Text

Issuance of Amendment No. 270 Adoption of TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b
	ML22322A109
Person / Time
Site:	Waterford
Issue date:	02/17/2023
From:	James Drake; Plant Licensing Branch IV
To:	; Entergy Operations
	Klett A
References
	EPID L-2021-LLA-0014
	Download: ML22322A109 (52)
	v • d • e

February 17, 2023 Site Vice President Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093

SUBJECT:

WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT NO. 270 RE: ADOPTION OF TSTF-505, REVISION 2, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4b (EPID L-2021-LLA-0014)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 270 to Renewed Facility Operating License (RFOL) No. NPF-38 for the Waterford Steam Electric Station, Unit 3. The amendment consists of changes to the technical specifications (TSs) in response to your application dated February 8, 2021, as supplemented by letters dated April 8, 2021, May 16, 2022, August 19, 2022, and October 13, 2022.

The amendment revises the TS requirements to permit the use of risk-informed completion times in accordance with Technical Specifications Task Force (TSTF) Traveler (TSTF-505),

Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b.

A copy of the related Safety Evaluation is also enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Jason J. Drake, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-382

Enclosures:

1. Amendment No. 270 to NPF-38

2. Safety Evaluation cc: Listserv

ENTERGY OPERATIONS, INC.

DOCKET NO. 50-382 WATERFORD STEAM ELECTRIC STATION, UNIT 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 Renewed License No. NPF-38

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Operations, Inc. (EOI), dated February 8, 2021, as supplemented by letters dated April 8, 2021, May 16, 2022, August 19, 2022, and October 13, 2022, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2 of Renewed Facility Operating License No. NPF-38 is hereby amended to read as follows:

2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 270, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of its date of issuance and shall be implemented within 120 days from the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jennifer L. Jennifer L. Dixon-Herrity Date: 2023.02.17 Dixon-Herrity 11:32:30 -05'00' Jennifer L. Dixon-Herrity, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. NPF-38 and the Technical Specifications Date of Issuance: February 17, 2023

ATTACHMENT TO LICENSE AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38 WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382 Replace the following pages of Renewed Facility Operating License No. NPF-38 and the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility Operating License REMOVE INSERT Technical Specifications REMOVE INSERT 3/4 3-4a 3/4 3-4a 3/4 3-17 3/4 3-17 3/4 3-18 3/4 3-18 3/4 3-18a 3/4 3-18a 3/4 3-18b 3/4 3-18b 3/4 5-3 3/4 5-3 3/4 6-9 3/4 6-9 3/4 6-15 3/4 6-15 3/4 6-16 3/4 6-16 3/4 6-18 3/4 6-18 3/4 6-19 3/4 6-19 3/4 6-20 3/4 6-20 3/4 7-4 3/4 7-4 3/4 7-9 3/4 7-9 3/4 7-9b 3/4 7-9b 3/4 7-11 3/4 7-11 3/4 7-12 3/4 7-12 3/4 7-43 3/4 7-43 3/4 8-1 3/4 8-1 3/4 8-2 3/4 8-2 3/4 8-2a 3/4 8-2a 3/4 8-9 3/4 8-9 3/4 8-14 3/4 8-14 6-10 6-10

--- 6-11

the NRC of any action by equity investors or successors in interest to Entergy Louisiana, LLC that may have an effect on the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the Commissions regulations set forth in 10 CFR Chapter I and is subject to all applicable provisions of the Act and to the rules, regulations and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

1. Maximum Power Level EOI is authorized to operate the facility at reactor core power levels not in excess of 3716 megawatts thermal (100% power) in accordance with the conditions specified herein.

2. Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 270, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the renewed license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. Antitrust Conditions (a) Entergy Louisiana, LLC shall comply with the antitrust license conditions in Appendix C to this renewed license.

(b) Entergy Louisiana, LLC is responsible and accountable for the actions of its agents to the extent said agent's actions contravene the antitrust license conditions in Appendix C to this renewed license.

AMENDMENT NO. 270

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC. WATERFORD STEAM ELECTRIC STATION, UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By application dated February 8, 2021 (Reference 1), as supplemented by letters dated April 8, 2021 (Reference 2), May 16, 2022 (Reference 3), August 19, 2022 (Reference 4), and October 13, 2022 (Reference 5), Entergy Operations, Inc (Entergy, the licensee) submitted a license amendment request (LAR) for Waterford Steam Electric Station, Unit 3 (Waterford 3). The amendment would revise technical specification (TS) requirements to permit the use of risk-informed completion times (RICTs) for actions to be taken when limiting conditions for operation (LCOs) are not met. The proposed changes are based on Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF [Risk-Informed TSTF] Initiative 4b, dated July 2, 2018 (TSTF-505) (Reference 6). The U.S. Nuclear Regulatory Commission (NRC, the Commission) issued a final revised model safety evaluation (SE) to be used when preparing a plant-specific SE of an LAR to adopt TSTF-505, on November 21, 2018 (Reference 7). The licensee has proposed variations from the TS changes approved in TSTF-505, which are provided in section 2.3 of the LAR and evaluated in section 3.0 of this SE. The NRC staff participated in a regulatory audit in October 2021. The NRC staff performed the audit to ascertain the information needed to support its review of the application and develop requests for additional information (RAIs), as needed. On October 7, 2021, the NRC staff issued an audit plan (Reference 8). The supplemental letters dated May 16, 2022, August 19, 2022, and October 13, 2022, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on May 18, 2021 (86 FR 26954). Enclosure 2

2.0 REGULATORY EVALUATION

Title 10 of the Federal Code of Regulations (10 CFR) Part 50 provides the general provisions for Domestic Licensing of Production and Utilization Facilities. The general provisions include but are not limited to establishing the regulatory requirements that a licensee must adhere to for the submittal of a license application. The NRC staff has identified the following applicable sections within 10 CFR Part 50 for the staffs review of a licensees application to adopt TSTF-505. 10 CFR 50.36, Technical Specifications, paragraphs (c)(2), Limiting conditions for operation, and (c)(5), Administrative controls 10 CFR 50.55a, Codes and standards, paragraph (h), Protection and safety systems 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants (i.e., the Maintenance Rule) NRC Regulatory Guides (RGs) provide one way to ensure that the codified regulations continue to be met. The NRC staff considered the following guidance, along with industry guidance endorsed by the NRC, during its review of the proposed changes: RG 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated March 2009 (Reference 9) and RG 1.200, Revision 3, Acceptability of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated December 2020 (Reference 10). RG 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, dated January 2018 (Reference 11). RG 1.177, Revision 2, Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications, dated January 2021 (Reference 12). NRC Regulation (NUREG)-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs [Probabilistic Risk Assessments] in Risk-Informed Decisionmaking, dated March 2017 (Reference 13). NUREG-1432, Standard Technical Specifications Combustion Engineering Plants [STS], Volume 1, Specifications, and Volume 2, Bases, dated September 2021 (Reference 14). NUREG-0800, Standard Review Plan [SRP] for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition, Section 16.1, Risk-Informed Decision Making: Technical Specifications, dated March 2007 (Reference 15). The licensees submittals cite RG 1.200, Revision 2 as applicable guidance. The updates in RG 1.200, Revision 3 do not include any technical changes that would impact the consistency with Nuclear Energy Institute (NEI) Topical Report (TR) NEI 06-09-A, Risk-Informed Technical Specifications Initiative 4b, Risk-Managed Technical Specifications (RMTS) Guidelines dated October 2012 (Reference 16). The NRC staff issued a final model SE approving NEI 06-09 with limitations and conditions on May 17,

2007 (Reference 17). Therefore the NRC staff finds the RG 1.200, Revision 3 is also applicable for use in the adoption of TSTF-505. 2.1 Description of Risk-Informed Completion Time Program The TS LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When an LCO is not met, the licensee must shut down the reactor or follow any remedial or required action (e.g., testing, maintenance, or repair activity) permitted by the TSs until the condition can be met. The remedial actions (i.e., ACTIONS) associated with an LCO contain conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s) (CT). The CTs are referred to as the front stops in the context of this SE. For certain conditions, the TSs require exiting the Mode of Applicability of an LCO (i.e., shut down the reactor). The licensees submittal requested approval to add a RICT program to the Administrative Controls Section of the TS, and to modify selected CTs to permit extending the CTs, provided risk is assessed and managed as described in NEI 06-09-A. Consistent with the section on NUREG-1432, Standard Technical Specifications Combustion Engineering Plants (STS), in table 1, Conditions Requiring Additional Technical Justification, of TSTF-505 for Conditions Requiring Additional Technical Justification (Specifications), the licensee provided several plant-specific LCOs and associated Actions proposed to be included in the RICT program, along with additional justification. NRC staff review of these variations and the justification is provided in section 3.0 of this SE. The licensee is proposing no changes to the design of the plant or any operating parameter, and no new changes to the design-basis in the proposed changes to the TSs. The effect of the proposed changes when implemented will allow CTs to vary, based on the risk significance of the given plant configuration (i.e., the equipment out of service at any given time), provided that the system(s) retain(s) the capability to perform the applicable safety function(s) without any further failures (e.g., one train of a two-train system is inoperable). These restrictions on inoperability of all required trains of a system ensure that consistency with the defense-in-depth (DID) philosophy is maintained by following existing guidance when the capability to perform TS safety function(s) is lost. The proposed RICT program uses plant-specific operating experience for component reliability and availability data. Thus, the allowances permitted by the RICT program are directly reflective of actual component performance in conjunction with component risk significance.

3.0 TECHNICAL EVALUATION

3.1 Method of NRC Staff Review The NRC staff reviewed the licensees PRA peer review history and results, alternative methods, and proposed approaches to determine if they are technically acceptable for use in the proposed RICT extensions. The NRC staff also reviewed the licensees proposed RICT program to determine if it provides the necessary administrative controls to permit CT extensions for consistency with NEI 06-09-A. An acceptable approach for making risk-informed decisions about proposed TS changes, including both permanent and temporary changes, is to show that the proposed licensing basis

(LB) changes meet the five key principles provided in RGs 1.174 and 1.177 and the three-tiered approach outlined in 1.177. Key Principle 1: Evaluation of Compliance with Current Regulations Paragraph 50.36(c)(2) of 10 CFR states, in part, that LCOs are the lowest functional capability or performance levels of equipment required for safe operation of the facility and that when a limiting condition for operation of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the technical specifications until the LCO can be met. The CTs in the current TSs were established using experiential data, risk insights, and engineering judgement. The RICT program provides the necessary administrative controls to permit extension of CTs and, thereby, delay reactor shutdown or Required Actions, if risk is assessed and managed appropriately within specified limits and programmatic requirements and the safety margins and DID remains sufficient. The option to determine the extended CT in accordance with the RICT program allows the licensee to perform an integrated evaluation in accordance with the methodology prescribed in NEI 06-09-A and proposed TS 6.5.19, Risk Informed Completion Time Program. The RICT is limited to a maximum of 30 days (termed the back stop). The typical CT is modified by the application of the RICT program as shown in the following example. The changed portion is indicated in italics. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One subsystem A.1 Restore subsystem 7 days inoperable. to OPERABLE status. OR In accordance with the Risk-Informed Completion Time Program In attachment 2, Proposed Technical Specification Changes (Mark-up), and enclosure 1, List of Revised Required Actions to Corresponding PRA Functions, to the LAR, as supplemented, the licensee provided a list of the TSs, associated LCOs, and Required Actions for the CTs that included modifications and variations from the approved TSTF-505. The modifications and variations consisted of proposed changes to the Required Actions and CTs. Furthermore, consistent with table 1 of TSTF-505 for Waterford 3 TSs 3.4.3.1, 3.6.1.3, 3.7.1.6, 3.5.2, 3.6.2.1, and 3.6.2.2 in section 2.3 in attachment 1, Description and Assessment of the Proposed Change, to the LAR, as supplemented, the licensee included additional technical justification to demonstrate the acceptability for including these TSs in the RICT program. The NRC staff reviewed the proposed changes to the TSs, associated LCOs, Required Actions and CTs provided by the licensee for the scope of the RICT program and concluded that, with the incorporation of the RICT program, the required performance levels of equipment specified in LCOs are not changed, only the required CT for the Required Actions are modified, such that 10 CFR 50.36(c)(2) will continue to be met. Based on the discussion provided above, the NRC

staff finds that the RICT program provided in section 2.0 of this SE, LCOs, Required Actions, and CTs meet the first key principle of RG 1.174 and RG 1.177. Although the following TS Actions proposed for a RICT provide for redundancy of function, the NRC staff determined that they are either identified for further evaluation within TSTF-505 or are outside the general scope of TSTF-505: LCO 3.6.2.1, Action a. for one containment spray system inoperable. LCO 3.6.2.2, Action for one train of containment cooling inoperable. LCO 3.7.1.2, Action d. for an inoperable emergency feedwater (EFW) system that remains capable of delivering at least 100 percent flow to either steam generator (SG). LCO 3.7.1.5, Action for one inoperable main steam line isolation valve (MSIV). LCO 3.7.1.7, Action a. for an inoperable atmospheric dump valve (ADV) automatic actuation channel. The above TS Actions are, therefore, evaluated further in section 3.3, Conditions Requiring Additional Technical Justification, of this SE, to ensure that the function will be maintained. Key Principle 2: Evaluation of DID In RG 1.174, the NRC identified the following considerations used for evaluation of how the licensing basis change is maintained for the DID philosophy: Preserve a reasonable balance among the layers of defense. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty. Preserve adequate defense against potential CCFs [common cause failures]. Maintain multiple fission product barriers. Preserve sufficient defense against human errors. Continue to meet the intent of the plants design criteria. The licensee requested to use the RICT program to extend the existing CTs for the respective TS LCOs prescribed in attachment 2 to the LAR, as supplemented. For the TS LCOs, in attachment 2 and enclosure 1 to the LAR, as supplemented, the licensee provided a description and assessment of the redundancy and diversity for the proposed changes. The NRC staffs

evaluation of the proposed changes for these LCOs assessed Waterford 3s redundant or diverse means to mitigate accidents to ensure consistency with the plant licensing basis requirements using the guidance prescribed in RGs 1.174 and 1.177 and TSTF-505, to ensure adequate DID (for each of the functions) to operate the facility in the proposed manner (i.e., that the changes are consistent with the DID criteria). to the LAR, as supplemented, provided information supporting the Waterford 3 evaluation of the redundancy, diversity, and DID for each TS LCO and TS Required Action as it relates to instrumentation and control (I&C), and electrical power systems. The NRC confirmed that the following TS LCOs are consistent with the DID philosophy: TS 3.3.1, Reactor Protective Instrumentation TS 3.3.2, Engineered Safety Features Actuation System Instrumentation TS 3.8.1.1, A.C. [Alternating Current] Sources - Operating TS 3.8.2.1, DC [Direct Current] Sources - Operating TS 3.8.3.1, Onsite Power Distribution Systems - Operating ) For the TS LCOs specific to I&C (TSs 3.3.1 and 3.3.2), the NRC staff reviewed the specific trip logic arrangements, redundancy, backup systems, manual actions, and diverse trips specified for each of the protective safety functions and associated instrumentation as described in the associated Updated Final Safety Analysis Report (UFSAR) (Reference 18) sections, and as reflected in enclosure 1 to the LAR, as supplemented, for each I&C LCO above. The NRC staff verified, that in accordance with the Waterford 3 UFSAR and equipment and actions credited in enclosure 1 to the LAR, as supplemented, in all applicable operating modes, the affected protective feature would perform its intended function by ensuring the ability to detect and mitigate the associated event or accident when the CT of a channel is extended. Furthermore, the NRC staff concludes that there is sufficient I&C redundancy, diversity, and DID to protect against CCFs and potential single failure for the Waterford 3 instrumentation systems evaluated in enclosure 1 to the LAR during a RICT. There is at least one diverse means specified by the licensee for initiating mitigating action for each accident event, thus providing DID against a failure of instrumentation during the RICT for each TS LCO. The DID specified by the licensee does not overly rely on manual actions as the diverse means; therefore, there is not over-reliance of programmatic activities as compensatory measures. Therefore, the NRC staff finds that the intent of the plants design criteria (e.g., safety functions) for the above TS LCOs related to I&C are maintained. For the TS LCOs specific to electrical power systems, the Waterford 3 UFSAR states that the plant is designed such that the safety functions are maintained assuming a single failure within the electrical power system. Single failure requirements are typically suspended for the time that a plant is not meeting an LCO (i.e., in an ACTION statement). The NRC staff reviewed the information the licensee provided in the LAR, as supplemented, for the proposed TS LCOs and TS Bases, and the UFSAR to verify the capability of the affected electrical power systems to perform their safety functions (assuming no additional failures) is maintained. The staff verified that the design success criteria for the affected TS LCO stated in table E1-1, List of Revised Required Actions to Corresponding PRA Functions, of enclosure 1 to the LAR supplement dated May 16, 2022, reflect the redundant or absolute minimum electrical power source/subsystem required to be operable to support the safety functions necessary to mitigate postulated design-basis accidents (DBAs), safely shutdown the reactor, and maintain the reactor in a safe shutdown condition. In addition, the NRC staff reviewed the risk management action (RMA) examples which provide reasonable assurance that the appropriate RMAs will be

implemented to monitor and control risk. The NRC staff finds that the intent of the plants design criteria (e.g., safety functions) applicable to the electrical power related TS LCOs provided above are maintained. The NRC staff notes that while in a TS LCO condition, the redundancy of the function will be temporarily relaxed and, consequently, the system reliability will be degraded accordingly. The NRC staff examined the design information from the Waterford 3 UFSAR, and the risk-informed TS LCO conditions for the affected safety functions. Based on this information, the NRC staff confirmed that under any given DBA evaluated in the Waterford 3 UFSAR, the affected protective features maintain adequate DID. Considering that the CT extensions will be implemented in accordance with the guidance in NEI 06-09-A that also considers RMAs, and the redundancy of the offsite and onsite power system, the NRC staff finds that the plant will maintain adequate DID. Therefore, the NRC staff finds the electrical power system related TS LCOs proposed by the licensee in enclosure 1 to the LAR, as supplemented, are acceptable for the RICT program. The NRC staff reviewed all TS LCOs proposed by the licensee in enclosure 1 to the LAR, as supplemented, and concludes that the proposed changes do not alter the ways in which the Waterford 3 systems fail, do not introduce new CCF modes, and the system independence is maintained. The NRC staff finds that some proposed changes reduce the level of redundancy of the affected systems, and this reduction may reduce the level of defense against some CCFs; however, such reductions in redundancy and defense against CCFs are acceptable due to existing diverse means available to maintain adequate DID against a potential single failure during a RICT. The NRC staff finds that extending the selected CTs with the RICT program following loss of redundancy, but maintaining the capability of the system to perform its safety function, is an acceptable reduction in DID during the proposed RICT period provided that the licensee identifies and implements compensatory measures in accordance with the RICT program during the extended CT. Based on the above, the NRC staff finds that the licensees proposed changes are consistent with the NRC-endorsed guidance prescribed in NEI 06-09-A and satisfy the second key principle in RGs 1.174 and 1.177. Additionally, the NRC staff concludes that the changes are consistent with the DID philosophy as described in RG 1.174. Key Principle 3: Evaluation of Safety Margins Paragraph 50.55a(h) of 10 CFR requires, in part, that protection systems of nuclear power reactors of all types must meet the requirements specified in this paragraph. Section 2.2.2, Technical Specification Change Maintains Sufficient Safety Margin (Principle 3), of RG 1.177 states, in part, that sufficient safety margins are maintained when: Codes and standards or alternatives approved for use by the NRC are met. Safety analysis acceptance criteria in the final safety analysis report are met or proposed revisions provide sufficient margin to account for analysis and data uncertainties. The licensee is not proposing in its LAR to change any quality standard, material, or operating specification. In the LAR, the licensee proposed to add a new program, the RICT Program, in

section 5.0, Administrative Controls, of the Waterford 3 TSs, which would require adherence to NEI 06-09-A. NEI 06-09-A, Condition 2 in part, stipulates for the TS LCOs and action requirements to which the RMTS will apply, the LAR will provide justification with comparison of the TS functions to the PRA modeled functions of the structures, systems, or components (SSCs) subject to those LCO actions or an appropriate disposition or programmatic restriction will be provided. The NRC staff evaluated the effect on safety margins when the RICT program is applied to extend the CT up to a backstop of 30 days in a TS condition with sufficient trains remaining operable to fulfill the TS safety function. Although the licensee will be able to have design-basis equipment out of service longer than the current TS allows, any increase in unavailability is expected to be insignificant and is addressed by the consideration of the single failure criterion in the design-basis analyses. Acceptance criteria for operability of equipment are not changed and, if sufficient trains remain operable to fulfill the TS safety function, the operability of the remaining train(s) ensures that the current safety margins are maintained. The NRC staff finds that if the specified TS safety function remains operable, sufficient safety margins would be maintained during the extended CT of the RICT program. Safety margins are also maintained if PRA functionality is determined for the inoperable train, which would result in an increased CT. Credit for PRA functionality, as described in NEI 06-09-A, is limited to the inoperable train, loss-of-offsite power (LOOP), or component. Based on the above, the NRC staff finds that the design basis analyses for Waterford 3 remain applicable and unchanged, that sufficient safety margins would be maintained during the extended CT, and that the proposed changes to the TSs do not include any change in the standards applied or the safety analysis acceptance criteria. The NRC staff concludes that the proposed changes meet 10 CFR 50.55a(h) and, therefore, Key Principle 3. Key Principle 4: Change in Risk Consistent with the Safety Goal Policy Statement NEI 06-09-A provides a methodology for a licensee to evaluate and manage the risk impact of extensions to TS CTs. Permanent changes to the fixed TS CTs are typically evaluated by using the three-tiered approach described in SRP section 16.1 and RG 1.177. This approach addresses the calculated change in risk as measured by the change in core damage frequency (CDF) and large early release frequency (LERF), as well as the incremental conditional core damage probability (ICCDP) and incremental conditional large early release probability (ICLERP); the use of compensatory measures to reduce risk; and the implementation of a configuration risk management program (CRMP) to identify risk-significant plant configurations. The NRC staff evaluated the licensees processes and methodologies for determining that the change in risk from implementation of RICTs will be small and consistent with the intent of the Commissions Safety Goal Policy Statement. In addition, the NRC staff evaluated the licensees proposed changes against the three-tiered approach in RG 1.177 for the licensees evaluation of the risk associated with a proposed TS CT change. The results of the NRC staffs review are discussed below. Tier 1: PRA Capability and Insights Tier 1 evaluates the impact of the proposed changes on plant operational risk. The Tier 1 review involves two aspects: (1) scope and acceptability of the PRA models and their application to the

proposed changes and (2) a review of the PRA results and insights described in the licensees application. In enclosures 2, Information Supporting Consistency with Regulatory Guide 1.200, Revision 2, and enclosure 4, Information Supporting Justification of Excluding Sources of Risk Not Addressed by the PRA Models, to the LAR and enclosure 1, Response to Request for Additional Information, to the LAR supplement dated May 16, 2022, the licensee identified the following modeled hazards and alternate methodologies that the licensee used to assess the risk contribution for extending the CT of a TS LCO in the proposed Waterford 3 RICT program. Internal Events PRA (IEPRA) model (includes internal floods) Internal Fire PRA (FPRA) model Seismic Hazard: a CDF penalty of 4.24E-06 per year, and a LERF penalty of 1.94E-06 per year Other External Hazards: screened out from RICT program based on appendix 6-A of the American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard ASME/ANS RA-Sa-2009 Addenda to ASME RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications (ASME/ANS RA-SB-2005 PRA standard) (Reference 19) Evaluation of Modeled PRAs In attachment 4, Revisions to Information Supporting Consistency with Regulatory Guide 1.200, Revision 2, to enclosure 1 to the LAR supplement dated May 16, 2022, the licensee confirmed that the PRA models have been peer reviewed. The internal events PRA was peer reviewed using the ASME/ANS RA-Sb-2005 PRA standard as endorsed by RG 1.200, Revision 2 Further the IEPRA and FPRA were peer reviewed using the ASME/ANS RA-Sa-2009 PRA standard. This included for the IEPRA focused-scope peer reviews on internal flooding, LERF, and human reliability analysis. The licensee stated that it conducted an independent assessment process for closure of the facts and observations (F&Os) resulting from these peer reviews. The NRC staff confirmed that the licensee performed closure of the F&Os consistent with Appendix X to NEI 05-04, 07-12, and 12-13, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (Reference 20), as endorsed in RG 1.200, Revision 3. The NRC evaluated the remaining open F&Os, along with their dispositions. In the LAR supplement dated August 19, 2022, the licensee proposed a license condition to resolve the open human reliability F&Os using the NRC accepted process, which is described in section 3.0 of this SE. The portions of RG 1.200, Revisions 2 and 3 discussed above do not include any technical changes that would impact the consistency with NEI 06-09-A; therefore, the NRC staff finds these portions of the RG also applicable for use in the licensees adoption of TSTF-505. During the approval process for the Waterford 3 LAR to adopt the 10 CFR 50.69 categorization process, the NRC staff became aware that the Waterford 3 FPRA had not yet incorporated the updated ignition frequencies provided in NUREG-2169, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009 (Reference 21). In response to NRC PRA Licensing Branch A (APLA) RAI 08.a and APLA RAI 09, the licensee performed sensitivity studies to assess the impact of the fire ignition frequencies on the RICT program. The

results of the sensitivity studies are found in updated table E1-2, In Scope TS/LCO Conditions RICT Estimate, in attachment 2 of enclosure 1 to the LAR supplement dated May 16, 2022, which demonstrate a significant impact on several RICT estimates. Therefore, the licensee provided, in enclosure 5, List of Regulatory Commitments, to the LAR supplement dated May 16, 2022, a commitment to incorporate the NUREG-2169 ignition frequencies prior to RICT program implementation. During the approval process for the Waterford 3 LAR to adopt the 10 CFR 50.69 categorization process, the NRC staff became aware that the Waterford 3 PRA models incorporate flexible and diverse coping (FLEX) strategies and equipment. The NRC staff noted in the NRC memorandum, Assessment of the NEI 16-06, 'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis, dated May 30, 2017 (Reference 22), certain uncertainties related to FLEX modeling associated with failure rates of portable equipment, and that these uncertainties should be considered in the PRA models to stay consistent with the ASME/ANS RA-Sa-2009 PRA standard, . In response to APLA RAI 02 in the LAR supplement dated May 16, 2022, the licensee stated that sensitivity studies were performed on the proposed RICT TS LCOs with no FLEX credit that demonstrated minimal impact on RICT values. The licensee considered PRA modeling uncertainties and their potential impact on the RICT program and identified, as necessary, the applicable RMAs to limit the impact of these uncertainties. In response to APLA RAI 08.b and c in the LAR supplement dated May 16, 2022, the licensee discussed the identification of key assumptions and sources of uncertainty along with providing the dispositions for impact on the risk-informed application or applicable sensitivities. In enclosure 2 to this supplement the licensee evaluated the Waterford 3 PRA models to identify the key assumptions and sources of uncertainty for this application, consistent with the RG 1.200, Revision 2, definitions, and using sensitivity and importance analyses to place bounds on uncertain processes, identify alternate modeling strategies, and provide information to users of the PRA. The NRC staff finds that the licensee performed an adequate assessment to identify the potential sources of uncertainty, and that the identification of the key assumptions and sources of uncertainty was appropriate and consistent with the guidance in NUREG-1855 and associated Electric Power Research Institute (EPRI) TRs 1016737, Treatment of Parameter and Model Uncertainity for Probabilistic Risk Assessments, December 2008 (Reference 23) and 1026511, Practical Guidance of the Use of Probabilistic Risk Assessment in Risk-informed Applications with a Focus on the Treatment of Uncertainty, December 2012 (Reference 24). Therefore, the NRC staff finds that the licensee has satisfied the guidance in RGs 1.174 and 1.177 and that the identification of assumptions and treatment of model uncertainties for risk evaluation of extended CTs is appropriate for this application and is consistent with the guidance in NEI 06-09-A. The NRC staff reviewed the peer review history of the PRA models provided by the licensee in attachment 4 of enclosure 1 to the LAR, as supplemented. The licensee adequately applied the guidance for establishing PRA technical acceptability for the models. The NRC staff further considered the key assumptions and sources of uncertainty identified by the licensee, the proposed use of surrogates in the PRA models for evaluating RICTs for specific TS functions, and credit for FLEX. Therefore, the NRC staff finds that the scope and technical acceptability of the IEPRA and FPRA are commensurate with the RICT application for use in the integrated decision-making process and consistent with RG 1.174.

Evaluation of Seismic Hazard The licensees proposed approach for including the seismic risk contribution in the RICT calculation is to add a seismic CDF penalty and a seismic LERF penalty to each RICT calculation. The proposed seismic CDF penalty is based on using the plant-specific seismic hazard curves developed in response to the Near-Term Task Force Recommendation 2.1 as noted in the letter from Entergy dated March 27, 2014 (Reference 25), a plant-level high confidence of low probability of failure (HCLPF) capacity of 0.15g referenced to peak ground acceleration (PGA), and a composite beta factor of 0.4 to represent the uncertainty parameter for seismic capacity. Using these inputs, the licensee calculated a seismic CDF penalty of 4.24E-06 per year. The NRC staff finds the proposed method to determine the seismic CDF acceptable because it is consistent with the approach used in NRC Generic Issue (GI)-199 , Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, dated September 2, 2010 (Reference 26). The NRC staff performed an independent convolution using the input parameters identified by the licensee to confirm the proposed seismic CDF penalty. The proposed seismic LERF penalty is based on a convolution of the estimated seismic CDF, as described above, with a limiting fragility for containment isolation failure, which was also assumed to be an HCLPF of 0.15g referenced to PGA. Using these inputs, the licensee calculated a seismic LERF penalty of 1.94E-06 per year. The NRC staff finds the proposed method to determine the seismic LERF penalty acceptable because the use of a 0.15g PGA HCLPF as the fragility for containment isolation is conservative. The licensee did not address the incremental risk associated with seismic-induced LOOP in the LAR, as supplemented. The NRC staff used a typical LOOP fragility and site-specific seismic hazard curve to estimate the seismic-induced LOOP frequency at a level of approximately 1E-5 per year. This frequency is approximately 1 percent of the total internal events 24-hour non-recovered LOOP frequency at a level of 1E-3 per year already addressed in the internal events PRA model. Therefore, the NRC staff finds that the exclusion of seismic-induced LOOP frequency from the non-recovered LOOP frequency in the internal events PRA model has an insignificant impact on the RICT program calculations. The NRC staff finds that, during RICTs for SSCs credited in the design-basis to mitigate seismic events, the licensees proposed methodology captures the risk associated with seismically induced failures of redundant SSCs, because such SSCs are assumed to be fully correlated. In summary, the NRC staff finds the licensees proposal to use the seismic CDF contribution of 4.24E-06 per year and a seismic LERF contribution of 1.94E-06 per year acceptable for the licensees RICT program for Waterford 3 because (1) the licensee used the most current site-specific seismic hazard information for Waterford 3, (2) the licensee provided justification to use a plant-level HCLPF value of 0.15g, which is higher than the value of 0.1g used in the GI-199 evaluation, and kept the same composite beta factor of 0.4 used in the GI-199 evaluation, (3) the licensee determined a seismic LERF penalty based on its estimate of seismic CDF combined with a containment isolation fragility of 0.15g PGA HCLPF, and (4) the licensee will add the baseline seismic risk to RICT calculations with an assumption of fully correlated failures, which is conservative for SSCs credited in seismic events, while any potential for non-conservative results for SSCs that are not credited in seismic events is small or nonexistent.

Evaluation of Other External Hazards The licensee evaluated external hazards in table E4-1, External Hazard Evaluation, in enclosure 4 to the LAR. This table was subsequently revised in the LAR supplement dated May 16, 2022. For the external flooding hazard, the licensee concluded that this hazard has an insignificant contribution to risk based on updated plant data, flood history, and new measures for risk management as well as the focused evaluation for external flooding using Waterford Steam Electric Station, Unit 3 - Flood Focused Evaluation Assessment (CAC No. MF9710; EPID L-2017-JLD-0009), (Reference 27). The NRC staff reviewed the licensees considerations of external flooding hazards for Waterford 3 and finds that the external flooding hazard has an insignificant contribution to configuration risk and can be excluded from the calculation of the proposed RICTs, which is consistent with the conclusion of the NRC staffs review of the licensees flood focused evaluation assessment. For the extreme winds and tornados hazard, the licensee stated that this hazard has an insignificant contribution to risk based on the plant being designed for extreme winds and tornado loadings that are substantially higher than those required from the plants licensing basis and the thickness of tornado missile barriers protecting safety-related SSCs. In addition, the licensee stated that extreme winds and tornados are considered for the initiating events analysis for the PRA model for a LOOP. The NRC staff reviewed the licensees evaluation of the risk from extreme winds and tornados and finds that it is acceptable for the RICT program because the impacts from the extreme winds and tornado are considered in the PRA model as a LOOP event. For the other external hazards evaluated in table E4-1 in enclosure 4 to the LAR, the licensee concluded that these external hazards have an insignificant contribution to risk and proposed that these external hazards be screened out from the RICT program. The other external hazards evaluated in table E4-1 are those identified for consideration in non-mandatory appendix 6-A of the ASME/ANS RA-Sa-2009 PRA standard , which provides a guide for identification of most of the possible external events for a plant site. The NRC staff reviewed the licensees evaluation of other external hazards and finds that the licensees consideration of risk from other external hazards is acceptable because the screening criteria used in table E4-1 are the same as the criteria presented in supporting requirements for screening external hazards EXT-B1 and EXT-C1 of the ASME/ANS RA-Sa-2009 PRA standard. In summary, the NRC staff finds that the contributions from external flooding, extreme winds and tornados, and other external hazards have an insignificant contribution to configuration risk and can be excluded from the calculation of the proposed RICTs because they either do not challenge the plant or they are bounded by the external hazards analyzed for the plant. Application of PRA Models, Results, and Insights in the RICT Program At the time of the submittal of the LAR and the LAR supplement dated April 8, 2021, the licensee had not fully completed the development of its CRMP tool (hereinafter referred to as the real-time risk (RTR) model). In response to NRC staff RAI 03, the licensee described the activities planned for completing the RTR model. The licensee described the planned verification and validation and benchmarking activities, which include comparison with prior models, and results from the maintenance rule models. The licensee also described its process for the RTR model update. The licensee tracks any changes to the PRA or the plant (including engineering changes, procedure revisions, licensing revisions, model improvement, plant-specific data changes, and industry research) and assesses their impact on the model. The

licensee described its criteria for performing planned updates to the RTR model and interim updates to the RTR model when significant model changes occur that would impact the RICT calculations. In its response to NRC staff RAI 07, the licensee described its process for deriving RMAs, which includes predetermined RMAs for procedures, RTR model insights, and operator experience, as well examining the list of important trains provided by the RTR model. Based on the above, the NRC staff finds that the Waterford 3 PRA models and RTR model used will continue to reflect the as-built, as-operated plant consistent with RG 1.200, Revision 2 for ensuring PRA acceptability is maintained. Therefore, the NRC staff concludes that the proposed application of the Waterford 3 RICT program is appropriate for use in the adoption of TSTF 505 for performing RICT calculations. In enclosure 5, Baseline CDF and LERF, to the LAR, as supplemented, the licensee provided the estimated total CDF and LERF of the base PRA models to demonstrate that Waterford 3 meets the 1E-4/year CDF and 1E-5/year LERF criteria of RG 1.174 consistent with the guidance in NEI 06-09-A and that these guidelines will be satisfied for implementation of a RICT program. The licensee has incorporated NEI 06-09-A into TS 6.5.19. The estimated current total CDF and LERF for Waterford 3 PRAs meet the RG 1.174 guidelines; therefore, the NRC staff concludes that the PRA results and insights to be used by the licensee in the RICT program will continue to be consistent with NEI 06-09-A. Tier 1: Conclusions Based on the above, the NRC staff finds that the licensee has satisfied the intent of Tier 1 for determining the PRA acceptable, and that the scope of the PRA models (i.e., IEPRA and FPRA) and other evaluated hazards, external hazards, and seismic methodology is appropriate for this application. Tier 2: Avoidance of Risk-Significant Plant Configurations Tier 2 evaluates the capability of the licensee to identify and avoid risk-significant plant configurations that could result if equipment, in addition to that associated with the proposed change, is taken out of service simultaneously, or if other risk-significant operational factors, such as concurrent system or equipment testing, are also involved. The limits established for entry into a RICT program and for RMA implementation are consistent with the guidance of Nuclear Utility Management and Resources Council (NUMARC) 93-01, Revision 4F, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, dated April 2018 (Reference 28) endorsed by RG 1.160, Revision 4, Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, dated August 2018 (Reference 29), as applicable to plant maintenance activities. The LAR also explains that RMAs will be implemented, in accordance with current plant procedures, no later than the time at which the 1E-06 ICCDP or 1E-07 ICLERP threshold is reached and under emergent conditions when the instantaneous CDF and LERF thresholds are exceeded. The NRC staff concludes that the RICT program requirements, that includes limits established for entry into a RICT, and implementation of RMAs are consistent with NEI 06-09-A. Therefore, the NRC staff finds that the proposed changes are consistent with the intent of Tier 2.

Tier 3: Risk-Informed Configuration Risk Management Tier 3 stipulates that a licensee should develop a program that ensures that the risk impact of out-of-service equipment is appropriately evaluated prior to performing any maintenance activity. The proposed RICT program establishes a CRMP based on the underlying PRA models. The CRMP is then used to evaluate configuration-specific risk for planned activities associated with the RMTS extended CT, as well as emergent conditions that may arise during an extended CT. This required assessment of configuration risk, along with the implementation of compensatory measures and RMAs, is consistent with the principle of Tier 3 for assessing and managing the risk impact of out of service equipment. Paragraph 50.36(c)(5) of 10 CFR identifies administrative controls as 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. In enclosure 8, Attributes of the CRMP Model, to the LAR, the licensee confirmed that future changes made to the baseline PRA models and changes made to the online model (i.e., CRMP) are controlled and documented by plant procedures. In enclosure 10, Program Implementation, to the LAR, the licensee identified the attributes that the RICT program procedures will address, which are consistent with NEI 06-09-A. In response to APLA RAI 03.b in the LAR supplement dated May 16, 2022, the licensee committed to update licensee procedure EN-DC-151, PRA Maintenance and Update, to include criteria to perform an interim update to the PRA models used in the RTR model specific to the RICT application. The NRC staff finds that the licensee has identified appropriate administrative controls consistent with NEI 06-09-A and that it will continue to meet 10 CFR 50.36(c)(5). Based on the licensees incorporation of NEI 06-09-A in the TSs, as discussed in LAR attachment 1; use of RMAs as discussed in LAR enclosure 12, Risk Management Action Examples; and because the proposed changes are consistent with the Tier 3 guidance, the NRC staff finds the licensees Tier 3 program is acceptable and supports the proposed implementation of the RICT program. Key Principle 4: Conclusions The licensee has demonstrated the technical acceptability and scope of its PRA models and alternative methods, including consideration of the impact of seismic events, non-seismic external hazards, and other hazards, and that the models can support implementation of the RICT program for determining extensions to CTs. The licensee has made proper consideration of key assumptions and sources of uncertainty. The risk metrics are consistent with the approved methodology of NEI 06-09-A and the acceptance guidance in RGs 1.174 and 1.177. The RICT program will be controlled administratively through plant procedures and training and follows the NRC-approved methodology in NEI 06-09-A. Based on the above, the NRC staff concludes that the RICT program satisfies the fourth key principle and is, therefore, acceptable. Key Principle 5: Performance Measurement Strategies - Implementation and Monitoring RGs 1.174 and 1.17 establish the need for an implementation and monitoring program to ensure that extensions to TS CTs do not degrade operational safety over time and that no adverse degradation occurs due to unanticipated degradation or common cause mechanisms. 1, Monitoring Program, to the LAR states that the SSCs in scope of the RICT program are also in scope of 10 CFR 50.65 for the Maintenance Rule. The Maintenance Rule monitoring programs will provide for evaluation and disposition of unavailability impacts, which

may be incurred from implementation of the RICT program. Furthermore, in enclosure 11 to the LAR, the licensee confirmed that the cumulative risk is calculated at least every refueling cycle, but the recalculation period does not exceed 24 months, which is consistent with NEI 06-09-A. The NRC staff concludes that the RICT program satisfies the fifth key principle because: (1) the RICT program will monitor the average annual cumulative risk increase as described in NEI 06-09-A, thereby ensuring that the program, as implemented, continues to meet guidance in RGs 1.174 and 1.177 guidance for small risk increases; and (2) all affected SSCs are within the Maintenance Rule program, which is used to monitor changes to the reliability and availability of those SSCs. 3.2 Optional Changes and Variations from TSTF-505 The NRC staff evaluated the proposed use of RICTs in the optional changes and variations discussed above in section 2.0 in conjunction with evaluating the proposed use of RICTs in each of the individual LCOs, Required Actions, and CTs. The NRC staffs evaluation of the licensees proposed use of RICTs in the variations against the key safety principles is discussed above. Based on the above section, the NRC staff finds that each of the five key principles have been met and concludes that the proposed optional changes and variations are acceptable 3.3 Conditions Requiring Additional Technical Justification 3.3.1 Additional Technical Justification Specified in TSTF-505 Table 1 of TSTF-505 contains a list of LCO conditions that may be proposed for inclusion in the RICT program where additional information may be necessary to explain why the condition would not represent a loss of specified safety function as used in the RICT program. Suggestions are provided in the table, but the suggestions may not be all encompassing for all plants. Licensees should provide sufficient information when adopting the listed Required Actions to justify that the condition does not represent a loss of specified safety function as used in the RICT program. The following information was obtained from the LAR and UFSAR, related to the containment and plant systems TSs where TSTF-505 specified the need for additional technical justification: STS 3.6.6A LCO Containment Spray and Cooling Systems Atmospheric and Dual) As indicated in table E1-1 of enclosure 1 to the LAR, the containment spray and cooling systems are treated separately under TS LCOs 3.6.2.1, Containment Spray System, and 3.6.2.2, Containment Cooling System. The conditions proposed for risk-informed allowed outage times involve inoperability of only one of two trains for each system, and the two conditions may be entered simultaneously. This consideration is consistent with the design-basis described in Section 6.2.2, Containment Heat Removal Systems, of the Waterford Unit 3 UFSAR, which assumes the operation of one spray and one cooling train providing containment cooling for accident mitigation. Section 2.3, Optional Changes and Variations, of attachment 1 to the LAR provides the following information addressing TS LCO 3.6.2.1 and TS LCO 3.6.2.2: NUREG 1432 includes a single technical specification for Containment Spray and Cooling Systems. Waterford 3 has individual TS for each system. The

cooling function of the systems is redundant. However, iodine removal is a function of the Containment Spray system but not the Containment Cooling System. The Waterford 3 RICT program will include a completion time for a single Containment Spray train inoperable. The program will also include a completion time for one of the two trains of Containment Cooling being inoperable. Having both Containment Spray trains inoperable is a loss of iodine removal function and will not be included in the RICT program. Thus, the operability of one spray train ensures that the iodine removal function would be maintained. These systems are explicitly modeled in the PRA, and the PRA success criteria are consistent with the design-basis. Therefore, the proposed actions would not present a loss of function condition for the containment cooling or iodine removal functions. STS 3.7.2.A, LCO [Two] MSIVs shall be OPERABLE One MSIV inoperable in MODE 1 As indicated in table E1-1 of enclosure 1 to the LAR, the MSIVs are explicitly modeled in the PRA. The PRA success criteria are consistent with the design-basis. The Waterford 3 MSIVs are double-disk gate valves that prevent flow in either direction. Thus, closure of one valve, as required for DBAs involving a loss of main steam line pressure boundary integrity, would result in isolation of one SG from the break, consistent with the design-basis. Therefore, the proposed action would not present a loss of function condition for the SG isolation function. 3.3.2 Plant-Specific LCO Variations from TSTF-505 Two plant-specific LCOs and associated Actions for which the licensee is proposing to apply the RICT program that are variations from the STSs in NUREG-1432 and the evaluated actions in TSTF-505 were identified in enclosure 13, Waterford 3 to Standard Technical Specification Cross Reference, to the LAR. The NRC staff found the following TS LCO Actions to differ significantly from those considered for inclusion in TSTF-505: LCO 3.7.1.2, Action d for an inoperable EFW system that remains capable of delivering at least 100 percent flow to either SG. LCO 3.7.1.7, Action a for an inoperable ADV automatic actuation channel. The NRC staff evaluated the information provided for these plant-specific LCOs and associated Actions to confirm that the condition does not represent a TS loss of function. Acceptability is based on appropriate constraints to preclude application of a RICT when a loss of function condition exists or if the system functions are not appropriately modeled in the PRA to reflect the risk associated with a loss of redundancy. For conditions under LCO 3.7.1.2, Action d, which applies for an inoperable EFW system that remains capable of delivering at least 100 percent flow to either SG, the safety function is maintained because, as specified in the Action, adequate flow can be delivered to either SG for decay heat removal and accident mitigation functions. TSTF-505 includes consideration of a RICT for STS LCO 3.7.5.B, which applies when one auxiliary feedwater train is inoperable for reasons other than a turbine-driven pump inoperable as a result of one inoperable steam supply. This condition is similar to the proposed conditions under TS 3.7.1.2, Action d, which applies when the EFW system can deliver 100 percent flow to each SG. When 100 percent flow can be delivered to each SG with operable EFW equipment, equipment equivalent to at least one train of EFW is operable. In addition, the bases for TS LCO 3.7.1.2, Action d, describes

that the action applies to conditions including an inoperable turbine-driven pump, both motor-driven pumps inoperable, or one of two redundant flow paths. Therefore, these conditions are equivalent to the STS LCO 3.7.5.B accepted for a RICT in TSTF-505. In the LAR supplement dated May 16, 2022, the licensee provided estimated RICTs for an inoperable turbine-driven pump and two inoperable motor-driven pumps in table E1-2, which supports evaluation of the PRA model of the system. Therefore, there is no loss of function condition associated with the condition, and a RICT is appropriate. LCO 3.7.1.7 Action a applies when one ADV automatic actuation channel is inoperable. The accident analyses credit automatic ADV actuation at power levels above 70 percent of full power to mitigate certain DBA conditions, and the success criterion is that one of two ADVs actuate to remove heat from the reactor coolant system through the SGs. The action ensures that one ADV actuation channel is operable and, therefore, represents only a loss of redundancy. In the LAR supplement dated May 16, 2022, table E1-1 indicates that the PRA model includes automatic actuation of the valves and the PRA success criteria are consistent with the design-basis. Therefore, there is no loss of function condition associated with the condition, and a RICT is appropriate. 3.4 Changes to the Operating License In the LAR and in the licensees responses to NRC staff RAIs there were certain specific actions that the NRC staff identified as being necessary to support the conclusion that the implementation of the proposed RICT program met the requirements of the RICT. The NRC staff finding on the acceptability of the implementation of the RICT program for the TS LCOs in this SE is dependent on the completion of the following license condition. In the LAR supplement dated August 19, 2022, the licensee proposed the following changes to the Waterford 3 renewed facility operating license: Entergy will complete closure of the four Human Reliability Analysis (HRA) Finding level Facts and Observations (F&Os) identified as Finding Numbers HR 1-2, HR 7-1, HR 7-3, and HR 7-4 in Table A3-2 of Entergy letter to NRC, dated April 25, 2022, and in Table E2-2 of Entergy letter to NRC, dated May 16, 2022, using an accepted NRC process (Nuclear Energy Institute (NEI) Appendix X to NEI 05-04, NEI 07-12, and NEI 12-13) prior to implementation of 10 CFR 50.69 and the risk-informed completion time (RICT) program. The NRC staff notes that prior approval would be required for a change to the RICT program, or the implementation of the RICT program as described in TS 6.5.19, and the implementation item in the LAR supplement dated May 16, 2022. The NRC staff finds that the above license condition is acceptable because it adequately implements the RICT program using models, methods, and approaches consistent with applicable guidance that are acceptable to the NRC. The NRC staff, through an onsite audit or during future inspections, may choose to examine the closure of the license condition, with the expectation that any issues discovered during this review, or concerns with regard to its adequate completion, would be tracked and dispositioned appropriately under the licensees corrective action program and could be subject to appropriate NRC enforcement action.

3.5 Technical Evaluation Conclusion

The NRC staff has evaluated the proposed changes against each of the five key principles in RGs 1.174 and RG 1.177, including the optional variations from the approved TSTF 505. The NRC staff concludes that the proposed changes satisfy the key principles of risk-informed decision-making identified in RG 1.174 and RG 1.177 and, therefore, the requested adoption of the proposed changes to the TSs, implementation items, and associated guidance, is acceptable to the NRC staff to ensure that the Commissions regulations continue to be met.

4.0 STATE CONSULTATION

In accordance with the Commissions regulations, the Louisiana State official was notified of the proposed issuance of the amendment on January 25, 2022. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding which was published in the Federal Register on May 18, 2021 (86 FR 26954), that the amendment involves no significant hazards consideration, and there has been no public comment on such finding. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions 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.

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20. Anderson, V. K., Nuclear Energy Institute, letter to S. Rosenberg, U.S. Nuclear Regulatory Commission, Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations, dated February 21, 2017 (ML17086A431).

21. U.S. Nuclear Regulatory Commission and Electric Power Research Insitute, Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009, NUREG-2169 and EPRI 3002002936, dated January 2015 (ML15016A069).

22. Reisi-Fard, M., U.S. Nuclear Regulatory Commission, memorandum to J. G. GiItter, U.S.

Nuclear Regulatory Commission, Assessment of the Nuclear Energy Institute 16-06,

   'Crediting Mitigating Strategies in Risk-Informed Decision Making,' Guidance for Risk-Informed Changes to Plants Licensing Basis, dated May 30, 2017 (ML17031A269).

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24. Electirc Power Research Insitute, Practical Guidance of the Use of Probabilistic Risk Assessment in Risk-informed Applications with a Focus on the Treatment of Uncertainty, EPRI TR-1026511, December 2012.

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26. Hiland, P., U.S. Nuclear Regulatory Commission, memorandum to B. W., Sheron, U.S.

Nuclear Regulatory Commission, Safety/Risk Assessment Results for Generic Issue 199, 'Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants', dated September 2, 2010 (ML100270582).

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Waterford Steam Electric Station, Unit 3 - Flood Focused Evaluation Assessment (CAC No. MF9710; EPID L-2017-JLD-0009), dated February 26, 2018 (ML17171A128).

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Principal Contributors: J. Evans M. Biro D. Wu S. Alferink E. Kleeh V. Goel M. Li J. Wilson S. Jones Date: February 17, 2023

ML22322A109 *by email OFFICE NRR/DORL/LPL4/PM* NRR/DORL/LPL4/LA* NRR/DRA/APLA/BC* NRR/DRA/APLC/BC* NAME JDrake PBlechman RPascarelli SVasavada DATE 11/17/2022 12/7/2022 02/08/2023 02/02/2023 OFFICE NRR/DEX/EICB/BC* NRR/DEX/EEEB/BC* NRR/DSS/STSB/BC* NRR/DSS/SCPB/BC* NAME MWaters WMorton VCusumano BWittick DATE 02/15/2023 02/16/2023 02/07/2023 02/02/2023 OFFICE OGC* NRR/DORL/LPL4/BC* NRR/DORL/LPL4/PM* NAME ELicon JDixon-Herrity JDrake DATE 02/14/2023 02/17/2023 02/17/2023}}

v t e Letter Sequence Approval
EPID:L-2021-LLA-0014, TSTF-505 (Approved, Closed)
Initiation Acceptance, Acceptance Supplement Administration Questions 13 April 2022 Answers Results Approval, Approval Other: ML23059A259
MONTHYEAR2021-04-08 [Table View]

ML22322A109

Text

SUBJECT:

Dear Sir or Madam:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

3.5 Technical Evaluation Conclusion

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

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