ML25097A184

Supplement to the Regulatory Audit Plan in Support of the Review of the LAR to Increase Flexibility in Mode Restraints and the LAR for Shutdown Actions Mode Change Restrictions

ML25097A184
Person / Time
Site:	Vogtle
Issue date:	04/16/2025
From:	John Lamb NRC/NRR/DORL/LPL2-1
To:	Coleman J Southern Nuclear Operating Co
References
EPID L-2024-LLA-0171, EPID L-2025-LLA-0011
Download: ML25097A184 (18)

Text

April 16, 2025 Jamie M. Coleman Regulatory Affairs Director Southern Nuclear Operating Co., Inc.

3535 Colonnade Parkway Birmingham, AL 35243

SUBJECT:

VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 - SUPPLEMENT TO THE REGULATORY AUDIT PLAN IN SUPPORT OF THE REVIEW OF THE LICENSE AMENDMENT REQUEST TO INCREASE FLEXIBILITY IN MODE RESTRAINTS AND THE LICENSE AMENDMENT REQUEST FOR SHUTDOWN ACTIONS MODE CHANGE RESTRICTIONS (EPID NOS. L-2024-LLA-0171 AND L-2025-LLA-0011)

Dear Ms. Coleman:

By letter dated December 19, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24354A169), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for Vogtle Electric Generating Plant (Vogtle), Units 3 and 4. The proposed LAR (L-2024-LLA-0171) would change the Technical Specifications (TSs), to increase flexibility in mode restraints. The licensees submittal indicates that the proposed changes are similar to those described in Technical Specification Task Force (TSTF) 359, Revision 9, Increased Flexibility in Mode Restraints, with enhancements similar to those identified in TSTF-529, Revision 4, Clarify Use and Application Rules, for Limiting Condition for Operation (LCO) 3.0.4.

By letter dated January 23, 2025 (ML25023A275), SNC submitted an LAR for Vogtle, Units 3 and 4. The licensees submittal stated that the proposed LAR (L-2025-LLA-0011) requests to modify TSs to revise unnecessarily restrictive and potentially conflicting Required Actions that may also restrict mode changes while shutdown.

By letter dated March 26, 2025 (ML25077A233), the U. S. Nuclear Regulatory Commission (NRC) staff identified the need, for both of these reviews, to conduct a regulatory audit and provided an audit plan to examine SNCs non-docketed information with the intent to gain understanding, verify information, or identify information that will require docketing to support the basis of the licensing or regulatory decision. In the interest of efficiency and effectiveness, the NRC staff is combining its audit efforts into a single activity for both reviews.

The NRC staff added questions 15 through 34 in the attached supplement and will conduct the audit virtually via Microsoft Teams using a licensee-established electronic portal available to NRC staff from approximately April 22 through August 15, 2025, with formal audit meetings to be scheduled during this period as needed. The NRC staff reserves the right to extend the audit, if necessary. The detailed audit plan is enclosed with this letter.

J. Coleman If you have any questions, please contact me at (301) 415-3100 or by email at John.Lamb@nrc.gov.

Sincerely,

/RA/

John G. Lamb, Senior Project Manager Plant Licensing Branch II-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos.52-025 and 52-026

Enclosure:

Audit Plan cc: Listserv

Enclosure SUPPLEMENT TO THE REGULATORY AUDIT PLAN BY THE OFFICE OF NUCLEAR REACTOR REGULATION TO SUPPORT THE REVIEW OF THE LICENSE AMENDMENT REQUEST TO INCREASE FLEXIBILITY IN MODE RESTRAINTS AND LICENSE AMENDMENT REQUEST FOR SHUTDOWN ACTIONS MODE CHANGE RESTRICTIONS SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 3 AND 4 DOCKET NOS.52-025, AND 52-026

1.0 BACKGROUND

By letter dated December 19, 2024 (Agencywide Documents and Access Management System (ADAMS) Accession No. ML24354A169), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) for Vogtle Electric Generating Plant (Vogtle), Units 3 and 4. The proposed LAR (L-2024-LLA-0171) would change the Technical Specifications (TSs), to increase flexibility in mode restraints. The licensees submittal indicates that the proposed changes are similar to those described in Technical Specification Task Force (TSTF) 359, Revision 9, Increased Flexibility in Mode Restraints, with enhancements similar to those identified in TSTF-529, Revision 4, Clarify Use and Application Rules, for Limiting Condition for Operation (LCO) 3.0.4.

By letter dated January 23, 2025 (ML25023A275), SNC submitted an LAR for Vogtle, Units 3 and 4. The licensees submittal stated that the proposed LAR (L-2025-LLA-0011) requests to modify TSs to revise unnecessarily restrictive and potentially conflicting Required Actions that may also restrict mode changes while shutdown.

By letter dated March 26, 2025 (ML25077A233), the U. S. Nuclear Regulatory Commission (NRC) staff identified the need, for both of these reviews, to conduct a regulatory audit and provided an audit plan to examine SNCs non-docketed information with the intent to gain understanding, verify information, or identify information that will require docketing to support the basis of the licensing or regulatory decision. In the interest of efficiency and effectiveness, the NRC staff is combining its audit efforts into a single activity for both reviews.

The NRC staff from the Office of Nuclear Reactor Regulation (NRR) has initiated its review of the LAR in accordance with NRR Office Instruction LIC-101, License Amendment Review Procedures (ML19248C539).

2.0 REGULATORY AUDIT BASIS A regulatory audit is a planned license-or regulation-related activity that includes the examination and evaluation of primarily non-docketed information associated with the LAR. An audit is conducted to gain understanding, to verify information, and to identify information that will require docketing to support the basis of a licensing or regulatory decision. An audit will assist the NRC staff in efficiently conducting its review and gaining insights to the licensees processes and procedures. Information that the NRC staff relies upon to make the safety determination must be submitted on the docket. This audit will be conducted in accordance with NRR Office Instruction LIC-111, Regulatory Audits, Revision 1, dated October 2019 (ML19226A274), with exceptions noted within this audit plan.

The NRC staff will perform the audit to support its evaluation of whether SNCs LAR can be approved per Title 10 of the Code of Federal Regulations (10 CFR), Section 50.90, Application for amendment of license, construction permit, or early site permit. The NRC staffs review will be informed by NUREG-0800, Standard Review Plan. The audit will assist the NRC staff with understanding the licensees proposed TS changes and associated technical and regulatory analysis.

3.0 INFORMATION AND OTHER MATERIAL NECESSARY FOR THE AUDIT Questions 15 through 34 were added to the audit plan dated March 26, 2025. Questions 1 through 14 were part of the audit plan dated March 26, 2025, and are repeated here for continuity.

Item #

Audit Questions/Discussions 1

L-2025-LLA-0011 - Describe what actions will be taken if TS 3.3.8, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, Condition J is entered due to three or more inoperable channels while the unit is in Modes 4 or 5, assuming there is no other inoperable equipment. Would actions be taken to establish Reactor Coolant System (RCS) vented as is currently required? If the proposed Required Action to initiate action to restore channel(s) to OPERABLE status is performed, is there an associated Completion Time to have the channels restored to OPERABLE?

Could the plant continue in Modes 4 or 5 indefinitely with these channels inoperable?

2 L-2025-LLA-0011 - Function 15 (Core Makeup Tank (CMT) Level - Low 3) in TS Table 3.3.8-1 is to ensure continued passive injection of borated water to the RCS following a small-break loss-of-coolant accident (SBLOCA). If this function is inoperable in Mode 4 or 5, and Condition J is entered, discuss differences in the plant/operator response to a SBLOCA if the RCS is vented (as currently required) -

versus the RCS not vented.

3 L-2025-LLA-0011 - Describe what actions will be taken if two items that are currently in conflict with each other are both inoperable from Modes 4 or 5. For example, what actions would be taken if both TS 3.3.8, ESFAS Instrumentation, Condition J (which currently requires establishing RCS vented) and Condition M (which currently requires establishing a pressurizer level 20% with the RCS pressure boundary intact) are entered. If the proposed required action to initiate action to restore Item #

Audit Questions/Discussions channel(s) to OPERABLE status is performed for both, is there a Completion Time to have the channels restored to OPERABLE? Could the plant remain in Modes 4 or 5 indefinitely with both functions inoperable?

4 L-2024-LLA-171. Section 4.2, Precedent, states, On April 6, 2022, Dominion Energy submitted a request to revise the TS for Millstone Power Station, Units 2 and 3 (Reference 12). The amendment (Reference 13) revised the TS requirements for mode change limitations to adopt the provisions of TSTF-359, Increased Flexibility in Mode Restraints. The availability of TSTF-359 for adoption by licensees was announced in the Federal Register on April 4, 2003 (68 FR 16579), as part of the [Consolidated Line-Item Improvement Process] CLIIP. The NRC staff reviewed the Federal Register notice, which stated, The staff believes that good practice should dictate that plant startup should normally be initiated only when all required equipment is operable and that startup with inoperable equipment must be the exception rather than the rule. The NRC staff did not find this wording in the LAR application. Explain where this wording can be found in the LAR or justify why it is acceptable for the language and regulatory context to be omitted or excluded in the proposed changes.

5 L-2024-LLA-171. Regarding Attachment 4 to NL-24-0412, Assessment for Precluding Use of LCO 3.0.4.b, the discussion regarding Mode 5 and Mode 6 appears to be incomplete and not consistent with TSs. As discussed in NUREG-2124, Final Safety Evaluation Report Related to the Combined Licenses for Vogtle Electric Generating Plant, Units 3 and 4, most of the low-power and shutdown risk occurs when the RCS is open and possibly drained, which occurs in Modes 5 and 6. In this condition, the passive residual heat removal (PRHR) system is not available for decay heat removal.

a. The discussion of Mode 5 does not include Mode 5 conditions with the RCS intact, which permits PRHR to remove decay heat as discussed in TS LCO 3.5.5. The Mode 5 discussion does not reference or discuss the automatic depressurization system (ADS) and associated TS LCO 3.4.12, Shutdown, Mode 5, RCS Intact, which would be expected to be included in this assessment for precluding the use of LCO 3.0.4.b. Explain why this discussion and context is not included or justify the technical and regulatory basis for its exclusion.

b. The discussion of Mode 5 conditions with the RCS open does not explain that the RCS is required to be open by ADS stages 1, 2, and 3 as stipulated in TS LCO 3.4.13, ADS-Shutdown, RCS Open.

Explain why this discussion is not included or justify the technical and regulatory basis for its exclusion.

c. Attachment 4 states, Loss of RCS heat removal can be mitigated by one of two trains of RNS [normal residual heat removal system],

gravity injection from RNS, or gravity injection from the IRWST [in containment refueling water storage tank]. This discussion appears to omit the need for ADS stage 4 to be opened when using gravity injection to prevent the surge line from flooding the pressurizer post RCS boiling. Attachment 4 does not appear to consider the context of Item #

Audit Questions/Discussions the requirement for ADS stage 4 to be operable when gravity injection is needed to mitigate a loss of decay heat removal. Explain why this discussion of the need for ADS stage 4 to be opened when using gravity injection is not included or justify the technical and regulatory basis for its exclusion.

6 L-2024-LLA-171. Regarding attachment 4 to NL-24-0412, no statements are made regarding containment closure, a key fission product barrier and defense-in-depth barrier. Explain the containment closure TSs consideration and use of the proposed LCO 3.0.4b?

7 L-2024-LLA-171. NUMARC 91-06, Guidelines for Industry Actions to Assess Shutdown Management, is the technical basis for the shutdown 10 CFR 50.65(a)(4) evaluations, as described in NUMARC 93-01. NUMARC 93-01 was developed for operating pressurized water reactors (PWRs) and boiling water reactors (BWRs).

Operating BWRs and PWRs do not have comparable passive emergency core cooling systems (ECCS) or regulatory treatment of non-safety systems (RTNSS) systems, as referenced in SECY 94-084, Policy and Technical Issues Associated with the Regulatory Treatment of Non-Safety Systems in Passive Plant Designs. In the Vogtle Units 3 and 4 design, general design criterion (GDC) 34 is met in Modes 5 and 6 by the automated and passive actuation of the gravity-driven core cooling system combined with actuation of ADS, which is covered in Technical Specifications. According to LCO 3.5.7, IRWST - Shutdown, MODE 5, and LCO 3.5.8, IRWST - Shutdown, MODE 5, only one injection flow path and one containment recirculation isolation valve need to be operable in Modes 5 and 6 for sustained decay heat removal. How is this lack of redundancy considered in and the maintenance rule a (4) risk evaluations?

8 L-2024-LLA-171. The risk significant RNS system and risk significant support systems for RNS provide non-safety related decay heat removal when the RCS is open. The risk significant RNS system and its support systems have availability controls when the RCS is not intact and could be drained, which are referenced in the final safety analysis report (FSAR) Table 16.3-1, List of Investment Protection Short-term Availability Controls. In this context, the NRC staff has three questions.

a. How are the attributes of decay heat removal capability, inventory control, and power availability from NUMARC 93-01, Revision 4f, Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Section 11.3.6, Assessment Methods for Shutdown Conditions, assessed in context of the RTNSS controls, which do not appear to be discussed in the LAR?

b. Is the low-power shutdown probabilistic risk assessment (PRA) that was referenced in revision 10 of the FSAR used for these evaluations?

c. How are these risk evaluations for 3.0.4.b consistent with Generic Letter 88-17, Loss of Decay Heat Removal - 10 CFR 50.54(f),

recommendations which are considered key voluntary initiatives to be maintained in SECY 97-168, Issuance for Public Comment of Item #

Audit Questions/Discussions Proposed Rulemaking For Shutdown and Fuel Storage Pool Operation?

9 L-2024-LLA-171. Regarding Attachment 3 to NL-24-0412, Associated Technical Specification Bases Changes, the results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented. Federal Register Notice 68 FR 16579 states, NUMARC 93-01 Section 11 requires that this assessment process be proceduralized. Furthermore, Inspection Procedure 71111.13 provides inspection guidance on, among other things, the verification of the performance of maintenance risk assessments, the adequacy of risk assessments and the management of the resulting risk. How can the adequacy or sufficiency of these risk assessments be ensured if these risk assessments do not need to be documented?

Regulatory Guide (RG) 1.174, Revision 3, An Approach for Using Probabilistic Risk Assessment in Risk-informed Decisions on Plant-Specific Changes to the Licensing Basis, provides different approaches for the review of deterministic applications with risk insights versus risk-informed applications. Section 2.4, Description of the Proposed Changes, of the Vogtle 3 and 4 mode restraints LAR states that the proposed LCO 3.0.4.b, applies a consistent, risk-informed basis in all conditions when changing MODES or other specified conditions in the Applicability while relying on ACTIONS, except when applying LCO 3.0.4.a or LCO 3.0.4.c. Section 3.0, Technical Evaluation, of the Vogtle 3 and 4 shutdown actions LAR states that the proposed changes do not impact the generic AP1000 shutdown evaluation initiating event frequencies or risk insights. Are either the Vogtle 3 and 4 Mode Restraints LAR or the Shutdown Actions LAR intended to be risk-informed submittals compliant with RG 1.174, Revision 3? Also, please explain how the risk insights address defense-in-depth and safety margins.

10 L-2024-LLA-171. Attachment 4 to NL-24-0412 states, considerations above are addressed in the AP1000 probabilistic risk assessment (PRA) developed to support the application for Design Certification (ML042230497), because it included a low-power and shutdown risk assessment, and the first approach taken in this qualitative assessment is to use results and insights from the assessment of low-power and shutdown (LPSD) risk to identify any systems or components that are potentially risk significant during shutdown modes (Modes 4, 5, 6), hot standby (Mode 3), or startup (Mode 2). In several places throughout the submittal, the LAR references using results and insights from the LPSD risk assessment.

NUREG-0800, Revision 3, Section 19.1 Determining the Technical Adequacy of Probabilistic Risk Assessment for Risk-Informed License Amendment Requests After Initial Fuel Load (ML12193A107), Subsection III.1.3, Parts of the PRA Model Used in Application, states, When using risk insights based on a PRA model, the PRA model should reasonably represent the as-built and as-operated plant. Has the LPSD PRA used to support the AP1000 design certification application been updated to reflect the current Vogtle 3 & 4 plant configuration and operation? If not, why does the LPSD PRA model reasonably represent the as-built and as-operated plant, and why are the risk results and risk insights appropriate for this LAR?

Item #

Audit Questions/Discussions 11 L-2024-LLA-171. Attachment 4 to NL-24-0412 states that the final consideration is whether the equipment, if inoperable, can be adequately assessed by the applicable method(s) of risk assessment employed for a potential use of LCO 3.0.4.b. What are the applicable method(s) this statement is referencing? How were they evaluated in Attachment 4 to determine which LCOs would not be able to use the allowances of LCO 3.0.4.b?

12 L-2024-LLA-171. Attachment 4 to NL-24-0412 states, Since initiating events during RCS drained conditions (one foot below the vessel flange or lower) contribute approximately 95% of the total shutdown Core Damage Frequency (CDF), common cause failure of the above components are significant only for drained conditions.

Drained conditions occur in:

Mode 5 with the RCS open (vented) and RCS level at the flange or lower, or Mode 6 with the RCS level at the flange or lower These risk conclusions do not appear to be consistent with the LPSD PRA results documented in NUREG-1793, Final Safety Evaluation Report Related to Certification of the AP1000 Standard Design. In contrast, NUREG-1793, Chapter 19, Severe Accidents, states, over 90 percent of the AP1000 shutdown risk occurs during vented, drained conditions. This plant configuration occurs during cold shutdown when the RCS boundary is open (via stages 1, 2, and 3 of the ADS),

and the RCS is drained to reach midloop conditions so that nozzle dams can be installed in the hot-and cold-legs to perform SG Maintenance. Consequently, the dominant portion of shutdown risk occurs when the RCS is vented, and PRHR is no longer functional. This high-risk period starts when the RCS is vented, not when RCS level is at the one foot below the flange or lower. It is not apparent if only RCS drained conditions were assessed or if all conditions were considered in the evaluating risk. Explain if lower risk operating states have been assessed, including complications, or if only high-risk operating states assessed and justify the basis for the chosen risk evaluation.

13 L-2024-LLA-171. Attachment 4 to NL-24-0412 states that condition I requires exiting the Applicability by initiating action to remove the upper internals. Since the TS Applicability includes a condition for which the equipment is potentially risk significant based on insights from the LPSD risk assessment, the final consideration is whether the inoperable channels can be adequately assessed by the shutdown defense-in-depth risk assessment. Attachment 4 to NL-24-0412 also states that the TS Applicability includes modes/conditions for which the equipment is risk significant based on insights from the LPSD risk assessment, so the final consideration is whether inoperable flow paths in ADS stage 4 can be adequately assessed by the shutdown defense-in-depth risk assessment. Clarify how these inoperable channels would be adequately evaluated using the shutdown defense-in-depth risk assessment.

Item #

Audit Questions/Discussions 14 L-2024-LLA-171. Attachment 4 to NL-24-0412 states that the LPSD PRA concluded that a very small contribution to shutdown CDF is from the conditions of RCS pressure boundary intact and pressurizer level 20%; therefore, the equipment is not risk significant for the TS 3.4.12 Applicability and use of LCO 3.0.4.b is reasonable. NUREG-0800, Revision 3, Section 19.1 states the following, To assess the technical adequacy of the PRA input for a decision, the licensee should identify which parts of the PRA are used to provide the results that will be compared to acceptance criteria or guidelines that apply to the application. Attachment 4 of the Vogtle 3 and 4 mode restraints LAR states the allowance for LCO 3.0.4.b is reasonable for Modes 1, 2, 3, and 4 and for Mode 5 with the RCS pressure boundary intact and pressurizer level 20% because there are no components/systems with a significant contribution to LPSD CDF for those modes/conditions. Justify why components/systems were not evaluated for individual modes/conditions.

15 L-2025-LLA-0011. Section 2.3 of the LAR makes an argument to justify the change that may not accurately characterize the options operators would have in the situation with conflicting requirements in required actions. Specifically, the LAR states, When Immediately" is used as a Completion Time, the Required Action should be pursued without delay and in a controlled manner. Thus, actions taken contrary to this direction would not be allowed, i.e., actions taken to exit the Applicability that would necessitate a water level change that conflict with the Required Action.

Technical Specification (TS) Actions only apply when the plant is operating with the applicable mode. This is consistent with 10 CFR 50.36 which specifically requires a licensee to either shut down or follow any remedial action permitted by their TS when an LCO is not met. Based on the regulation shutting down the reactor (or exiting the applicable mode) should always be an option. Vogtle, Units 3 and 4, LCO

3.0.2 states

Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and 3.0.6.

If the LCO is met, or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required, unless otherwise stated. [Emphasis added]

This is further supported by the Vogtle, Units 3 and 4, Bases for LCO 3.0.2 which state, Completing the Required Actions is not required when an LCO is met, or is no longer applicable, unless otherwise stated in the individual Specifications.

[Emphasis added]

Item #

Audit Questions/Discussions

a. Please explain whether the intent of the above statement from Section 3.2 of the LAR is stating that in the example case provided in the LAR, the licensee does not believe that exiting the Applicability is an option or if they believe that any time the plant is in a Condition with a specified immediate action, operators are prevented from exiting the Applicability.

b. If, as stated in the LAR, a Completion Time of Immediately prevents operator from taking any other action than the specified Required Action, doesnt the portions of the proposed change to revise the Required Actions to Initiate action to restore [compliance with the LCO] with an Immediately Completion Time result in the same problem? The only choice would be to restore compliance with the specified LCO. Section 3.0 of the LAR states that exiting the Applicability would be an option with the proposed structure, but exiting the Applicability does not restore compliance. It places the plant in a mode where compliance is not required.

The LCO would still not be met. If exiting the Applicability is acceptable with the proposed revision, then why isnt it acceptable with current technical specification requirements. The NRC notes that the licensee is obligated to enter all applicable LCOs concurrently along with Required Actions and Completion Times for each associated LCO Condition.

For example, LCO 3.5.3, CMTs - Shutdown, RCS Intact, Action E could be entered in Mode 5 with the RCS intact after failure to restore the inoperable CMT to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. In this case, the current requirement is to exit the applicability by venting the RCS. The current action seems appropriate if the IRWST is operable. Under the proposed change, the operator would E.1 Initiate Action to be in MODE 5, (which would already be complete in this case) AND E.2 Initiate action to restore required CMT to OPERABLE status. The proposed addition of Action E.2 tells the operator to keep trying to restore after they have already been unable to do so for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Would the operator (based on the discussion in Section 2.3 of the LAR) be prevented from venting the RCS to exit the Applicability in this case?

16 L-2025-LLA-0011. In Section 3.0 of the LAR, the technical evaluation of the TS 3.3.8, Containment Pressure, Action P states that there is a potential conflict between the Action P requirement to Open a containment air flow path 6 inches in diameter and LCO 3.6.7 which requires capability of closing the containment prior to steaming. The LAR states, Required Action P.3 represents a potential conflict with a necessary post-event action. The priority of closing the containment during such an event should not be constrained by a conflicting TS required action that would apply once the Applicability was reentered when there was no longer an open containment air flow path 6 inches in diameter.

The NRC staff has the following questions about this discussion in the LAR:

Item #

Audit Questions/Discussions

a. Please explain what is meant by post-event action. TSs do not provide procedures for mitigation of events. If the event occurs, then the licensee would follow their event response or emergency operating procedures to mitigate the event. This question also applies to the LAR technical evaluation of TS 3.3.9, Function 15 (as well as several other TSs evaluated in Section 3.0 of the LAR), which also cites an apparent conflict with a necessary post-event action.

b. The Vogtle, Units 3 and 4, Bases for LCO 3.6.7 state, In MODES 5 and 6, there is no potential for steam release into the containment immediately following an accident.

Pressurization of the containment could only occur after heatup of the IRWST due to Passive Residual Heat Removal Heat Exchanger (PRHR HX) operation (MODE 5 with RCS intact) or after heatup of the RCS with direct venting to the containment (MODE 5 with reduced RCS inventory or MODE 6 with the refueling cavity not fully flooded) or after heatup of the RCS and refueling cavity (MODE 6 with refueling cavity fully flooded). The time from loss of normal cooling until steam release to the containment for four representative sets of plant conditions is shown in Figure B 3.6.7-1 as a function of time after shutdown. Because local manual action may be required to achieve containment closure it is assumed that the containment hatches, air locks and penetrations must be closed prior to steaming into containment.

Based on Table 3.6.7-1, Operators would have several hours as a minimum to close containment before steaming could begin in MODE 5. Please explain how Action P would constrain closure of containment in event response.

17 L-2025-LLA-0011. In TS 3.3.8, the initial Actions for an inoperable channel(s) is to first place the inoperable channel in bypass or trip (Actions A and B). If Actions A or B are successful, then the Actions C through Q would not apply. The proposed revisions to TS 3.3.8 Actions include replacing the Actions to exit the applicability with Actions requiring restoring compliance with the LCO. Why wouldnt it be appropriate to give operators the option to place the inoperable channels in trip or bypass?

18 L-2025-LLA-0011. TS 3.3.9, Actions A and B require restoring inoperable manual initiation channels to operable status with completion times of 48 and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If these Actions are not met, then the current requirement to exit the Applicability is required. Is there reason to believe that these restoration times of Actions A and B are inadequate? The proposed changes to the Actions G, H, J and L would simply require continued attempts to restore operability with no requirement to ever exit the applicability. Please discuss how an operator would determine when it is appropriate to exit the Applicability.

Item #

Audit Questions/Discussions 19 L-2025-LLA-0011. Does inoperability of TS 3.3.9, Function 15, Containment Vacuum Relief Valve Actuation - Manual Initiation, represent a loss of function condition? The technical evaluation of the proposed changes relate to this function does not discuss alternative ways (e.g., automatic initiation) of accomplishing the function when manual initiation is inoperable.

20 L-2025-LLA-0011. TS 3.5.3, Action E.1 is marked up to state, Initiate action to be in Mode 5 with an immediate Completion Time (CT). This appears to be inconsistent with other specifications which allow 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to go from Mode 4 to Mode 5. Why is this action using an immediate CT in lieu of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for this action?

21 L-2025-LLA-0011. The default actions proposed for revision in the LAR, in some cases, can only be entered after having already failed to restore operability (or compliance with the LCO) for a period of time. The default actions in these cases are proposed to be revised to state initiate action to restore operability or compliance with the LCO. Why are the words initiate action to necessary? The action to restore would have already been started in response to the previous Required Actions. Why wouldnt it be more appropriate to state restore operability or compliance with an immediate CT? For instance, LCO 3.5.3 provides at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> (and as much as 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) to restore the required CMT to operable status prior to entry into Condition E. Condition E cannot be entered without having failed to restore operability in accordance with Required Actions A.1, B.1 and D.1.

As stated in Section 1.3 of the Vogtle, Units 3 and 4, TS, [w]hen Immediately is used as a Completion Time, the Required Action should be pursued without delay and in a controlled manner. Initiate action to makes sense when the default Condition can be entered directly. But if the default Condition can only be entered after having time to restore compliance with the LCO, its unclear as to why the words are necessary. The Writers Guide for Plant-Specific Improved Technical Specifications (TSTF-GG-05-01) makes it clear that an immediate CT is appropriate in cases where It is desired to accomplish an action without delay (given the potential surrounding circumstances), but the time necessary to complete the action may vary widely based on a number of unknowns. In this situation, it may be inappropriate to require the completion of the action within a specific time.

22 L-2025-LLA-0011. The technical evaluation for modifying Required Action E.1 of TS 3.5.3, CMTs - Shutdown, RCS Intact states removing the portion of the action that exits the Applicability has no adverse effect on safe plant operation. However, the updated final safety analysis report (UFSAR), Subsection 19E.4.8.2, Loss of Normal Residual Heat Removal System Cooling in MODE 4 and MODE 5 with Reactor Coolant System Intact. states, [f]ollowing the loss of RNS cooling, there is no active mechanism for heat removal from the RCS. And, as further described in that Section of the UFSAR, the CMT is credited in mitigating a loss of RNS event.

Condition E would only be entered if operators had been unable to restore the required CMT to operable status over the previous 24 to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Please explain how staying in the Applicability when operators are clearly having difficulty restoring an inoperable CMT would not represent at least some increase in plant risk? The LAR statement seems to contradict the technical specification. If there is no impact on plant safety for operating in MODE 5 without any CMTs, then why would one be Item #

Audit Questions/Discussions required to be operable? Note, this question similarly applies to the technical evaluation in the LAR for TS 3.5.5, PRHR HX - Shutdown, RCS Intact, Action E.

23 Regulatory Basis:

Criterion 34Residual heat removal. A system to remove residual heat shall be provided. The system safety function shall be to transfer fission product decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded. Suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities shall be provided to assure that for onsite electric power system operation (assuming offsite power is not available) and for offsite electric power system operation (assuming onsite power is not available) the system safety function can be accomplished, assuming a single failure.

Commission Direction:

SRM to SECY-97-168, Issuance For Public Comment Of Proposed Rulemaking Package For Shutdown And Fuel Storage Pool Operation, states, The Commission expects the staff to continue to monitor licensee performance, through inspections and other means, in the area of shutdown operations to ensure that the current level of safety is maintained. The regulatory analysis assumed two trains of RHR and support equipment consistent with Standard Technical Specifications. The regulatory analysis credited voluntary initiatives, and it also credited equipment required by TS and equipment recommended to be available based on guidance from Generic Letter 88-17 and NUMARC 91-06.

The 1995 Commission Probabilistic Risk Analysis (PRA) policy statement states, The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the art in PRA methods and data and in a manner that complements the NRCs deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

Audit Question:

L-2025-LLA-0011. There are multiple examples in the Vogtle 3 and 4 Shutdown Actions LAR (ML25023A275) where a potential conflict is identified with the limiting condition for operations (LCO) for TS 3.6.7, Containment Penetrations. Numerous LCOs require the opening of a containment air flow path 6 inches, the containment equipment hatch, or the containment airlock while LCO 3.6.7 requires the capability of closing the containment prior to steaming (e.g., loss of decay heat removal that would lead to steaming in containment).

a. Would adding a Note to all the applicable TS Conditions, which clarifies that the containment opening does not have to be maintained when containment closure is required to be implemented (post-event), resolve the conflict in TS?

b. If adding clarifying notes is not desired, justify why the submitted changes to all the LCO Actions would result in an improvement in safety during shutdown operations. Include in this discussion any impacts an opening in containment 6 inches (including the containment equipment hatch or Item #

Audit Questions/Discussions the containment airlock) has on the baseline risk profile in MODEs 5 and 6.

c. Does the Vogtle 3 and 4 NUMARC 93-01 shutdown risk management implementation procedure address the risk posed by the potential for excessive containment cooling resulting in a negative containment pressure below the design limit?

24 L-2025-LLA-0011. There are multiple examples within the Vogtle 3 and 4 Shutdown Actions LAR where it is stated that events requiring actuation of ADS Stage 1, 2, and 3 or ADS Stage 4 do not assume a minimum RCS water level. Additionally, there is a statement that events requiring isolation of the Chemical and Volume Control System (CVS) letdown isolation valves do not assume a minimum RCS level.

a. Are the event sequences where ADS actuation or CVS letdown isolation is fully successful the only event sequences that do not assume a minimum RCS water level? If a minimum RCS water level is assumed for other event sequences, what are the other event sequences and what is the assumed required water level?

b. Is a 20% pressurizer water level necessary for preventing core damage in any low-power shutdown event sequence? If a 20% pressurizer water level is necessary, what are the event sequences where such a level is required?

c. Is a pressurizer water level above the P-12 interlock necessary for preventing core damage in any low-power shutdown event sequence? If a pressurizer water level above the P-12 interlock is necessary, what are the event sequences where such a level is required? Please include a discussion on the necessity of the automatic features which are enabled when pressurizer water level is above the P-12 interlock setpoint.

d. What RCS water inventory or water levels are credited in the Vogtle 3 and 4 NUMARC 93-01 shutdown risk management implementation procedures? If RCS inventory / levels are credited, please describe the conditions for which they are credited.

25 L-2025-LLA-0011. There are multiple examples within the Vogtle 3 and 4 Shutdown Actions LAR where it states that once the RCS is vented, the complement of safety systems available to respond to an event is reduced.

a. How do the baseline design certification LPSD PRA risk profiles of the AP1000 in an RCS vented state compare with the baseline risk profiles with the RCS pressure boundary intact?

b. What are the primary drivers for the differences in the baseline design certification LPSD PRA risk profiles of the AP1000 RCS vented state compared with the baseline risk profiles with the RCS pressure boundary intact?

c. Are the RCS vented or RCS pressure boundary intact states credited in the Vogtle 3 & 4 NUMARC 93-01 shutdown risk management implementation procedures? If these states are credited, what are they credited for?

26 L-2025-LLA-0011. There are multiple examples within the Vogtle 3 and 4 Shutdown Actions LAR where it is stated that the removal of the upper internals would preclude Item #

Audit Questions/Discussions normal outage progression to proceed to Mode 5 with the reactor vessel head on, which would establish an additional fission product boundary.

a. How do the baseline risk profiles of the AP1000 with the upper internals removed compare with being in Mode 5?

b. Is the removal of the upper internals credited in the Vogtle 3 and 4 NUMARC 93-01 shutdown risk management implementation procedures? If it is credited, please describe the conditions for which they are credited?

27 L-2025-LLA-0011. There are multiple examples within the Vogtle 3 and 4 Shutdown Actions LAR where Mode 6 LCO actions to initiate action to establish water level 23 feet above the top of the reactor vessel flange are eliminated.

a. Is there a LPSD plant operating state that has a lower baseline risk profile than establishing water level 23 feet above the top of the reactor vessel flange (other than when all fuel is removed from the core)?

b. If there is no LPSD plant operating state with a lower baseline risk profile, justify why giving operators additional flexibility for the associated TS LCOs would not impact the following:

i. LPSD evaluation provided with the generic AP1000 Design Certification Document ii. Defense-in-depth model iii. LPSD risk assessment 28 L-2025-LLA-0011. On page E-20 of the Vogtle 3 and 4 Shutdown Actions LAR, the following statement is made with respect to TS 3.3.9, Function 7, ADS Stage 4 Actuation - Manual Initiation, Actions H & I: Once the RCS is vented (i.e., the first portion of Required Action H.2) the remaining requirement to establish 20%

pressurizer level is not required to be completed since the Applicability would be exited.

To exit the mode of applicability (Mode 5) for this function and Action (H.2),

operators would have to take the plant to either Mode 4 or Mode 6. vented is defined in the Vogtle 3 & 4 TS as a condition when all required flow paths in ADS stage 1, 2, and 3, or alternative flow path with equivalent area Mode 4 and Mode 6, as defined in Section 1.1 of the Vogtle 3 and 4 TS, does not require the RCS to be vented.

a. How does the act of taking the RCS to a vented status affect the requirement to establish a level in the pressurizer?

b. Would removing the reactor vessel head satisfy any requirement to establish RCS vented? If removing the reactor vessel head would satisfy the requirements to establish RCS vented, why would providing an OR Action for H.2 to immediately initiate action to enter Mode 6 address the conflict with the 20% pressurizer level requirement?

29 L-2025-LLA-0011. Starting at the bottom of page E-24 of the Vogtle 3 and 4 Shutdown Actions LAR, the following statement is made with respect to TS 3.4.12 Item #

Audit Questions/Discussions ADS - Shutdown, RCS Intact: Action D could be reached without operability of ADS Stage 1, 2, and 3 being impacted. Describe how Action D of TS 3.4.12 could be reached with ADS Stage 1, 2, and 3 still being operable. TS 3.4.12 is specific to ADS operability, and in order to reach Action D of the TS, ADS Stage 1, 2, and 3 would have to be inoperable.

30 L-2025-LLA-0011. Does the Vogtle 3 and 4 NUMARC 93-01 implementing procedure credit all systems whose TS will be altered by the Vogtle 3 and 4 Shutdown Action LAR? If any system is not credited, how will the risk mitigated by these systems be evaluated with respect to shutdown risk management?

31 L-2025-LLA-0011. On page E-30 of the Vogtle 3 and 4 Shutdown Actions LAR, the following statement is made with respect to TS 3.6.7 Containment Penetrations:

Neither Required Actions B.1.1 or B.1.2 would exit the Mode 5 or 6 applicability.

Are these containment penetrations covered in all operating modes by other TS -

such as TS 3.6.1 Containment and TS 3.6.3 Containment Isolation Valves? If they are not covered in all operating modes, please explain how a containment penetration could be inoperable in Modes 5 and 6 but operable in Modes 1-4. If they are covered in all operating modes, justify why allowing more flexibility to leave the Mode 5 or 6 applicability is necessary.

32 L-2025-LLA-0011. What is the definition of RCS Intact or RCS Open as it applies to the Vogtle 3 and 4 TS, and identify where this definition is located?

33 L-2025-LLA-0011. On page E-13 of the Vogtle 3 & 4 Shutdown Actions LAR, the following statement is made: The VEGP Units 3 and 4 UFSAR Section 19.19 (Table 19.19-1) identifies the AP1000 shutdown CDF of 1.91 E-7 from internal events, internal floods, and internal fires, Later in the paragraph it states that the AP1000 shutdown CDF is approximately an order of magnitude below the VEGP at power CDF of 1.53E-06 (UFSAR Table 19.9-1) from internal events, internal floods, internal fires, and seismic events. Why is seismic CDF not considered at shutdown like it is at power?

34 L-2025-LLA-0011. The NRC staff reviewed the Vogtle Unit 3 and 4 FSAR, Revision 12, Table 19E.2-1 Evaluation of a Loss of RNS at Mid-Loop With No IRWST Injection. Time to RCS boiling after a loss of RNS (normal residual heat removal system) was estimated as 3 minutes, time to having an empty hot leg was estimated as 20 minutes, and time to core uncovery was estimated as 42 minutes. Actuation of fourth-stage ADS valves occurs on Low-4 (empty) hot leg level with a two-out-of-two logic. Actuation of fourth-stage ADS causes actuation of IRWST injection. If the fourth-stage ADS valves do not open, there is no apparent capability for gravity injection in this passive design.

Regarding the risk significance of this configuration, Vogtle, Units 3 and 4, FSAR, Revision 12, Section 19.19, Core Damage and Severe Release Frequency from Events at Shutdown, states that the dominant accident sequences comprise 95.3 percent of the level 1 shutdown PRA core damage frequency. These dominant sequences consist of:

Item #

Audit Questions/Discussions Loss of component cooling or service water system initiating event during drained condition with a contribution of 76.7 percent of the CDF Loss of RNS initiating event during drained condition with a contribution of 10.4 percent of the CDF Loss of offsite power initiating event during drained condition with a contribution of 8.2 percent of the CDF All three sequences rely on actuation of ADS Stage 4 and the gravity injection through the gravity-driven core cooling system to prevent core damage. During RCS vented conditions, the first three stages of ADS are required by TS to be opened, and, therefore, decay heat removal using the PRHR and the CMTs is not feasible.

The verification of ADS Stage 4 operability before entering into reduced inventory operation would be consistent with NUMARC 91-06 guideline, Prior to entering a REDUCED INVENTORY condition, equipment requirements that provide or support KEY SAFETY FUNCTIONS should be verified. Generic Letter 88-17 also states that prior to entering into a reduced inventory condition, Provide at least two available or operable means of adding inventory to the RCS that are in addition to pumps that are a part of the normal DHR systems. In addition, ADS Stage 4 and the gravity-driven core cooling system meet GDC 34.

a) How have NUMARC 91-06 and Generic Letter 88-17 guidelines been used to address the proposed TS action statements associated with ADS Stage 4 and the gravity-driven core cooling system during RCS vented condition?

b) Is the operability of ADS Stage 4 verified before entering into reduced inventory operation?

i.

If ADS Stage 4 operability is not verified, justify why a verification of operability is not performed.

ii. If ADS Stage 4 operability is verified, describe how that verification is performed.

ML25097A184 NRR-106 OFFICE NRR/DORL/LPLII-1/PM NRR/DORL/LPLII-1/LA NRR/DRA/APLC/BC NRR/DSS/STSB/BC NAME JLamb KZeleznock SAlferink (A)

RElliott for SMehta DATE 04/07/2025 04/11/2025 04/11/2025 04/14/2025 OFFICE NRR/DORL/LPLII-1/BC NRR/DORL/LPLII-1/PM NAME MMarkley JLamb DATE 04/16/2025 04/16/2025