of the operations at Cordova Industrial Park pose any threat to QCNPS from explosion, explosive shock, resulting missiles, or toxic fumes release" yet there is no mention of the effect of the CF Industries Chemical complex on the plant. Please explain the impact of the CF Industries Chemical complex on the plant and either justify that the impact can be screened for this application or describe how the impact is included in the RICT program.

Constellation Response to APLC RAI-08 Per UFSAR Section 2.2.1, industrial areas located within a 5-mile radius of the plant include the CF Industries chemical complex, located 3.1 miles north of the site. Per UFSAR Section 2.2.2, CF Industries houses a chemical complex producing nitrogen fertilizers and agricultural chemicals. Triennial gas surveys are conducted to analyze potential effects on control room operators provided in the "Control Room Habitability Study for Quad Cities Units 1 and 2, Commonwealth Edison Company" (i.e., the Triennial Toxic Gas Survey).

CF Industries and the amount of ammonia that is stored and shipped to that location on a regular basis is the reason why the station has ammonia detectors installed on the intake of

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 54 of 72 Docket Nos. 50-254 and 50-265 both the Train A and Train B Control Room HVAC system. The ammonia detectors will isolate the Control Room HVAC system to protect the Control Room Operators when the ammonia concentration reaches a specific level.

Note: Toxic Gas Surveys conducted in the 1980's (post TMI actions) initially identified the ammonia threat when the Train B Control Room HVAC system was originally installed. Since that time, a toxic gas survey is done periodically as required by the Control Room Habitability Program to determine if a change to the ammonia detector setpoint is needed or other detectors need to be added to protect the Control Room Operators. UFSAR Section 6.4.4.2.3 (Protection Provisions) further discusses ammonia detectors installed at the station.

The QCNPS TSTF-505 LAR Table E4-16 and 10 CFR 50.69 LAR Attachment 4 use screening criteria C1 (event damage potential is less than events for which plant is designed) and C3 (event cannot occur close enough to the plant to affect it) to screen the "Industrial or Military Facility Accident" hazard. For CF Industries, the screening criteria that applies is C1 based on the ammonia detectors' design function to protect the Control Room Operators by isolating the Control Room HVAC system when the ammonia concentration reaches a specific level.

APLC RAI-09 The TSTF-505 and 10 CFR 50.69 LARs, Table E4-16 and Attachment 4 respectively, states that criterion "C4" (event is included in the definition of another event) and criterion "C5" (event develops slowly, allowing adequate time to eliminate or mitigate the threat) was used to screen the snow hazard. The LARs focus on potential flooding impacts, but not on the design basis roof live load or the maximum recorded snowfall for the site. It is unclear to the NR staff whether the risk of this hazard is adequately considered for this application.

Justify the screening of risk associated with snowfall from the application (e.g., by comparing historical maximum snowfall against the design basis, showing that the occurrence frequency of snowfall events that could challenge the plant is low).

Constellation Response to APLC RAI-09 A 25 psf snow load was used in both the Reactor Building and Turbine Building structural calculations per QDC-0020-S-0567, Revision 6 and QDC-0030-S-1611, Revision 2, in accordance with the Uniform Building Code (UBC) [UFSAR Section 3.8.4.1.2 and Reference 3]

The largest snowfall total ever recorded on a single day during the period 1901-2024 in Davenport, IA (~20 miles southwest of Quad Cities) was 16.4-inches which occurred on January 3, 1971. The most snow over a two-day period in Davenport was 17.1-inches during January 12-13, 1971. [Reference 4] Assuming a 20-inch snowfall, the equivalent snow load would be about 11 psf, well below the 25 psf snow load used in the design calculations.

Note 1: An inch of water weighs about 5.2 psf. [Reference 5]

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 55 of 72 Docket Nos. 50-254 and 50-265 Note 2: A conservative "rule-of-thumb" for snow ratios states that for every 10-inches of snow there is an equivalent of 1-inch of water (10:1 ratio) [Reference 6].

As a follow-up to the discussions on this question during the NRC audit, annual snowfall totals from the years 2000 to 2024 were accessed from the National Weather Service [Reference 7].

The largest snowfall cumulative annual total of 65.1 inches occurred during the year 2013-2014 at the Moline Quad Cities Airport.

Also, Constellation Procedure OP-AA-108-111-1001, "Severe Weather and Natural Disaster Guidelines", Revision 23, provides pre-storm guidelines at Step 4.6 for snow, blizzards, snow squalls, and ice accumulation or polar vortex/extreme cold. The pre-storm prioritization activities guidelines include consideration of roof snow loading. Attachment 1 of the procedure, "Hurricane/Blizzard/Flooding Guidelines" contains a number of specific actions such as verifying a snow removal plan is in place including equipment and consumables.

In addition, CEG procedure SY-AA-101-146, "Severe Weather Preparation and Response",

Revision 3, requires at Step 4.6 to perform the following:

x ENSURE that Security posts are maintained with adequate Snow / Ice Melt and other equipment applicable to the post.

x REVIEW any site-specific snow removal plans with Maintenance personnel, emphasizing areas required to be cleared for Security patrols, zone assessment and response.

x COORDINATE snow removal with Maintenance personnel, emphasizing areas required to be cleared for Security patrols, zone assessment, and response.

Based on the reported roof structural calculations cited above, reported snowfall totals, and implementation of the severe weather procedure, it is unlikely that enough snow would be allowed to accumulate on and damage roofs of critical buildings prior to removal. Therefore, the hazard screening Criterion of C5 (event develops slowly, allowing adequate time to eliminate or mitigate the threat) is appropriate, as well as Criterion C1 (event damage potential is less than events for which plant is designed).

EEEB RAI-01 According to the Quad Cities, Units 1 and 2, Updated Final Safety Analysis Report, Chapter 8, page 8.3-9, the plant has three emergency diesel generators (EDGs), one dedicated to each unit, and a common which is automatically connected to the unit in which loss of offsite power (LOOP) and loss-of-coolant accident (LOCA) occurs. Considering this arrangement, explain whether the plant (both units) can be safely shutdown in the following scenarios: LOOP for both units, a LOCA on a unit, and a single failure of EDG on another unit. If not, please explain the single-failure criteria as applied to the two units. This clarification will help understand/verify the Design Success Criteria considered in Table E-1 of the LAR.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 56 of 72 Docket Nos. 50-254 and 50-265 Constellation Response to EEEB RAI-01 The station is provided with sufficient and independent power sources to assure safe reactor shutdown under emergency conditions (worst case) on total loss of all offsite power (LOOP) concurrent with a design basis accident (DBA). The diesel generator (DG) system provides emergency source of ac power in the event all normal offsite power becomes unavailable. The system consists of three DGs: 1, 2, and 1/2, which is a shared DG. One DG can provide the necessary power for a safe unit shutdown during a LOOP with or without an accident occurring simultaneously. First, if a LOOP occurs and no accident signal is present from either unit, DG 1/2 is automatically connected to bus 13-1 (Unit 1), or 23-1 (Unit 2), depending on which unit loses power first. Second, if a LOOP occurs with a loss of coolant accident (LOCA) signal, DG 1/2 is automatically connected to the unit having the accident signal and is disconnected from the other unit. There are other manual capabilities within the system. For example, DG 1 can power the Unit 2 main 4160-V essential service system (ESS) bus, if desired, by manually closing two circuit breakers. Additionally, the DG system can backfeed to the main unit 4160-V auxiliary buses from the 4160-V ESS buses by manually closing two circuit breakers. However, such manual operations are possible only under certain specific conditions because interlock devices are installed to protect against possible fault conditions. For example, connecting DG 1 to Unit 2 4160-V ESS bus cannot be accomplished if DG 2 is already connected to that bus.

Such flexibility is an operational convenience designed and controlled such that safety is not jeopardized.

EEEB RAI-02 According to TSTF-505 LAR, Table E1-1, corresponding to TS Condition 3.8.1.B "One required DG inoperable," the design success criteria require "Two of out of three DGs", whereas the PRA success criteria require "One out of three DGs." Explain the reason for this difference.

Constellation Response to EEEB RAI-02 PRA success criteria are often different from design success criteria, as the requirements, goals, and objectives of a PRA are different from those of a design perspective. For example, design success criteria consider a concurrent single failure of an active component whereas PRA success criteria treats concurrent failures probabilistically. Specific to diesel generators, when AC-powered equipment is part of the success path in the QCNPS PRA, a single train of equipment is sufficient (e.g., to maintain core inventory and core cooling, for suppression pool cooling, etc.). Thus, only a single DG needs to be in operation to provide AC-power to a single train of equipment used in response to events postulated and addressed in the PRA.

Sufficiency of a single train is established by normal plant operation and operating history, plant procedures, and detailed thermal-hydraulics analyses.

EEEB RAI-03 Explain the single-failure criteria considered for the safety-related 250 VDC and 125 VDC systems as applied to the two units.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 57 of 72 Docket Nos. 50-254 and 50-265 Constellation Response to EEEB RAI-03 The DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure. The DC electrical power system is consistent with the recommendations of Safety Guide 6, March 10, 1971 and IEEE Std 308-1978.

125 VDC System The 125 V battery system of each unit is sized to start and carry the normal dc loads plus all dc loads required for safe shutdown on one unit and the operational loads required to limit the consequences of a design-basis event on the other unit, for a period of four hours following loss of offsite power plus a single active failure without taking credit for the battery charger. During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system. In case of loss of normal power to the battery charger, the DC loads are automatically powered from the associated battery. Each battery has adequate storage capacity to carry the required normal loads plus all loads required for safe shutdown on one unit and operations required to limit the consequences of a design basis event on the other unit for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This time period is deemed adequate to safeguard the plant until normal sources of power are restored.

The DC batteries associated with each unit are housed in a ventilated room apart from its charger and distribution buses. This arrangement ensures redundant subsystems are located in an area separated physically and electrically from the other subsystems to ensure that a single failure in one subsystem does not cause a failure in a redundant subsystem. There is no sharing between redundant Class 1E subsystems such as batteries, battery chargers, or distribution buses.

The Division 1 and 2 125 VDC electrical power sources provide control power to selected safety related equipment as well as circuit breaker control power for 4160 V, 480 V, control relays and annunciators. Each unit includes a 125 VDC source consisting of a 125 VDC battery and two 125 VDC full capacity chargers (normal and spare). Each 125 VDC unit source (125 VDC battery and associated chargers) supplies power to the associated unit Division 1 125 VDC electrical power distribution subsystem and the opposite unit Division 2 125 VDC electrical power distribution subsystem. The Division 1 and 2 125 VDC electrical power distribution subsystems provide power to redundant loads; therefore, both unit 125 VDC sources are needed to support the operation of both units. These sources are referred to as the Division 1 and 2 125 VDC electrical power sources since they supply the associated units Division 1 and 2 125 VDC electrical power distribution subsystems, respectively. In addition, the Division 2 125 VDC electrical power distribution subsystems provide control power to safety related loads common to both units such as the Standby Gas Treatment System. Therefore, the opposite unit Division 2 125 VDC electrical power distribution subsystem is needed to support the operations of the given unit. This source is referred to as the opposite unit's 125 VDC electrical power subsystem; however, it receives power from the given unit's battery and full capacity chargers.

The design also includes an alternate battery for each 125 VDC electrical power distribution subsystem. However, the design configuration of the alternate battery is susceptible to single failure and therefore, is not reliable as a normal 125 VDC source. The loads between the

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 58 of 72 Docket Nos. 50-254 and 50-265 redundant 125/250 VDC subsystem are not automatically transferable except for the Automatic Depressurization System and the 1/2 diesel generator, the logic circuits and valves of which are normally fed from the Division 1 125 VDC system.

The DC power distribution system is described in more detail in Bases for LCO 3.8.7, "Distribution SystemOperating," and LCO 3.8.8, "Distribution SystemShutdown."

The 125 VDC system utilizes one safety related battery for each unit. The buses fed from each battery are split such that The Unit 1 battery feeds all the Division 1 loads for Unit 1 and all the Division 2 loads on Unit 2. The Unit 2 battery has the same arrangement where it supplies all Division 1 loads on Unit 2 and all the Division 2 loads on Unit 1. These divisional buses supply 125 VDC power to the corresponding AC divisional loads (EDG and ECCS pumps) such that a single failure of one 125 VDC battery will only affect one division of AC loads leaving the other division intact. If the single failure is applied to the Unit 1 Battery, all the Division 1 loads would be lost, but the Division 2 load would still be operable to supply all the Division 2 loads which would allow for two RHR pumps and a Core Spray pump to mitigate the accident which is all that is required. Each battery has enough capacity to power all the accident loads on one unit and all the shutdown loads on the non-accident unit.

250 VDC System There are two 250-V systems, one per unit. The basic function of the 250-V battery is to supply electrical power to the dc distribution systems whenever the battery charger, which supplies the normal source of power, fails. The safety-related 250-V battery system of each unit is sized to start and carry the normal dc loads plus all dc loads required for safe shutdown on one unit, and the operational loads required to limit the consequences of a design-basis event on the other unit, for a period of four hours following loss of offsite power plus a single active failure without taking credit for the battery charger.

The 250 VDC system is set up in a similar fashion with one 250 VDC Battery for each unit. In this case, the unit battery supplies high pressure coolant injection (HPCI) on one unit and reactor core isolation cooling (RCIC) on the other unit. A single failure of the Unit 1 250 VDC battery would disable HPCI, but RCIC would still be available from the Unit 2 battery. The Automatic Depressurization System (ADS) is the actual back-up to HPCI with respect to protection of the core. HPCI will maintain reactor level until the vessel is depressurized. If HPCI is lost, the ADS system will depressurize the vessel allowing core spray to inject. The 250 VDC batteries are both Division 2, however, the ADS system is fed from Division 1 of the 125 VDC system, so a loss the 250 VDC battery (Division 2) is mitigated by the 125 VDC ADS system (Division 1) which is still available. The 250 VDC batteries have enough capacity to power both the accident loads on one unit and the required loads for shut down on the non-accident unit.

EEEB RAI-04 In the LAR, corresponding to TS Conditions 3.8.7.A, "One or more AC electrical power distribution subsystems inoperable" the proposed RICT program has a NOTE stating: "Only applicable when a loss of function has not occurred." Explain why a similar NOTE is not

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 59 of 72 Docket Nos. 50-254 and 50-265 proposed for the TS Condition 3.8.7.B, "One or more DC [direct current] electrical power distribution subsystems inoperable" and TS Condition 3.8.7.C, "One or more required opposite unit AC or DC electrical power distribution subsystems inoperable."

Constellation Response to EEEB RAI-04 Condition E is the loss of function condition for LCO 3.8.7. Thus, a separate note regarding loss of function is not needed for Conditions A, B, or C. Please see the response to STSB RAI-03.d for the corrected mark-up of LCO 3.8.7.

EEEB RAI-05 According to the LAR, Table E1-1, corresponding to the TS Condition 3.8.1.C "Two required offsite circuits inoperable," the design success criteria require "Two required offsite circuits."

Explain how the design success criteria will be met with "Two required offsite circuits inoperable."

Constellation Response to EEEB RAI-05 The design success criteria in Table E1-1 should say two (of three) DGs as shown in the attached mark-up. The second DG supports opposite unit's Class 1E AC Electrical Power Distribution System needed for LCO 3.6.4.3 Standby Gas Treatment (SGT) system, LCO 3.7.4 Control Room Emergency Ventilation (CREV) System (Unit 2 only), and LCO 3.7.5 Control Room Emergency Ventilation Air Conditioning (AC) System (Unit 2 only).

A revised copy of TSTF-505 LAR Table E1-1 is provided in Attachment 5. The new copy of Table E1-1 in Attachment 5 of this submittal supersedes the previous version of that table in its entirety.

EEEB RAI-06 According to the LAR, Table E1-1, the TS Condition 3.8.1.D states "One required offsite circuit inoperable OR one required DG inoperable." Please confirm that the TS Condition 3.8.1.D in Table E1-1 was meant to state "One required offsite circuit inoperable AND one required DG inoperable." Also, according to the Table E1-1, for the design success require "Two required sources." Explain the two required sources.

Constellation Response to EEEB RAI-06 TS Condition 3.8.1.D in Table E1-1 should state "AND" instead of "OR" in the TS Condition Description column, as it is supposed to match the current Technical Specification wording.

See the mark-up referenced by/attached to the EEEB-5 response for the change in context to the other Table E1-1 rows for Table E1-1.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 60 of 72 Docket Nos. 50-254 and 50-265 The two required sources are the other offsite circuit (either the opposite unit's reserve auxiliary transformer and associated unit crosstie breakers, or the respective unit's reserve auxiliary transformer) and the other division diesel generator.

A revised copy of TSTF-505 LAR Table E1-1 is provided in Attachment 5. The new copy of Table E1-1 in Attachment 5 of this submittal supersedes the previous version of that table in its entirety.

EICB RAI-01 In Section 3.1.2.3 "Evaluation of Instrumentation and Control Systems" of the TSTF-505 Revision 2 Model Safety Evaluation, the NRC clarifies that the basis of the staff's safety evaluation is to consider "a number of potential plant conditions allowed by the new TSs" and to consider "what redundant or diverse means were available to assist the licensee in responding to various plant conditions." The TSTF-505 Revision 2 position recommends that "at least one redundant or diverse means (e.g., other automatic features or manual action) to accomplish the safety functions (e.g., reactor trip, safety injection, or containment isolation) remain available during the use of the RICT." This approach is consistent with maintaining a sufficient level of defense-in-depth in accordance with RG 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant Specific Changes to the Licensing Basis,"

(ML100910006), and the guidance in Revision 1 of RG 1.177, "An Approach for Plant Specific, Risk Informed Decisionmaking: Technical Specifications," (ML100910008), which further describe the regulatory position with respect to defense-in-depth (including diversity). of the TSTF-505 LAR lists the functions of the Instrumentation and Control Systems and their design logics; however, this list does not provide NRC staff adequate information to verify that at least one redundant or diverse means will remain available to accomplish the intended I&C safety functions during the proposed risk informed completion time.

Describe other means that exist to initiate the safety function for each plant accident condition that each affected I&C function is currently designed to address. The evaluation of "diverse means," should identify the conditions that the functional unit responds to, and for each condition, other means (e.g., diversity, redundancy, or operator actions) that can be used.

Alternatively, provide additional information to demonstrate that defense-in-depth is maintained during the extended completion times for each function. This information is needed to demonstrate compliance with 10 CFR 50.36(c), and consistency with the implementing guidance in RG 1.174 and the TSTF-505, Revision 2.

Constellation Response to EICB RAI-01 The response to EICB RAI-01 will be separately submitted no later than April 5, 2024.

STSB RAI-01 In Attachment 2 of the TSTF-505 LAR, the proposed change to add RICTs for Quad Cities TSs Required Action 3.3.5.1.F.2 (place channel in trip) is, in part:

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 61 of 72 Docket Nos. 50-254 and 50-265 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from discovery of inoperable channel concurrent with HPCI [high-pressure coolant injection] or reactor core isolation cooling (RCIC) inoperable

OR In accordance with the Risk-Informed Completion Time Program

The NRC staff recognizes that the licensee's proposed change is consistent with the NUREG-1433 "Standard Technical Specifications General Electric BWR/4 Plants," Revision 4 (ML21272A358) TS markups in TSTF-505, Revision 2. However, it has been brought to staff's attention that some of the TSTF-505 markups contain errors, introducing potential for licensee actions to be less conservative than the original intent of the requirements. To modify completion times that include the phrase "from discovery," the RICT shall start at discovery instead of the time the TS action statement is entered, or the normal "time zero." The requirement is not clear when the RICT statement is separated from the "from discovery" statement.

To provide clarity, discuss how the proposed change ensures that the time of entry for the condition as at from discovery is clear or revise the placement of the proposed completion time for TS Required Action 3.3.5.1.F.2 between "96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />" and "from discovery." Also, provide a similar discussion or change for proposed revision to Required Action 3.3.5.1.G.2, which is formatted similarly.

Constellation Response to STSB RAI-01 Quad Cities has revised their proposed mark-ups of Required Actions 3.3.5.1.F.2 and 3.3.5.1.G.2 so that the option before the AND reads as follows (inserted text shown in italics):

96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> or in accordance with the Risk Informed Completion Time Program from discovery of inoperable channel concurrent with HPCS or reactor core isolation cooling (RCIC) inoperable The mark-up for the portion below the AND is unchanged from the original QCNPS submittal. A revised TS page mark-up is provided in Attachment 3 of this submittal. This mark-up aligns with the format previously approved for Nine Mile Point 2 for the same two Required Action statements.

STSB RAI-02 In TSTF-505, Revision 2, TS Example 1.3-8 appears to contain a typographical error. Assess whether the reference in the first paragraph to "Condition C" should be revised to "Condition B."

Constellation Response to STSB RAI-02 The corrected TS page addressing the typographical error is provided in Attachment 3.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 62 of 72 Docket Nos. 50-254 and 50-265 STSB RAI-03 Discuss why the following Required Actions have a proposed note that excludes loss of function (LOF) conditions for the RICT program when the associated LCO has a separate TS condition that addresses a LOF condition. (This question is seeking to understand, not necessarily to revise, the proposed TS.)

a. RA 3.3.2.2.A.1 (TS 3.3.2.2 Condition B addresses LOF)

b. RA 3.3.4.1.A.1 and A.2 (TS 3.3.4.1 Conditions B and C address LOF)

c. RA 3.3.6.1.A.1 (TS 3.3.6.1 Condition B addresses LOF)

d. RA 3.8.7.A.1 (TS 3.8.7 Condition E addresses LOF)

Constellation Response to STSB RAI-03 The following explanation of the loss of function note was previously provided in Attachment 1 Section 2.4 Item 5 of the QCNPS LAR:

For several of the Instrumentation Section 3.3 TS, a footnote is added to the proposed statement "or in accordance with the Risk Informed Completion Time Program" to ensure that a RICT is not applied when the actuation/trip function is lost. Under this circumstance, TSTF-505, Revision 2, specifies the addition of a Note that reads "Not applicable when [all] required [channels] are inoperable." Because the loss of function is dependent upon not only the number of inoperable channels, but also the combination of inoperable channels within the trip systems, CEG has chosen to replace the TSTF-505 Note with a footnote which reads "Not applicable when trip capability is not maintained,"

which accomplishes the intended purpose of the TSTF-505 Note.

LCO 3.3.2.2 is associated with the Feedwater System and Main Turbine High Water Level Trip Instrumentation. Based on the differences in Condition wording between the QDC and NUREG-1433 standard, the mark-up for TS 3.3.2.2.A should have been Insert B (no note), not Insert C (with note). A revised TS page mark-up is provided in Attachment 3.

LCO 3.3.4.1.A.1 is associated with the Anticipated Transient Without Scram Recirculation Pump Trip Instrumentation. It was confirmed that the TSTF-505 Traveler does not require additional justification for this application. While other BWRs have included the loss of trip capability function note with this LCO, on further review, CEG agrees the note is not necessary in this case due to the existence of a clear delineation between the Condition that applies when trip capability is maintained and the applicable Condition when the trip capability is not maintained.

The mark-up for TS 3.3.4.1.A is revised to use Insert B (without the note). A revised TS page mark-up is provided in Attachment 3.

LCO 3.3.6.1 is associated with the Primary Containment Isolation Instrumentation. It was confirmed that the TSTF-505 Traveler does not require additional justification for this application. While other BWRs have included the loss of trip capability function note with this LCO, on further review, CEG agrees the note is not necessary in this case due to the existence of a clear delineation between the Condition that applies when trip capability is maintained and

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 63 of 72 Docket Nos. 50-254 and 50-265 the applicable Condition when the trip capability is not maintained. The mark-up for TS 3.3.6.1.A is revised to use Insert B (without the note). A revised TS page mark-up is provided in Attachment 3.

LCO 3.8.7 is associated with the operating electrical power distribution subsystems. Based on the differences in LCO and Condition wording between the QDC and NUREG-1433 standard, the mark-up for TS 3.8.7.A should have been Insert B (no note), not Insert C (with note).

Conditions B and C are unchanged by this response. A revised TS page mark-up is provided in.

As part of a broader review of QCNPS loss of function conditions, it was determined that LCO 3.6.1.7.E is a loss of function condition. It was included in the original LAR submittal with the modifying loss of function note included, but upon further review, if there is one valve in each line that cannot open, there is always a loss of function anytime Condition E is entered and therefore RICT would never be applied. To avoid future confusion regarding the note, it was determined that the mark-up of LCO 3.6.1.7.E should be removed from the scope of the RICT Program (Attachment 3 coversheet indicates page 3.6.1.7-2 as removed). The mark-up of Table E1-1 in Attachment 5 is revised to reflect that loss of two valves in a single line is a loss of function condition. The corresponding row in the original TSTF-505 LAR Attachment 4 is revised to indicate 3.6.1.7.E is not included, as shown in Attachment 6. The above discussion supersedes the "additional justification" provided in Table E1-3 of the TSTF-505 LAR for 3.6.1.7.E.

STSB RAI-04 For TS Required Action 3.8.4.B.2, the current completion time is "[p]rior to exceeding 7 cumulative days per operating cycle of battery inoperability, on a per battery basis, as a result of maintenance or testing." The frontstop "7 cumulative days" is not fixed like all described in TSTF-505, Revision 2; it is variable depending on how many days have been used during the operating cycle for an inoperable battery. Applying a RICT to this required action is also inconsistent with the guidance in NEI 06-09-A that does not mention variable frontstop completion times.

Provide justification for the following questions or remove this proposed change:

a) How would the variable frontstop be evaluated to begin calculating a RICT?

i.

Provide an example where more than 1 day but less than 7 days has been applied to an inoperable battery in the operating cycle.

ii.

Provide an example of what would happen when TS 3.8.4 Condition B is applicable and the 7-day threshold has already been exceeded (e.g., RICT was used one time and exited during the same operating cycle).

b) Are there ever emergent maintenance and testing scenarios that would apply to TS 3.8.4 Condition B? If so, would a RICT be applied after the 7-day threshold were exceeded and what frontstop value would be selected?

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 64 of 72 Docket Nos. 50-254 and 50-265 Constellation Response to STSB RAI-04 CEG has decided to remove the application of RICT from TS 3.8.4.B.2. The mark-up for this item is removed from TSTF-505 LAR Attachment 2. A revised mark-up of TS page 3.8.4-2 is provided in Attachment 3. The related description in Attachment 4 is revised to reflect "Apply RICT?" being changed to No and the comment revised to reflect that RICT should not be applied to a variable frontstop completion time. Additionally, Enclosure 1 Table E1-1 and Table E1-2 are revised to remove the row for TS Condition 3.8.4.B. This update is shown in the revised copy of TSTF-505 LAR Table E1-1 is provided in Attachment 5.

STSB RAI-05 The proposed administrative controls for the RICT program in TS 5.5.15 paragraph "e" of of the TSTF-505 LAR was based on the TS markups of TSTF-505, Revision 2, for Quad Cities, Units 1 and 2. The NRC staff recognizes that the model safety evaluation (SE) for TSTF-505, Revision 2, contains improved phrasing for the administrative controls for the RICT program in TS 5.5.15 paragraph "e," namely the phrasing "approved for use with this program" instead of "used to support this license amendment." In lieu of the original phrasing in TS 5.5.15 paragraph "e", discuss whether the phrases "methods used to support Amendment # xxx" or, as discussed in the TSTF-505 model SE, "methods approved for use with this program" would provide more clarity for this paragraph.

Constellation Response to STSB RAI-05 QCNPS intends to adopt TSTF-591 [Reference 9], which replaces the current TS 5.5.15 paragraph "e", adds new paragraphs "f" that shifts the program from using Regulatory Guide 1.200 Revision 2 to Revision 3, and paragraph "g" that references a new TS 5.6.7. to this submittal provided the evaluation of the adoption of TSTF-591 into the QCNPS TS. Since TSTF-591 replaces paragraph "e" with new text, the specific change noted in this question would no longer be applicable. A revised TS page mark-up showing the entire proposed wording for the added TS 5.5.15 and TS 5.6.7 sections is included in Attachment 3. References

[1] U.S. NRC, "Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants,"

NUREG-1150, dated December 1990 (https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr1150/index.html)

[2] Letter from J.B. Hosmer (Commonwealth Edison Company) to U.S. NRC,

Subject:

"Quad Cities Station Units 1 and 2, Response to NRC Review of Individual Plant Examination Submittal - Internal Events," dated August 30, 1996 (ADAMS Accession No. 9609100313 or ML11284A194)

[3] TDBD-DQ-1, Revision 1, "Topical Design Basis Document Structural Design Criteria for Quad Cities Units 1 and 2 and Dresden Units 2 and 3"

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 65 of 72 Docket Nos. 50-254 and 50-265

[4] https://www.extremeweatherwatch.com/cities/davenport/most-daily-snow

[5] https://roofonline.com/weight-of-water

[6] https://www.weather.gov/arx/why_snowratios

[7] https://www.weather.gov/wrh/climate?wfo=dvn

[8] ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Addendum A to RAS-2008, ASME, New York, NY, American Nuclear Society, La Grange Park, Illinois, February 2009.

[9] Letter from A. Russell (U.S. NRC) to Technical Specifications Task Force,

Subject:

Final Model Safety Evaluation of Technical Specifications Task Force Traveler TSTF-591, "Revise Risk-Informed Completion Time (RICT) Program" (EPID L-2022-PMP-0003), dated December 18, 2023 (ADAMS Accession Nos. ML23325A213 and ML23325A214)

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 66 of 72 Docket Nos. 50-254 and 50-265 Enclosure to Constellation Response to APLC RAI-03 IDHEAS-ECA HRA Analysis - Operators Fail to Install Flood Barriers Prior to External Flood This enclosure provides additional supplemental information supporting the Constellation Responses to APLC RAI-03 and APLC RAI-04 HFE Definition Quad Cities Nuclear Power Station (QCNPS) has 14 credited flood barriers (including the north Turbine Building wall siding) at key openings around the power block to keep local intense precipitation (LIP) flood waters from entering the buildings during the event as shown in Figure APLC RAI-03 EA-1 [Reference E-1]. Seven of the barriers are temporary and passive requiring manual installation. The remaining barriers are permanently installed exterior doors or plates that do not require manual actions to perform their function. Figure APLC RAI-03 EA-1 shows all the barrier locations. Table APLC RAI-03 EA-1 summarizes the barriers that require manual closure along with their barrier numbers, description of the barrier locations, barriers that require manual installation, as well as their final barrier height, original barrier height, barrier type and maximum time to install the barrier.

The LIP barriers were designed to be installed quickly prior to a LIP event with approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of warning time. A proposed modification to QCOA-0010-16 will direct operators to install the barriers prior to the arrival of flood waters from a river flood to protect the site up to an elevation of 599 [Reference E-5]. There are 4 openings that require fastlog barriers, one swing gate, and two door panels. The fastlogs were originally designed for the LIP event, but EC 636914 will modify all fastlog-type barriers to use eight fastlogs and protect up to 4 feet above site grade.

This analysis applies the NRC IDHEAS-ECA method to quantify the human error probability (HEP) associated with operators failing to manually close barriers 1A, 1B, 4, 8, 9, 10 and 11 in accordance with the revised QCOA-0010-16 and QCOA-0010-22 [Reference E-8] prior to the arrival of a river flood. The critical steps to install the barriers are the same for both LIP and river flood, however, the river flood scenario has significantly more installation time available.

The flood barrier installation feasibility assessment is documented in Section 7.1.3 of the QCNPS IA [Reference E-2]. The following elements are included in the QCNPS IA justification for an adequate site response:

1. Defining critical path and time sensitive actions (TSAs)

2. Demonstrating all TSAs are feasible

3. Establishing unambiguous procedural triggers

4. Proceduralized and clear organizational response to flooding

5. Detailed flood response timeline In summary, the TSAs considered are the installation of the seven barriers that require manual actions. The longest barrier installation takes approximately 45 minutes and all seven could be done concurrently. Validation Plan #12 [Reference E-3] confirms the barriers can be installed with 8 equipment operators in approximately one hour and strain on these operators is minimal.

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 67 of 72 Docket Nos. 50-254 and 50-265 However, this analysis will assume that the barriers are installed in series to allow for additional time margin and remain demonstrably conservative. Operator training is provided for installation of the flood barriers in LN-FLEX.2, Section J [Reference E-4]. Given that the timeline is well defined, the entry conditions are laid out in the procedures, and training is provided, QCNPS has demonstrated that the site response to LIP is adequate.

Table APLC RAI-03 EA-1: LIP Barriers that Require Manual Closure Details Number1 Description1 Final Height3

[ft]

Original Height1

[ft]

Requires Manual Closure1 Barrier Type3 Max Time to Install2 1A U1 Turbine Building Roll Up Door 4.0 4.5 Yes Fastlogs (8) 45 mins 1B U2 Turbine Building Roll Up Door 4.0 4.5 Yes Fastlogs (8) 45 mins 4

1/2 EDG Interlock Door 4.0 4.0 Yes Door Panel 5 mins 8

LTD Building to Trackway 1 Door 4.0 2.5 Yes Fastlogs (8) 45 mins 9

Personnel Decon Room Door 4.0 4.0 Yes Fastlogs (8) 45 mins 10 Service Building to Trackway 1 Door 4.0 4.0 Yes Swing Gate 5 mins 11 Aux Electric Room Door 4.0 4.0 Yes Door Panel 5 mins 1QDC-0000-S-2089 [Reference E-1]

2This time is based on the conclusion that 8 log barriers can be installed in approximately 35 minutes

[Reference E-5] and 10 minutes was added for additional margin to provide a conservative estimate in the development of the human error probability.

3EC 636914 - Update to LIP Barriers [Reference E-5]

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 68 of 72 Docket Nos. 50-254 and 50-265 Figure APLC RAI-03 EA-1: LIP Barrier Locations at Quad Cities [Reference E-1]

[Figure E4-2, Enclosure 4, TSTF-505 LAR]

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 69 of 72 Docket Nos. 50-254 and 50-265 Quantification Breakdown of HFE using IDHEAS-ECA methodology The IDHEAS-ECA methodology [Reference E-6] requires the operator action to broken down into critical tasks. In IDHEAS-ECA a critical task can be cognitive, execution or both cognition and execution. This is a different breakdown from the EPRI HRA approach of defining critical tasks only for execution.

For each critical task the IDHEAS-ECA methodology considers five crew failure modes:

detection, understanding, decision making, action execution, and interteam coordination.

• Detection (D) is noticing cues or gathering information in the work environment.

• Understanding (U) is the integration of pieces of information with a persons mental model to make sense of the scenario or situation.

• Decision making (DM) includes selecting strategies, planning, adapting plans, evaluating options, and making judgments on qualitative information or quantitative parameters.

• Action execution (E) is the implementation of the decision or plan to change some physical component or system.

• Interteam coordination (Teamwork) (T) focuses on how various teams interact and collaborate on an action. The first four crew failure modes may be performed by an individual or a team, and interteam coordination is performed by multiple groups or teams.

Additionally, time is a failure mode which is applied to overall HFE as opposed to each critical task.

Table APLC RAI-03 EA-2 shows the critical tasks for the HFE "Operators fail to install flood barriers" and the appliable crew failure modes.

Table APLC RAI-03 EA-2: Identification of Critical Tasks and Failure Modes Critical Task Crew Failure Modes Main control room crew Diagnoses water level has reached 586 feet and dispatch local operators to begin barrier installations This critical task consists of cognition only and the crew failure mode is Understanding. This task requires that the operators understand that a flood warning is issued and that water levels will exceed 594 feet within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> or when actual river level is 586 feet.

Decision Making is not modeled as a crew failure mode since once the crew has diagnosed the water level, the procedure tells the crew exactly what to do. Operators are required to follow their procedures and in this scenario there is no reason for the crew to not trust their procedures and not implement them per their expected training requirements.

Install Barrier 1A Execution - Installation of each barrier is an execution only task.

Install Barrier 1B Install Barrier 4 Install Barrier 8

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 70 of 72 Docket Nos. 50-254 and 50-265 Table APLC RAI-03 EA-2: Identification of Critical Tasks and Failure Modes Critical Task Crew Failure Modes Install Barrier 9 Each barrier is modeled separately because each installation is performed by different independent teams of 2 crew members.

Install Barrier 10 Install Barrier 11 The crew failure modes timing and teamwork are not applicable to this analysis. Teamwork is not applicable to this HFE because the control room operators make the decision to send crews to install barriers but there is no critical interaction required between installations crews and the control room operators following the decision to start the barrier installation.

The timeline for this action starts when a flood warning is issued that water levels will exceed 594 feet within 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> or when actual river level is 586 feet. [Reference E-7, Step D.13.b].

The time available is 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />. The time required to install all 7 barriers is approximately 45 minutes if done concurrently. This time is based on Validation Plan #12 [Reference E-3].

However, it should be noted that this analysis assumes the barriers are installed in series and the maximum estimated time to install will be 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and 15 minutes. Given the time available is 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, this has a negligible impact on the evaluation of the HFE.

Table APLC RAI-03 EA-3: Evaluation of Each Critical Task Critical Task Performance Influencing Factors (PIF) associated with each critical task HEP Diagnose water level has reached 586 feet and dispatch local operators to begin barrier installations Scenario Familiarity - SF3 (Understanding only) - local intense precipitation events and river floods are not a common occurrence. However, these types of scenarios are trained on regularly (biennial for End of Cycle see APLC RAI -03b) This PFI is selected with a level of 1.

All other PIFs are considered to have NO impact on the decision to install the barriers.

1E-2 Install Barrier 1A Installation of each barrier is an execution only task.

Barriers 1A, 1B, 8, 9 consist of eight fastlog panels; each of which are heavy (113 lbs), there is potential for the gaskets to be affected by debris and the operators are instructed to use soap solution (if time permits).

Therefore, physical demand PD1 is selected for these 4 barriers.

Barriers 10 is a swing gate. Barriers 4 and 11 are door panels. These barriers are not physically demanding to operate and therefore the physical demand PIF is considered to have no impact (PD0).

1.5E-4 Install Barrier 1B 1.5E-4 Install Barrier 4 1.0E-4 Install Barrier 8 1.5E-4 Install Barrier 9 1.5E-4 Install Barrier 10 1.0E-4 Install Barrier 11 1.0E-4

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 71 of 72 Docket Nos. 50-254 and 50-265 Table APLC RAI-03 EA-3: Evaluation of Each Critical Task Critical Task Performance Influencing Factors (PIF) associated with each critical task HEP All other PIFs are ideal and have no impact on the performance to install barriers. This is based on the following justification:

x The installation of the barriers is covered in training at a frequency of every two years. There are no issues with scenario familiarity.

x All barrier installations are performed indoors so the external weather conditions have no impact on the operators performance. The indoor conditions have nominal lighting conditions, nominal temperatures and radiation levels.

x The task complexity to establish the barriers is straightforward with few procedure steps.

x The procedures steps to establish the barriers are straightforward and easy to use.

x There is no issue with sufficient staffing to perform all 7 barrier installations in the time available. This was shown in the validations.

Quantification Results The total HEP for this action is 1.1E-2. This value does not include any type of recovery. Given that the available time is 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />, recovery credit for peer checking or self-review could be credited. Currently, IDHEAs-ECA does not provide any guidance on how to quantify this type of recovery. The HEP of 1.1E-2 is therefore considered conservative.

Enclosure to Response to RAI-03 References E-1.

QDC-0000-S-2089, Evaluation of Flood Boundary Structures for Local Intense Precipitation, Revision 2, dated January 29, 2018 (available during the audit)

E-2.

QCNPS letter to NRC, "Exelon Generation Company, LLC Response to March 12, 2012, Request for Information Enclosure 2, Recommendation 2.1, Flooding, Required Response 3, Flooding Integrated Assessment Submittal," dated June 29, 2018

[Section 7.1.3] (ADAMS Accession No. ML18180A033)

E-3.

EC-EVAL 404409, Integrated Review of Flex Actions IAW NEI Validation Process Fukushima, dated April 20, 2016 [Validation Plan #12] (available during the audit)

E-4.

LN-FLEX.2, Licensed/Non-Licensed Operator Initial/Continual Training, FLEX Equipment, dated August 2019 (available during the audit)

QCNPS Request to Adopt TSTF-505 Response to Request for Additional Information Page 72 of 72 Docket Nos. 50-254 and 50-265 E-5.

EC 636914, Update to LIP Barriers to Assist the Station External Flood, dated June 3, 2022 (available during the audit)

E-6.

NUREG-2256, Integrated Human Event Analysis System for Event and Condition Assessment (IDHEAS-ECA), U.S Nuclear Regulatory Commission, dated October 2022 (ADAMS Accession No. ML22300A117)

E-7.

QCOA-0010-16, Revision 29, Flood Emergency Procedure (available during the audit)

E-8.

QCOA-0010-22, Revision 17, Local Intense Precipitation Response Procedure (available during the audit

ATTACHMENT 2 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Response to Request for Additional Information Related 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" Evaluation of the TSTF-591 Proposed Change

QCNPS Request to Adopt TSTF-505 Evaluation of the TSTF-591 Proposed Change Page 1 of 9 Docket Nos. 50-254 and 50-265

1.0 DESCRIPTION

OF PROPOSED CHANGES Since the submittal of the original license amendment request in June 2023, a subsequent Technical Task Force Traveler, TSTF-591, "Revised Risk-Informed Completion Time (RICT)

Program," (ADAMS Accession No. ML22081A224) has been reviewed and approved as a change to the Standard Technical Specifications (STS). Quad Cities Nuclear Power Station (QCNPS) has determined that is it desirable to incorporate the additional changes to the RICT Program described in TSTF-591 as part of their initial implementation of RICT as opposed to subsequently requesting adoption under the Consolidated Line Item Improvement Process (CLIIP). TSTF-591 revises the TSTF-505 proposed TS Section 5.5 "Risk Informed Completion Time Program," to reference Regulatory Guide (RG) 1.200, Revision 3, instead of Revision 2, and to make other changes to the program description wording. Additionally, TSTF-591 adds a new requirement to TS Section 5.6, "Reporting Requirements," to inform the NRC of future newly developed methods used to calculate a RICT.

2.0 ASSESSMENT

2.1 Applicability of Published Safety Evaluation CEG has reviewed TSTF-591, Revision 0, (ADAMS Accession no. ML22081A224) and the associated NRC model safety evaluation dated December 18, 2023 (ADAMS Accession No. ML23262B230). As described in the subsequent paragraphs, CEG has concluded that the technical basis is applicable to QCNPS Units 1 and 2, and supports incorporation of the additional TSTF-591 changes into the QCNPS plant-specific TS as part of the activities to create a RICT Program at QCNPS (i.e., the previously submitted request to adopt TSTF-505).

The addition of TSTF-591 TS changes to the in-progress review of the proposed changes for adoption of TSTF-505 contain no changes to the internal events / internal flooding and fire PRA models that that would be considered a Newly Developed Method, in accordance with the Pressurized Water Reactor Owners Group (PWROG) topical report, PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review," issued July 2020, as endorsed in Regulatory Guide 1.200 "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities", Revision 3.

2.2 Verifications and Regulatory Commitments The QCNPS TS utilizes different numbering than the STS on which TSTF-591 was based. This difference is administrative in nature and does not affect the applicability of TSTF-591 to QCNPS.

Additionally, QCNPS was not licensed to the 10 CFR 50, Appendix A, General Design Criteria (GDC). QCNPS UFSAR Section 3.1 "Conformance with NRC Design Criteria," provides an assessment against the 70 draft GDC published in 1967 and concludes that the plant-specific requirements are sufficiently similar to the Appendix A GDC. Therefore, this difference does not alter the conclusion TSTF-591, Revision 0 is applicable to QCNPS.

QCNPS Request to Adopt TSTF-505 Evaluation of the TSTF-591 Proposed Change Page 2 of 9 Docket Nos. 50-254 and 50-265 2.3 TS Changes to Adopt TSTF-591 The current request to adopt TSTF-505 adds a new section, TS 5.5.15, "Risk Informed Completion Time Program," to describe the RICT program. TSTF-591 revises the program description. Existing proposed paragraph e is replaced with the revised paragraph e shown below. Paragraphs f and g, also shown below, are added by TSTF-591.

e. A RICT calculation must include the following hazard groups: internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards using penalty factors, and extreme wind and tornado hazards using configuration-specific penalty factors for tornado missiles. Changes to these means of assessing the hazard groups require prior NRC approval.

f.

The PRA models used to calculate RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities."

g. A report shall be submitted in accordance with Specification 5.6.7 before a newly developed method is used to calculate a RICT.

In addition to the revision of TS 5.5.15, TSTF-591 adds a new specification, TS 5.6.7, as show below.

5.6.7 Risk Informed Completion Time (RICT) Program Upgrade Report A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT. The report shall include:

a.

The PRA models upgraded to include newly developed methods;

b.

A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c.

Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and

d.

All changes to key assumptions related to newly developed methods or their implementation.

Revised clean TS pages associated with the adoption of the additional TSTF-591 changes are provided in Attachment 3.

QCNPS Request to Adopt TSTF-505 Evaluation of the TSTF-591 Proposed Change Page 3 of 9 Docket Nos. 50-254 and 50-265

3.0 REGULATORY ANALYSIS

3.1 Applicable Regulatory Requirements and Guidance The regulation under Title 10 of the Code of Federal Regulations (10 CFR) 50.36(b) requires that:

Each license authorizing operation of a utilization facility will include technical specifications. The technical specifications will be derived from the analyses and evaluation included in the safety analysis report, and amendments thereto, submitted pursuant to [10 CFR] 50.34 ["Contents of applications; technical information"]. The Commission may include such additional technical specifications as the Commission finds appropriate.

The categories of items required to be in the TSs are listed in 10 CFR 50.36(c).

The regulation at 10 CFR 50.36(c)(5), states that administrative controls are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

NRC Regulatory Guides (RGs) provide one way to ensure that the regulations continue to be met. The RGs applicable to the adoption of TSTF-591 include:

x NUREG-0800, Revision 3, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [light-water reactor] Edition" (SRP):

o Chapter 19, Section 19.2, "Review of Risk Information Used to Support Permanent Plant-Specific Changes to the Licensing Basis: General Guidance,"

dated June 2007 (ADAMS Accession No. ML071700658).

o Chapter 16, Section 16.0, "Technical Specifications," March 2010 (ML100351425). Specifically, whether the proposed changes are consistent with NUREG-1433, "Standard Technical Specifications, General Electric, BWR/4 Plants" Volume 1, "specifications" and Volume 2, "Bases," Revision 5, dated September 2021 (ADAMS Accession Nos. ML21272A357 and ML21272A358, respectively) o Chapter 16, Section 16.1, "Risk-Informed Decision Making: Technical Specifications," March 2007 (ADAMS Accession No. ML070380228).

x NEI 06-09-A, Revision 0, "Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines" (ADAMS Accession No. ML063390639), provides guidance for risk-informed TS. The NRC issued a final SE approving NEI 06-09 on May 17, 2007 (ADAMS Accession No. ML071200238).

x NEI 17-07, Revision 2, "Performance of PRA Peer Reviews Using the ASME/ANS PRA Standard," provides guidance material for conducting and documenting a probabilistic risk assessment (PRA) peer review using the American Society of Mechanical Engineers

QCNPS Request to Adopt TSTF-505 Evaluation of the TSTF-591 Proposed Change Page 4 of 9 Docket Nos. 50-254 and 50-265 (ASME)/American Nuclear Society (ANS) PRA Standard, issued August 2019 (ADAMS Accession No. ML19231A182).

x PWR Owners' Group (PWROG) topical report PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review," establishes the definitions, processes, and technical requirements necessary to implement newly developed methods, issued July 2020 (ADAMS Accession No. ML20213C660). RG 1.200, Revision 3, endorsed specified portions of PWROG-19027-NP.

3.2 No Significant Hazards Consideration Determination Constellation Energy Generation, LLC (CEG) has evaluated the additional proposed changes to the Technical Specifications (TS) using the criteria in 10 CFR 50.92 and has determined that the proposed changes do not involve a significant hazards consideration.

TSTF-591 is an approved change to the Standard Technical Specifications (STS) that modifies the TSTF-505 new program description entitled the "Risk Informed Completion Time Program",

to reference Regulatory Guide (RG) 1.200, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 3, instead of Revision 2, and to make other administrative changes. TSTF-591 modified TS Section 5.6, "Reporting Requirements," to add a new requirement to inform the NRC of newly developed methods used to calculate a RICT.

As required by 10 CFR 50.91(a), an analysis of the issue of no significant hazards consideration is presented below:

1.

Do the proposed changes involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change updates the standard for maintaining and updating PRA models used to calculate a RICT from NRC-approved Regulatory Guide 1.200, Revision 2, to NRC-approved Regulatory Guide 1.200, Revision 3. A new report is added to inform the NRC when a newly developed RICT method is used. These proposed changes do not involve a significant increase in the probability of an accident previously evaluated because the changes involve no change to the plant or its modes of operation. The proposed changes do not increase the consequences of an accident because the design-basis mitigation function of the affected systems is not changed and the consequences of an accident during the extended Completion Time are no different from those during the existing Completion Time.

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

2.

Do the proposed changes create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

QCNPS Request to Adopt TSTF-505 Evaluation of the TSTF-591 Proposed Change Page 5 of 9 Docket Nos. 50-254 and 50-265 The proposed change does not change the design, configuration, or method of operation of the plant, with the exception of the flood mitigation strategy previously described in the original request to adopt TSTF-505. No new changes to the design, configuration, or method of operation of the plant are associated with the additional adoption of TSTF-591.

The proposed modification described in the TSTF-505 adoption amendment request only delays the time of flood water entry into the station without altering the mitigation strategy. The functionality, design, and implementation of the LIP barriers remains unchanged, except for minor upgrades such as increasing the height of one barrier. The current Updated Final Safety Analysis Report (UFSAR) described flood mitigation strategy remains available to mitigate the higher magnitude, low frequency floods that overtop the barriers at 599'.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3.

Do the proposed changes involve a significant reduction in a margin of safety?

Response: No.

The TSTF-505 adoption amendment request permits the extension of Completion Times provided that risk is assessed and managed in accordance with the NRC approved Risk Informed Completion Time Program (RICT). The TSTF-591 addition updates the standard for maintaining and updating PRA models used to calculate a RICT from NRC-approved Regulatory Guide (RG) 1.200, Revision 2, to NRC-approved RG 1.200, Revision 3. TSTF-591 also adds a new report requirement to inform the NRC when a newly developed RICT method is used. These additional changes help ensure the implemented risk-informed configuration management program assures adequate margins of safety are maintained.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above, CEG concludes that the proposed changes present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

3.2 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public. This is the same conclusion as the evaluation previously performed for the TSTF-505 adoption amendment request submitted by CEG letter dated June 8, 2023 (ADAMS Accession No. ML23159A249).

QCNPS Request to Adopt TSTF-505 Evaluation of the TSTF-591 Proposed Change Page 6 of 9 Docket Nos. 50-254 and 50-265 4.0 ENVIRONMENTAL EVALUATION The proposed changes would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed changes do not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed changes meet the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed changes. This is the same conclusion as the evaluation previously performed for the TSTF-505 adoption amendment request.

ATTACHMENT 3 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Response to Request for Additional Information Related 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" Revised Proposed Technical Specification Changes (Mark-Ups)

TS Pages 1.3-13 3.3.2.2-1 3.3.4.1-1 3.3.5.1-6 and -7 3.3.6.1-1 3.6.1.7-2 (removed) 3.8.4-2 3.8.7-1 and -2 5.5-14 5.6-5

Completion Times 1.3 Quad Cities 1 and 2 1.3-13 Amendment No. 199/195 1.3 Completion Times EXAMPLES EXAMPLE 1.3-7 (continued) is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.

IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner.

Insert A here

Insert A Example 1.3-8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem inoperable.

A.1 Restore subsystem to OPERABLE status.

7 days OR In accordance with the Risk Informed Completion Time Program B. Required Action and associated Completion Time not met.

B.1 Be in MODE 3.

AND B.2 Be in MODE 4.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 36 hours When a subsystem is declared inoperable, Condition A is entered. The 7 day Completion Time may be applied as discussed in Example 1.3-2. However, the licensee may elect to apply the Risk Informed Completion Time Program which permits calculation of a Risk Informed Completion Time (RICT) that may be used to complete the Required Action beyond the 7 day Completion Time. The RICT cannot exceed 30 days. After the 7 day Completion Time has expired, the subsystem must be restored to OPERABLE status within the RICT or Condition B must also be entered.

The Risk Informed Completion Time Program requires recalculation of the RICT to reflect changing plant conditions. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

If the 7 day Completion Time clock of Condition A has expired and subsequent changes in plant condition result in exiting the applicability of the Risk Informed Completion Time Program without restoring the inoperable subsystem to

OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start.

If the RICT expires or is recalculated to be less than the elapsed time since the Condition was entered and the inoperable subsystem has not been restored to OPERABLE status, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable subsystems are restored to OPERABLE status after Condition B is entered, Condition A is exited, and therefore, the Required Actions of Condition B may be terminated.

Feedwater System and Main Turbine High Water Level Trip Instrumentation 3.3.2.2 Quad Cities 1 and 2 3.3.2.2-1 Amendment No. 230/225 3.3 INSTRUMENTATION 3.3.2.2 Feedwater System and Main Turbine High Water Level Trip Instrumentation LCO 3.3.2.2 Four channels of Feedwater System and main turbine high water level trip instrumentation shall be OPERABLE.

APPLICABILITY:

THERMAL POWER 25% RTP.

---

NOTE---------------------------------------

Separate Condition entry is allowed for each channel.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more Feedwater System and main turbine high water level trip channels inoperable.

A.1

---

NOTE--------

Not applicable if inoperable channel is the result of an inoperable feedwater pump breaker.

Place channel in Trip.

7 days B.

Feedwater System and main turbine high water level trip capability not maintained.

B.1 Restore trip capability.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (continued)

Insert B here delete delete

ATWS-RPT Instrumentation 3.3.4.1 Quad Cities 1 and 2 3.3.4.1-1 Amendment No. 199/195 3.3 INSTRUMENTATION 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation LCO 3.3.4.1 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE:

a.

Reactor Vessel Water LevelLow Low; and b.

Reactor Vessel Steam Dome PressureHigh.

APPLICABILITY:

MODE 1.

ACTIONS

---

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more channels inoperable.

A.1 Restore channel to OPERABLE status.

OR A.2

---

NOTE---------

Not applicable if inoperable channel is the result of an inoperable breaker.

Place channel in trip.

14 days 14 days (continued)

Insert B here Insert B here

ECCS Instrumentation 3.3.5.1 Quad Cities 1 and 2 3.3.5.1-6 Amendment No. 199/195 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F.

As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

F.1 Declare Automatic Depressurization System (ADS) valves inoperable.

AND F.2 Place channel in trip.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of loss of ADS initiation capability in both trip systems 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from discovery of inoperable channel concurrent with HPCI or reactor core isolation cooling (RCIC) inoperable AND 8 days (continued)

Insert C here Insert F here Insert F wording:

or in accordance with the Risk Informed Completion Time Program

ECCS Instrumentation 3.3.5.1 Quad Cities 1 and 2 3.3.5.1-7 Amendment No. 199/195 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G.

As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

G.1 Declare ADS valves inoperable.

AND G.2 Restore channel to OPERABLE status.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> from discovery of loss of ADS initiation capability in both trip systems 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> from discovery of inoperable channel concurrent with HPCI or RCIC inoperable AND 8 days H.

Required Action and associated Completion Time of Condition B, C, D, E, F, or G not met.

H.1 Declare associated supported feature(s) inoperable.

Immediately Insert C here Insert F here Insert F wording:

or in accordance with the Risk Informed Completion Time Program

Primary Containment Isolation Instrumentation 3.3.6.1 Quad Cities 1 and 2 3.3.6.1-1 Amendment No. 199/195 3.3 INSTRUMENTATION 3.3.6.1 Primary Containment Isolation Instrumentation LCO 3.3.6.1 The primary containment isolation instrumentation for each Function in Table 3.3.6.1-1 shall be OPERABLE.

APPLICABILITY:

According to Table 3.3.6.1-1.

ACTIONS

---

NOTE-------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more required channels inoperable.

A.1 Place channel in trip.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for Functions 1.a, 2.a, 2.b, 3.d, 5.b, and 6.b AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for Functions other than Functions 1.a, 2.a, 2.b, 3.d, 5.b, and 6.b B.

One or more automatic Functions with isolation capability not maintained.

B.1 Restore isolation capability.

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (continued)

Insert B here Insert B here

DC SourcesOperating 3.8.4 Quad Cities 1 and 2 3.8.4-2 Amendment No. 199/195 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B.

---

NOTE---------

Only applicable if the opposite unit is in MODE 1, 2, or 3.

Division 1 or 2 125 VDC battery inoperable as a result of maintenance or testing.

B.1 Place associated OPERABLE alternate 125 VDC electrical power subsystem in service.

AND B.2 Restore Division 1 or 2 125 VDC battery to OPERABLE status.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Prior to exceeding 7 cumulative days per operating cycle of battery inoperability, on a per battery basis, as a result of maintenance or testing C.

---

NOTE---------

Only applicable if the opposite unit is in MODE 1, 2, or 3.

Division 1 or 2 125 VDC battery inoperable, due to the need to replace the battery, as determined by maintenance or testing.

C.1 Place associated OPERABLE alternate 125 VDC electrical power subsystem in service.

AND C.2 Restore Division 1 or 2 125 VDC battery to OPERABLE status.

72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 7 days (continued)

Insert B here

Distribution SystemsOperating 3.8.7 Quad Cities 1 and 2 3.8.7-1 Amendment No. 275/270 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Distribution SystemsOperating LCO 3.8.7 The following electrical power distribution subsystems shall be OPERABLE:

a.

Division 1 and Division 2 AC and DC electrical power distribution subsystems; and

b.

The portions of the opposite unit's AC and DC electrical power distribution subsystems necessary to support equipment required to be OPERABLE by LCO 3.6.4.3, "Standby Gas Treatment (SGT) System," LCO 3.7.4, "Control Room Emergency Ventilation (CREV) System" (Unit 2 only), LCO 3.7.5, "Control Room Emergency Ventilation Air Conditioning (AC) System" (Unit 2 only),

and LCO 3.8.1, "AC SourcesOperating."

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more AC electrical power distribution subsystems inoperable.

A.1 Restore AC electrical power distribution subsystems to OPERABLE status.

8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (continued)

Insert B here delete

Distribution SystemsOperating 3.8.7 Quad Cities 1 and 2 3.8.7-2 Amendment No. 275/270 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. One or more DC electrical power distribution subsystems inoperable.

B.1 Restore DC electrical power distribution subsystems to OPERABLE status.

2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. One or more required opposite unit AC or DC electrical power distribution subsystems inoperable.

---

NOTE------------

Enter applicable Condition and Required Actions of LCO 3.8.1 when Condition C results in the inoperability of a required offsite circuit.

C.1 Restore required opposite unit AC and DC electrical power distribution subsystems to OPERABLE status.

7 days D. Required Action and associated Completion Time of Condition A, B, or C not met.

---

NOTE-----------

LCO 3.0.4.a is not applicable when entering MODE 3.

D.1 Be in MODE 3.

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> E. Two or more electrical power distribution subsystems inoperable that, in combination, result in a loss of function.

E.1 Enter LCO 3.0.3.

Immediately Insert B here Insert B here delete

Programs and Manuals 5.5 5.5 Programs and Manuals Quad Cities 1 and 2 5.5-14 Amendment No. 248/243 5.5.14 Surveillance Frequency Control Program This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, Risk-Informed Method for Control of Surveillance Frequencies, Revision 1.

c.

The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

INSERT E

Reporting Requirements 5.6 5.6 Reporting Requirements Quad Cities 1 and 2 5.6-5 Amendment No. 294/290 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

18. EMF-2292 (P)(A) Revision 0, "ATRIUMTM-10: Appendix K Spray Heat Transfer Coefficients," Siemens Power Corporation, September 2000.

19. ANF-1358(P)(A) Revision 3, "The Loss of Feedwater Heating Transient in Boiling Water Reactors," Framatome ANP, September 2005.

20. EMF-CC-074(P)(A) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," Siemens Power Corporation, August 2000.

21. NEDC-33930P, Revision 0, "GEXL98 Correlation for ATRIUM 10XM Fuel," February 2021, as approved by NRC Staff SE dated December 15, 2022.

22. 006N8642-P, Revision 1, "Justification of PRIME Methodologies for Evaluating TOP and MOP Compliance for non-GNF Fuels," January 2022, as approved by NRC Staff SE dated February 6, 2023.

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i.e., report number, title, revision, date, and any supplements).

c.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d.

The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

INSERT F

Insert E 5.5.15 Risk Informed Completion Time Program This program provides controls to calculate a Risk Informed Completion Time (RICT) and must be implemented in accordance with NEI 06-09-A, Revision 0, "Risk-Managed Technical Specifications (RMTS) Guidelines." The program shall include the following:

a.

The RICT may not exceed 30 days; b.

A RICT may only be utilized in MODE 1 and 2; c.

When a RICT is being used, any change to the plant configuration, as defined in NEI 06-09-A, Appendix A, must be considered for the effect on the RICT.

1. For planned changes, the revised RICT must be determined prior to implementation of the change in configuration.

2. For emergent conditions, the revised RICT must be determined within the time limits of the Required Action Completion Time (i.e., not the RICT) or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the plant configuration change, whichever is less.

3. Revising the RICT is not required if the plant configuration change would lower plant risk and would result in a longer RICT.

d.

For emergent conditions, if the extent of condition evaluation for inoperable structures, systems, or components (SSCs) is not complete prior to exceeding the Completion Time, the RICT shall account for the increased possibility of common cause failure (CCF) by either:

1. Numerically accounting for the increased possibility of CCF in the RICT calculation; or

2. Risk Management Actions (RMAs) not already credited in the RICT calculation shall be implemented that support redundant or diverse SSCs that perform the function(s) of the inoperable SSCs, and, if practicable, reduce the frequency of initiating events that challenge the function(s) performed by the inoperable SSCs.

e.

A RICT calculation must include the following hazard groups:

internal flood and internal events using a PRA model, internal fires using a PRA model, seismic hazards using penalty factors, and extreme wind and tornado hazards using configuration-specific penalty factors for tornado missiles.

Changes to these means of assessing the hazard groups require prior NRC approval.

f.

The PRA models used to calculate RICT shall be maintained and upgraded in accordance with the processes endorsed in the regulatory positions of Regulatory Guide 1.200, Revision 3, "Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities."

g.

A report shall be submitted in accordance with Specification 5.6.7 before a newly developed method is used to calculate a RICT.

Insert F 5.6.7 Risk Informed Completion Time (RICT) Program Upgrade Report A report describing newly developed methods and their implementation must be submitted following a probabilistic risk assessment (PRA) upgrade associated with newly developed methods and prior to the first use of those methods to calculate a RICT.

The report shall include:

a.

The PRA models upgraded to include newly developed methods; b.

A description of the acceptability of the newly developed methods consistent with Section 5.2 of PWROG-19027-NP, Revision 2, "Newly Developed Method Requirements and Peer Review;"

c.

Any open findings from the peer-review of the implementation of the newly developed methods and how those findings were dispositioned; and d.

All changes to key assumptions related to newly developed methods or their implementation.

ATTACHMENT 4 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Response to Request for Additional Information Related 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" Revised Table A5-1 RICT Program Implementation Items

QCNPS Request to Adopt TSTF-505 Revised Table A5-1 RICT Program Implementation Items Page 1 of 1 Docket Nos. 50-254 and 50-265 Table A5-1 QCNPS RICT Program Implementation Items Source Description Implementation Item, Table E1-1, Technical Specification (TS) 3.6.1.7.A One or more lines with one reactor building-to-suppression chamber vacuum breaker not closed.

The structures, systems, and components (SSCs) are not modeled. The model will be updated to include these SSCs prior to exercising the RICT program for this TS., Table E1-1, Technical Specification (TS) 3.6.1.7.C One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

The SSCs are not modeled.

The model will be updated to include these SSCs prior to exercising the RICT program for this TS., Table E1-1, Technical Specification (TS) 3.6.1.7.E Two lines with one reactor building-to-suppression chamber vacuum breakers inoperable for opening.

The SSCs are not modeled.

The model will be updated to include these SSCs prior to exercising the RICT program for this TS., Section 5.2.3.3, LIP Flood Barrier Upgrades and Deployment LIP barriers are modified to protect the plant up to 599.0' Complete EC 636914, Update to LIP Barriers to Assist the Station External Flood, and EC 636912, Update to Station External Flood Response in Support of Risk Reduction, modifications (scope includes both physical plant and documentation / procedure changes).

ATTACHMENT 5 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Response to Request for Additional Information Related 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" Revised Table E1-1 In-Scope TS/LCO Conditions to Corresponding PRA Functions

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 1 of 17 E1-1 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.1.7.A One Standby Liquid Control (SLC) subsystem inoperable.

Two SLC trains Yes Provide a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown.

One of two trains Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.3.1.1.A One or more required channels inoperable.

Reactor Protection System (RPS)

Instrumentation outlined in Table 3.3.1.1-1 Not explicitly Provide reactor trip signal based on plant parameters.

One of two channels, taken twice See Section 5.1 for a detailed RPS signal diversity discussion.

Same as Design Success Criteria Individual RPS instrumentation inputs to the RPS logic system are not modeled in the PRA.

Common cause failure of the RPS electrical system is used as a conservative surrogate for failure of the RPS.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 2 of 17 E1-2 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.1.1.B One or more Functions with one or more required channels inoperable in both trip systems.

RPS Instrumentation outlined in Table 3.3.1.1-1 Not explicitly Provide reactor trip signal based on plant parameters.

One of two channels, taken twice See Section 5.1 for a detailed RPS signal diversity discussion.

Same as Design Success Criteria Individual RPS instrumentation inputs to the RPS logic system are not modeled in the PRA.

Common cause failure of the RPS electrical system is used as a conservative surrogate for failure of the RPS.

3.3.2.2.A One or more Feedwater System and main turbine high water level trip channels inoperable.

Reactor high level Feedwater System and main turbine trip instrumentation Not explicitly Feedwater System and main turbine trip to prevent water intrusion into turbines Two of four channels See Section 5.2 for a detailed diversity discussion.

Same as Design Success Criteria Common cause failure of high level feedwater trip failure will be used as a conservative surrogate for failure of any channel of feedwater instrumentation.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 3 of 17 E1-3 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.4.1.A One or more channels inoperable.

Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)

System includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of a recirculation pump trip Not explicitly Recirculation pump trip One of two channels See Section 5.3 for a detailed ATWS-RPT signal diversity discussion.

Same as Design Success Criteria Failure of recirculation pump breakers will be used as a surrogate for failure of the ATWS-RPT instrumentation 3.3.5.1.B As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

Emergency Core Cooling System (ECCS) actuation instrumentation for Core Spray (CS),

Low Pressure Coolant Injection (LPCI), High Pressure Coolant Injection (HPCI),

Diesel Generators (DGs)

See Notes 2 and 3 Yes (1.a, 1.b, 2.a, 2.b, 2.d, 2.j), not explicitly (3.a, 3.b)

Initiate ECCS (CS, LPCI, associated DG, HPCI)

Success of the ECCS instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.1-1.

See Section 5.4 for a detailed signal diversity discussion.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 4 of 17 E1-4 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.5.1.C As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

ECCS actuation instrumentation for CS, LPCI, DGs See Notes 2 and 3 Yes (1.c, 1.e, 2.c, 2.e, 2.g, 2.h, 2.i, 2.k), not explicitly 3.c, 3.g)

Initiate ECCS (CS, LPCI, associated DG)

Success of the ECCS instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.1-1.

See Section 5.4 for a detailed signal diversity discussion.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.3.5.1.D As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

ECCS actuation instrumentation for HPCI Not explicitly Initiate HPCI Success of the ECCS instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.1-1.

See Section 5.4 for a detailed signal diversity discussion.

Same as Design Success Criteria CCST level-low and suppression pool water level-high HPCI instrumentation not explicitly modeled, therefore the fail to start is used as surrogate.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 5 of 17 E1-5 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.5.1.E As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

ECCS actuation instrumentation for CS, LPCI, HPCI, DGs See Notes 2 and 3 Not explicitly Initiate ECCS (CS, LPCI, associated DG)

Success of the ECCS instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.1-1.

See Section 5.4 for a detailed signal diversity discussion.

Same as Design Success Criteria ECCS pump discharge flow-low, bypass, not explicitly modeled. Minimum flow valves fail to open used as surrogates.

3.3.5.1.F As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

ECCS actuation instrumentation for Automatic Depressurization System (ADS)

Not explicitly Initiate ADS Success of the ECCS instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.1-1.

See Section 5.4 for a detailed signal diversity discussion.

Same as Design Success Criteria ADS actuation instrumentation not explicitly modeled; relief valves fail to open used as surrogates.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 6 of 17 E1-6 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.5.1.G As required by Required Action A.1 and referenced in Table 3.3.5.1-1.

ECCS actuation instrumentation for ADS Not explicitly Initiate ADS Success of the ECCS instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.1-1.

See Section 5.4 for a detailed signal diversity discussion.

Same as Design Success Criteria ADS actuation instrumentation not explicitly modeled; relief valves fail to open used as surrogates.

3.3.5.3.B As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

Reactor Vessel level Instrumentation supporting Reactor Core Isolation Cooling (RCIC) automatic initiation Yes Initiate RCIC Success of the RCIC instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.3-1.

See Section 5.5 for a detailed RCIC signal diversity discussion.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 7 of 17 E1-7 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.5.3.D As required by Required Action A.1 and referenced in Table 3.3.5.3-1.

CCST and suppression pool level instrumentation supporting RCIC automatic initiation Not explicitly Initiate RCIC Success of the RCIC instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.5.3-1.

See Section 5.5 for a detailed RCIC signal diversity discussion.

Same as Design Success Criteria CCST level-low and torus water level-high RCIC instrumentation not explicitly modeled, therefore fail to open for injection valves are used as surrogates.

3.3.6.1.A One or more required channels inoperable Primary Containment Isolation Instrumentation outlined in Table 3.3.6.1-1 Not explicitly Initiate closure of primary containment isolation valves Success of the primary containment instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.6.1-1.

See Section 5.6 for a detailed signal diversity discussion.

Same as Design Success Criteria The logic for primary containment isolation is not modeled in detail. Therefore, a surrogate is chosen that represents a failure of the containment isolation signal.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 8 of 17 E1-8 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.3.6.3.A One relief valve inoperable due to inoperable channel(s).

Relief Valve Set Instrumentation outlined in Table 3.3.6.3-1 Not explicitly Mitigate overpressurization transients Success of the Relief Valve Set instrumentation is dependent upon individual instrumentation channel functions outlined in Table 3.3.6.3-1 Same as Design Success Criteria Relief valve instrumentation is not explicitly modeled, therefore fail to open for the relief valves are used as surrogates.

3.3.8.1.A One or more channels inoperable.

Loss of Power (LOP) instrumentation includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to trip offsite power circuits and start the DGs Not explicitly Undervoltage functions for each 4160 V ESS bus Two of two trains See Section 5.7 for a detailed LOP signal diversity discussion.

Failure of DG autostart, load shed, and undervoltage relays will be used as a surrogate for failure of the LOP instrumentation.

Failure of DG autostart, load shed, and undervoltage relays will be used as a surrogate for failure of the LOP instrumentation.

3.4.3.A One relief valve inoperable.

Five relief valves (Four ERV and one Target Rock Valve)

See Note 1 Yes Relieve pressure during a limiting event and maintain ASME Code limit on reactor overpressure Four of five relief valves One valve (non-ATWS), 12 of 13 valves (ATWS)

SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 9 of 17 E1-9 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.5.1.B One LPCI subsystem inoperable for reasons other than Condition A OR one Core Spray subsystem inoperable.

Two LPCI subsystems, two CS subsystems Yes Low pressure injection into the reactor pressure vessel (RPV)

Two LPCI pumps, one CS pump One LPCI subsystem or one CS subsystem SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.5.1.C One LPCI pump in each subsystem inoperable.

Two LPCI subsystems Yes Low pressure injection into the RPV One LPCI pump in each subsystem One of two LPCI subsystems SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.5.1.E Two LPCI subsystems inoperable for reasons other than Condition C.

Two LPCI subsystems Yes Low pressure injection into the RPV Low pressure injection function is achieved through two redundant CS subsystems One train of one low pressure source SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.5.1.G HPCI System inoperable.

HPCI System Yes High pressure injection into the RPV Injection function is achieved by RCIC being operable and low pressure ECCS injection/spray subsystems in conjunction with ADS Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 10 of 17 E1-10 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.5.1.H One ADS valve inoperable.

Five ADS valves (Four ERV, one Target Rock valve)

See Note 1 Yes Provide depressurization of RCS during small break LOCA if HPCI cannot maintain RPV level Four of five valves Two of five valves SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.5.3.A RCIC System inoperable.

RCIC System Yes Provide makeup water and maintain RPV level above top of core following RPV isolation with loss of coolant flow from feedwater system HPCI is operable Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.6.1.2.C Primary containment air lock inoperable for reasons other than Condition A or B.

Primary containment air lock equipment Not explicitly Primary containment isolation One of two primary containment air lock doors maintain boundary Maintain primary containment The primary containment air locks are not modeled therefore a pre-existing containment failure will be used as a conservative surrogate.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 11 of 17 E1-11 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.6.1.3.A One or more penetration flow paths with one primary containment isolation valve (PCIV) inoperable for reasons other than Condition D.

PCIVs Not explicitly Minimize the loss of reactor coolant inventory and establish the primary containment boundary during accidents One of two isolation valves per penetration Same as Design Success Criteria Not all PCIVs are modeled therefore a pre-existing containment failure will be used as a conservative surrogate.

3.6.1.6.A One low set relief valve inoperable.

Two low set relief valves Not explicitly Prevents excessive short duration S/RV cycles with valve actuation at the relief setpoint.

One of two valves Same as Design Success Criteria The low set mode of the relief valves is not explicitly modeled in the PRA, however the failure of the relief valves to open can be used as a surrogate since failure of the relief valves to open bounds the risk of the low set mode.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 12 of 17 E1-12 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.6.1.7.A One or more lines with one reactor building-to-suppression chamber vacuum breaker not closed.

Reactor building-to-suppression chamber vacuum breakers No Relieve vacuum when primary containment depressurizes below reactor building pressure One of vacuum breakers closed in each of the two lines N/A The SSCs are not modeled. The model will be updated to include these SSCs prior to exercising the RICT program for this TS. This is an Implementation Item identified in.

3.6.1.7.C One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

See TS 3.6.1.7.A Two vacuum breakers open in one of two paths See TS 3.6.1.7.A 3.6.1.7.E Two lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

See TS 3.6.1.7.A.

Two vacuum breakers open in one of two paths N/A This is a loss of function condition and RICT should not be applied.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 13 of 17 E1-13 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.6.1.8.A One required suppression chamber-to-drywell vacuum breaker inoperable for opening.

Suppression chamber-to-drywell vacuum breakers Not explicitly Relieve vacuum in the drywell 7 of 12 vacuum breakers Same as Design Success Criteria The opening function of the suppression chamber to drywell vacuum breakers is not modeled in the PRA. The vapor suppression function is modeled and is used as a surrogate.

3.6.2.3.A One RHR suppression pool cooling subsystem inoperable.

Two Residual Heat Removal (RHR) subsystems Yes Maintain primary containment peak pressure and temperature below design limits One of four RHR loops One of four RHR loops SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.6.2.6.A One RHR drywell spray subsystem inoperable.

Two RHR subsystems See Note 4 Yes Adequately scrub inorganic iodines and particulates from primary containment atmosphere One of four RHR loops One of four RHR loops SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.7.1.B One RHRSW pump in each subsystem inoperable.

Two Residual Heat Removal Service Water (RHRSW) subsystems Yes Provide cooling water for the RHR system heat exchangers following a DBA or transient One of two pumps in each subsystem One of two pumps in each subsystem SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 14 of 17 E1-14 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.7.1.C One RHRSW subsystem inoperable for reasons other than Condition A.

See TS 3.7.1.B.

One of two subsystems One of two subsystems See TS 3.7.1.B.

3.7.9.A Safe Shutdown Makeup Pump (SSMP) System inoperable SSMP Yes Provide RPV makeup during RPV isolation with loss of feedwater flow One of one train Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.1.A One required offsite circuit inoperable.

Reserve Auxiliary Transformers (RATs), Unit Auxiliary Transformers (UATs),

associated breakers, and offsite power supplies Yes Provide power to safety related buses One of two offsite AC power sources See Section 7 for additional details.

As needed to supply supported functions SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.1.B One required DG inoperable.

Three DGs and support systems Yes Provide power to safety related buses when offsite power to them is lost Two out of three DGs See Section 7 for additional details.

One out of three DGs SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 15 of 17 E1-15 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.8.1.C Two required offsite circuits inoperable.

RATs, UATs, associated breakers, and offsite power supplies Yes Provide power to safety related buses Two out of three DGs See Section 7 for additional details.

As needed to supply supported functions SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.1.D One required offsite circuit inoperable AND one required DG inoperable.

RATs, UATs, associated breakers, offsite power supplies, three DGs and support systems Yes Provide power to safety related buses Two required sources (one offsite circuit and one DG)

See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.4.A One 250 VDC electrical power subsystem inoperable.

Two 250 VDC subsystems Yes Supply DC loads during normal operation One out of two subsystems See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.4.C Division 1 or 2 125 VDC battery inoperable, due to the need to replace the battery, as determined by maintenance or testing.

Division 1 and Division 2 125 VDC batteries Yes Supply DC loads during normal operation One out of two subsystems See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 16 of 17 E1-16 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.8.4.D Division 1 or 2 125 VDC electrical power subsystem inoperable for reasons other than Condition B or C.

Division 1 and Division 2 125 VDC subsystem Yes Supply DC loads during normal operation One out of two subsystems See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.4.E Opposite unit 125 VDC electrical power subsystem inoperable.

Opposite unit Division 2 subsystem Yes Supply DC loads during normal operation One out of one subsystem See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.7.A One or more AC electrical power distribution subsystems inoperable.

Two AC distribution subsystems, associated buses, MCCS, and distribution panels Yes Provide AC power distribution to required divisional loads to shut down reactor and maintain in safe condition One of two subsystems See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

3.8.7.B One or more DC electrical power distribution subsystems inoperable.

Two 125 VDC distribution subsystems, two 250 VDC subsystems Yes Provide DC power distribution to required divisional loads to shut down reactor and maintain in safe condition One of two subsystems for 125 and 250 VDC See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

QCNPS Request to Adopt TSTF-505 Revised Table E1-1 In-Scope TS/LCO Docket Nos. 50-254 and 50-265 Page 17 of 17 E1-17 Table E1-1: In-Scope TS/LCO Conditions to Corresponding PRA Functions TS Condition TS Condition Description SSCs Covered by TS LCO SSCs Modeled in PRA?

Function Covered by TS LCO Design Success Criteria PRA Success Criteria Other Comments 3.8.7.C One or more required opposite unit AC or DC electrical power distribution subsystems inoperable.

Two AC distribution subsystems, associated buses, MCCS, distribution panels, two 125 VDC distribution subsystems, two 250 VDC subsystems Yes Provide AC/DC power distribution to required divisional loads to shut down reactor and maintain in safe condition One of two subsystems for each AC/DC distribution system See Section 7 for additional details.

Same as Design Success Criteria SSCs modeled in PRA consistent with TS scope and can be explicitly included in the RTR tool for the RICT program.

Table E1-1 Notes:

1. The five relief valves consist of four electromagnetic relief valves (ERV) and the Target Rock safety/relief valve. The Target Rock valve performs both relief valve and safety valve functions and is credited as one of each valve.

2. QCNPS has two independent core spray divisions (subsystems or loop). Each division consists of a 4500 gal/min capacity pump, valves, piping and an independent circular sparger ring inside the inner shroud just over the core. Suction water is supplied by the suppression pool.

3. LPCI is a functional mode of the RHR system. LPCI has two divisions (subsystems or loop) which consists of a heat exchanger, two RHR pumps in parallel, and associated piping. The two divisions of LPCI are cross connected by a single header, making it possible to supply either division from the pumps in the other division.

4. Each of the two RHR drywell spray subsystems contains two pumps, one heat exchanger, drywell spray valves, and a spray header inside the drywell. Each RHR drywell spray subsystem is capable of recirculating water from the RHR suppression pool through a heat exchanger and dispersed through the RHR drywell spray nozzles. The LOCA radiological dose analysis credits the RHR drywell spray system for scrubbing radionuclides from the drywell air space.

ATTACHMENT 6 Quad Cities Nuclear Power Station Docket Nos. 50-254 and 50-265 Facility Operating License Nos. DPR-29 and DPR-30 Response to Request for Additional Information Related 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" Excerpts of Revised TSTF-505/QCNPS Cross-References from Attachment 4

QCNPS Request to Adopt TSTF-505 Excerpts of Revised TSTF-505/QCNPS Cross-References from Attachment 4 Page 1 of 

Docket Nos. 50-254 and 50-265 Tech Spec Description TSTF-505 Tech Spec QCNPS Tech Spec Apply RICT?

Comments One line with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

Restore the vacuum breaker(s) to OPERABLE status.

3.6.1.7.C.1 3.6.1.7.C.1 Yes TSTF-505 change is incorporated.

Two lines with one or more reactor building-to-suppression chamber vacuum breakers inoperable for opening.

Restore all vacuum breakers in one line to OPERABLE status.

3.6.1.7.D.1 3.6.1.7.E.1 No The loss of a single breaker in both lines represents a loss of function, which is exclusion criterion 1.

Therefore, QCNPS 3.6.1.7.E.1 is outside the scope of TSTF-505.

Suppression Chamber-to-Drywell Vacuum Breakers 3.6.1.8 3.6.1.8 One required suppression chamber-to-drywell vacuum breaker inoperable for opening.

Restore one vacuum breaker to OPERABLE status.

3.6.1.8.A.1 3.6.1.8.A.1 Yes TSTF-505 changes are incorporated.