ML25268A010
| ML25268A010 | |
| Person / Time | |
|---|---|
| Site: | Watts Bar  |
| Issue date: | 11/21/2025 |
| From: | Kimberly Green Plant Licensing Branch II |
| To: | Erb D Tennessee Valley Authority |
| Green K | |
| References | |
| EPID L-2024-LLA-0175 | |
| Download: ML25268A010 (0) | |
Text
November 21, 2025 Mr. Delson C. Erb Vice President, OPS Support Tennessee Valley Authority 1101 Market Street, LP 4A-C Chattanooga, TN 37402-2801
SUBJECT:
WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 - REGULATORY AUDIT
SUMMARY
REGARDING LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2, PROVIDE RISK-INFORMED EXTENDED COMPLETION TIMES - RITSTF INITIATIVE 4B (EPID L-2024-LLA-0175)
Dear Mr. Erb:
By letter dated December 27, 2024, the Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for Facility Operating License Nos. NPF-90 and NPF-96 for Watts Bar Nuclear Plant, Units 1 and 2, respectively. The proposed amendments would adopt Technical Specifications Task Force (TSTF) Traveler TSTF-505, Revision 2, Provide Risk-informed Extended Completion Times, RITSTF Initiative 4b, to permit the use of risk-informed technical specification completion times for certain actions required when a limiting condition for operation (LCO) is not met. Additionally, TVA requested to adopt Traveler TSTF-439-A, Revision 2, Eliminate Second Completion Times Limiting Time From Discovery of Failure To Meet an LCO.
The Nuclear Regulatory Commission (NRC) staff reviewed TVAs LAR and determined that a regulatory audit would assist in the timely completion of the LAR review. The NRC staff conducted a regulatory audit that consisted of a remote audit March 17 - September 5, 2025.
An audit plan was provided to Ms. Hulvey of your staff by letter dated February 26, 2025. The audit plan included a list of items to be provided on an electronic portal. This list was augmented by a list of audit questions that was transmitted to Ms. Shawna Hughes by email dated June 18, 2025. The audit questions are contained in an attachment to the enclosure. The staff audited these documents and held discussions with members of TVA on the LAR relevant to the NRC staffs review.
As a result of the audit, no requests for additional information were issued. A summary of the audit is enclosed.
Sincerely,
/RA/
Kimberly J. Green, Senior Project Manager Plant Licensing Branch II-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-390 and 50-391
Enclosure:
Audit Summary cc: Listserv
Enclosure REGULATORY AUDIT
SUMMARY
RELATED TO LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATIONS TO ADOPT TSTF-505, REVISION 2 TENNESSEE VALLEY AUTHORITY WATTS BAR NUCLEAR PLANT, UNITS 1 AND 2 DOCKET NOS. 50-390 AND 50-391
1.0 BACKGROUND
By letter dated December 27, 2024, Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for Watts Bar Nuclear Plant, Units 1 and 2 (WBN) (Agencywide Documents Access and Management System Accession No. ML24362A110). The proposed amendments would revise certain technical specifications (TSs) to permit the use of a risk-informed completion time (RICT) for actions to be taken when a limiting condition for operation (LCO) is not met. The proposed changes to implement an RICT are based on a Technical Specifications Task Force (TSTF) Traveler, TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (ML18183A493). The U.S. Nuclear Regulatory Commission (NRC or Commission) staff issued a final model safety evaluation approving TSTF 505, Revision 2, on November 21, 2018 (ML18269A041). The LAR also proposed to implement TSTF-439, Revision 2, Eliminate Second Completion Times Limiting Time from Discovery of Failure to Meet an LCO (ML051860296).
Information that the NRC staff relies upon to make the safety determination must be submitted on the docket. However, the NRC staff may review supporting information retained as records under Title 10 of the Code of Federal Regulations (10 CFR) 50.71 or 10 CFR 54.37, which, although not required to be submitted as part of the licensing action, would help the NRC staff better understand the licensees submitted information. To support its review of the LAR, the NRC staff issued a regulatory audit plan on February 26, 2025 (ML25051A176). The purpose of the audit is to review the documentation related to the subject of its application (e.g., calculations and reports) that were not submitted on the WBN docket, to acquire additional understanding about the amendment request, and to determine whether additional information is needed to be docketed to complete the NRC staffs safety evaluation.
2.0 AUDIT ACTIVITIES An audit team consisting of NRC staff conducted a remote regulatory audit to support the review of the LAR from March 17, 2025, through September 5, 2025. The purpose of the audit was to gain an understanding of the information needed to support the NRC staffs licensing decision regarding the LAR and to develop requests for additional information (RAIs), if needed. of this audit summary lists the NRC staff, TVA staff, and Jensen Hughes individuals that took part in or attended the audit. Throughout the audit period, the NRC audit team used an internet-based portal provided by TVA to review primarily non-docketed information related to the application. A formal, virtual audit meeting with the TVA was held from July 28-30, 2025, to facilitate technical discussions of audit questions according to the audit plan. A list of the audit questions discussed during the virtual audit meeting is provided in attachment 2.
Technical discussions were focused on the following major areas: probabilistic risk assessment, external hazards, TSs, digital instrumentation, and electrical engineering. On July 31, 2025, the NRC staff provided a brief conclusion of the formal, virtual audit meeting including audit objectives that were met and details on the path forward. There were no open items from audit discussions and no deviations from the audit plan. Non-docketed information provided by the licensee in response to audit questions and the formal meeting is listed in attachment 3 of this audit summary.
3.0 RESULTS OF THE AUDIT TVA stated that it would supplement the application with additional information based on the audit discussions. TVA submitted a supplement by letter dated August 20, 2025 (ML25232A156). The NRC staff will review TVAs supplement to assess if any additional information is needed to complete its review of the LAR.
Attachments:
- 1. List of Audit Participants
- 2. List of Audit Questions Discussed During the Audit
- 3. List of Documents Audited During the Audit List of Audit Participants NRC TVA Kimberly Green Bill Victor Jigar Patel Shawna Hughes Bob Pascarelli Gerry Kindred Jay Robinson Dan Kearnaghan Edmund Kleeh Mark Nicholson Sunwoo Park Jacob Johnson Andrea Russell Jonathan Gore Ellery Coffman Hongbing Jiang Norbert Carte Daniel Silverstein Nick Lovelace, Jensen Hughes Adrienne Brown Kelly Wright, Jensen Hughes Adakou Foli Jeff Schappaugh, Jensen Hughes Charles Moulton Sunwoo Park Evan Davidson List of Audit Questions Probabilistic Risk Assessment Licensing Branch A (APLA) Audit Questions QUESTION 01 - Probabilistic Risk Assessment (PRA) Model Uncertainty Analysis Process The NRC staff SE to Nuclear Energy Institute (NEI) 06-09-A specifies that the LAR should identify key assumptions and sources of uncertainty and assess and disposition each as to its impact on the risk-managed TS (RMTS) application. Section 2.3.4 of NEI 06-09-A states that PRA modeling uncertainties shall be considered in application of the PRA base model results to the RICT program and that sensitivity studies should be performed on the base model prior to initial implementation of the RICT program on uncertainties that could potentially impact the results of a RICT calculation. NUREG-1855, Revision 1, Guidance on the Treatment of Uncertainties Associated with PRAs in Risk-Informed Decision-Making, [Final] Report, dated March 2017 (ML17062A466), presents guidance on the process of identifying, characterizing, and qualitatively screening model uncertainties.
LAR enclosure 9 states that the process for identifying key assumptions and sources of uncertainty for the internal events (including internal floods), fire, and seismic PRAs was performed using the guidance in NUREG-1855, Revision 1. The LAR indicates that in addition to plant-specific assumptions and sources of uncertainty from the internal events (including internal floods), fire, and seismic PRA notebooks, that generic industry sources of uncertainty were also reviewed for applicability presented in Electric Power Research Institute (EPRI)
Topical Report (TR) 1016737, Treatment of Parameter and Model Uncertainty for Probabilistic Risk Assessments, and EPRI TR 1026511, Practical Guidance of the Use of Probabilistic Risk Assessment in Risk-informed Applications with a Focus on the Treatment of Uncertainty. The LAR states that no assumptions and sources of uncertainty were identified to have the potential to impact the TSTF-505 application (i.e., no key assumptions and sources of uncertainty). The LAR presents two sources of modeling uncertainty (i.e., modeling of FLEX and digital equipment) in enclosure 9, tables E9-1, E9-3 and E9-4 that are based on review of regulatory correspondence to other nuclear utilities; TVA states they were not determined to be key sources of uncertainty for WBN. No other assessment of candidate assumptions or sources of uncertainty was provided in the LAR, and the LAR did not identify any sensitivity studies to support its conclusions or identify any Risk Management Actions (RMAs) needed for LCO
[limiting conditions for operation] conditions that could be impacted by modeling uncertainty.
The word screened is not used in LAR enclosure 9, but it appears that a master compilation of plant-specific and generic industry PRA modeling assumptions and sources of uncertainty was screened using a set of criteria to determine that none of the applicable PRA modeling assumptions and sources of uncertainty are key to this application. It is not clear to NRC staff what evaluation criteria were used to consistently evaluate plant-specific and generic sources of uncertainty to conclude that none are key for this application. Therefore, address the following:
- a. Describe and justify the criteria used to consistently evaluate a comprehensive list of internal events (including internal floods), fire, and seismic PRA modeling assumptions and sources uncertainty (including those associated with plant-specific features, modeling choices, and generic industry concerns) to conclude none are key to the WBN RICT program.
- b. Discuss and provide the results of sensitivity studies (if any) that were performed to evaluate an identified assumption or source of uncertainty for its impact on the RICT calculations.
- c. Discuss additional RMAs (if any) that will be used to address sources of PRA modeling uncertainty: (1) describe how these RMAs will be identified prior to the implementation of the RMTS program, consistent with the guidance in section 2.3.4 of NEI 06-09; and (2) provide RMA examples that may be considered during a RICT program entry to minimize any potential adverse impact from this uncertainty and explain how these RMAs are expected to reduce the risk associated with this uncertainty.
QUESTION 02 - Consideration of Shared Systems in RICT Calculations Regulatory Guide (RG) 1.200, Revision 2, states, [t]he base PRA serves as the foundational representation of the as-built and as-operated plant necessary to support an application.
Table 1 of LAR enclosure 8 indicates the existence of cross-ties between units and identifies several systems that are shared. It is not clear to NRC staff how these systems are shared, whether they can support both units in an accident, and how the shared systems are credited for each unit in the PRA models. NRC staff notes that for certain events, such as dual unit events (e.g., loss of offsite power), it may be appropriate to only credit the shared systems for one unit. Therefore, address the following:
- a. Explain how shared systems are modeled in the Real Time Risk (RTR) model for each unit in a dual unit event demonstrating that shared systems are not over-credited in the RTR model.
- b. If the RTR model does not address the impact of events that can create a concurrent demand for the system shared by both units, then justify that this exclusion has an inconsequential impact the RICT calculations.
QUESTION 03 - Impact of Seasonal Variations The Tier 3 assessment in RG 1.177 [Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications] stipulates that a licensee should develop a program that ensures that the risk impact of outofservice equipment is appropriately evaluated prior to performing any maintenance activity. NEI 06-09-A and its associated NRC SE state that, for the impact of seasonal changes, either conservative assumptions should be made or the PRA should be adjusted appropriately to reflect the current (e.g., seasonal or time of cycle) configuration.
LAR enclosure 8, section 3, on attributes of the RTR model, states, There are no seasonal variations included in the WBN OTMHM [one top multi hazard model]. It is not clear to NRC staff why no mechanism or criteria are used to determine when PRA adjustments need to be made in the RTR due to changes in Tennessee River water temperature, nor seasonal or time of cycle dependencies. Therefore, address the following to clarify the treatment of seasonal and time of cycle variations:
- a. Discuss any PRA modeling adjustments made in the RTR to account for seasonal and time of cycle variations during a RICT evolution. Also, explain how these adjustments are made and how this approach is consistent with the guidance in NEI 0609-A and its associated NRC SE.
- b. Describe the criteria used to determine when PRA adjustments due to seasonal or time of cycle variations need to be made in the RTR and what mechanism initiates these changes.
- c. Describe the number of emergency raw cooling water (ERCW) pumps that are required due to river water temperature throughout the year; and discuss the impact of the ERCW header temperature on the number of component cooling water (CCW) heat exchangers required to mitigate modeled events. Describe how the RTR conservatively models the success criteria of ERCW pumps and CCW heat exchangers for all seasons in order to have no seasonal variations included in the model.
QUESTION 04 - In-Scope LCOs and Corresponding PRA Modeling The NRC SE for NEI 06-09-A specifies that the LAR should provide a comparison of the TS functions to the PRA modeled functions to show that the PRA modeling is consistent with the licensing basis assumptions or to provide a basis for when there is a difference. LAR enclosure 1, table E1-1 identifies each TS LCO Condition proposed to be in the RICT program, describes whether the systems and components participating in the TS LCO are modeled in the PRA, and compares the design basis and PRA success criteria. For certain TS LCO Conditions, the table explains that the associated SSCs [structures, systems and components] are not explicitly modeled in the PRAs but their unavailability will be represented using a surrogate event that fails the function performed by the SSC. For one LCO Condition, the LAR did not provide enough description for NRC staff to conclude that the PRA modeling will be sufficient for the proposed LCO Condition. Therefore, address the following:
- a. LAR table E1-1 states for TS LCO 3.7.5 (Auxiliary Feedwater System) Condition B (One AFW [auxiliary feedwater] train inoperable in MODE 1, 2, or 3 for reasons other than Condition A) that the design basis success criterion is Two of three AFW pumps and the PRA success criterion is One of three AFW Pumps. The comment column for this entry states that limited accident scenarios, such as specific ATWS transients, do require two AFW pumps. It is not clear to NRC staff why the PRA success criteria is not conservatively modeled like the design success criteria to include all accident scenarios.
Therefore, address the following:
- i.
Describe all accident scenarios that require two AFW pumps and justify why the PRA success criteria of one of three AFW pumps is adequate for these scenarios with respect to this application.
ii. Describe any adjustments of AFW success criteria made to the RTR model for emergent conditions similar to accident scenarios provided in response to (i) OR provide a justification for why no changes are made to the RTR model for emergent conditions similar to accident scenarios provided in response to (i).
QUESTION 05 - Modeling of the Reactor Coolant Pump (RCP) Shutdown Seal The PRA model for the Generation Ill Seals was approved by the NRC in the final SE of TR PWROG-14001-P, Revision 1, PRA Model for the Generation Ill Westinghouse Shutdown Seal, dated August 23, 2017 (ML17200C875).
Consistent with the guidance in RG 1.174, Revision 3, [An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis]
that the PRA scope, level of detail and technical acceptability be based on the as-built and as-operated and maintained plant, and reflect operating experience at the plant, address the following:
- a. Clarify what kind of seals are installed in each RCP in WBN, Units 1 and 2, and whether the current PRA models include credit for the Westinghouse Generation Ill (SHIELD)
RCP seals.
- b. If any PRA model include credit for the Westinghouse Generation Ill RCP seals, address the following:
- i.
Confirm that the limitations and conditions in the NRC SE for PWROG-14001-P, Revision 1, are met.
ii. if exceptions to the limitations and conditions exist, identify all the exceptions and justify their impact on the application.
iii. Clarify whether the Generation Ill Westinghouse RCP seal model has been peer-reviewed as part of the internal events PRA and fire PRA (FPRA) peer-reviews.
iv. If this RCP seal model has not been peer reviewed, justify why the addition of this model is not considered a PRA upgrade requiring a focused-scope peer review.
- v. If the addition of RCP seal model qualifies as a PRA upgrade, provide the results from the focused-scope peer review including the associated Facts and Observations (F&Os) and their resolutions.
Probabilistic Risk Assessment Licensing Branch B (APLB) Audit Questions QUESTION 06 - FPRA Assumption: Control Building Fire Scenarios For fires in the Control Building, MCR [main control room] abandonment appears to be the criteria for a fire to be considered an Appendix R fire and thus included in the FPRA. In the rest of the plant, a fire-induced plant trip (manual or automatic) appears to be sufficient to indicate a fire is an Appendix R fire, but a fire-induced plant trip is not a MCR abandonment criteria for a Control Building fire. Therefore, such scenarios would not be included in the FPRA.
Can control building fire scenarios cause a plant trip, and if so, shouldnt those fire scenarios be included in the FPRA?
QUESTION 07 - FPRA Assumption: Model Asymmetry From table 8-14, Fire PRA Report Assumption Evaluation, in the FPRA Quantification Notebook:
Assumption: The Internal Events PRA has a modeling asymmetry. It assumes that Charging Pump A is initially operating at t=0. This asymmetry is propagated to the Fire PRA. Therefore, HACHG_F, which models the failure to ensure one charging pump is operating, is assigned to the startup of Charging Pump B.
This assumption may introduce a bias in risk-informed applications (such as 50.69 and risk-informed completion times). [emphasis added]
Source of Uncertainty Impact Evaluation: For the evaluation of fire risk in the baseline Fire PRA, the assumption is not a significant source of uncertainty and does not require a sensitivity study. But in future fire risk applications, the assumption would require further evaluation to ensure it does not introduce an undue bias. [emphasis added]
However, it does not appear that this is addressed in the LAR.
Describe the evaluation that was performed regarding this asymmetry, and its effect on the RICT program.
QUESTION 08 - Discussion Topic: Fire Hazard in Baseline Core Damage Frequency (CDF)
The staff would like to discuss the prominence of the fire hazard in the WBN baseline CDF values (>80 percent for each unit) and see if TVA has any insights into why this is the case.
Related to this would be the relative robustness of containment for the fire scenarios, which appear to lead to a large early release frequency at around one third as often as internal events scenarios.
Probabilistic Risk Assessment Licensing Branch C (APLC) Audit Questions QUESTION 09 - Seismic PRA Model Self-Assessment In section 3 of enclosure 2 of the LAR, TVA states, in response to Nuclear Regulatory Commission (NRC) concerns due to the lack of a TVA performed self-assessment of whether each F&O resolution was a PRA Upgrade or PRA Update, a re-review was performed during October 2019. To support this re-review, TVA prepared a self-assessment of the upgrade/update status of each F&O resolution to be used as input by the review team.
Explain whether any changes from the self-assessment re-review for the internal events PRA were incorporated into the seismic PRA model used for the RICT program.
QUESTION 10 - Seismic PRA Model Peer Review In section 5 of enclosure 2 of the LAR, TVA states, The March 2016 WBN Units 1 and 2 Seismic Probabilistic Risk Assessment (SPRA) was peer reviewed against the requirements of ASME/ANS RA-Sa-2009, the American Society of Mechanical Engineers (ASME)/American Nuclear Society (ANS) Probabilistic Risk Assessment (PRA) Standard.
In section 3.2.3 of enclosure 1 of the previously approved WBN application to adopt 10 CFR 50.69, TVA states, ASME/ANS PRA Standard ASME/ANS RA-Sb-2013 [] is credited for the peer review performed on the WBN SPRA.
Please address the following:
- b. If the last full-scope peer review of SPRA was conducted against the SPRA requirements in ASME/ANS RA-Sb-2013, justify that the use of this alternative approach to the NRC-endorsed approach addresses the technical elements for the development of a SPRA.
QUESTION 11 - Extreme Winds and Tornado Hazards Screening Provide specific evaluations or supporting documentation that substantiate the screening of extreme winds and tornado hazards from the RICT evaluations.
QUESTION 12 - External Flooding Hazards Screening Provide specific evaluations or supporting documentation that substantiate the screening of external flooding hazards from RICT evaluations.
Electrical Engineering Branch (EEEB) Audit Questions General Design Criteria (GDC) 17 requires, in part, that both offsite and onsite electrical power systems should be provided to permit the functioning of systems, structures, and components (SSCs) important to safety. The safety function for each system, assuming other is not functioning, assures fuel design limits and design conditions of the reactor coolant pressure boundary are not exceeded, and the core is cooled, and containment integrity and other vital functions are maintained for postulated accidents.
QUESTION 13 - Electrical Discussion Topic
, table E1-1, shows the same number of 14 TS changes as in attachments 2.1 and 2.2 including their individual wording for TS Conditions identified except for the following:
- i.
TS 3.8.9, Condition A for Unit 2 is not shown, only the one for same TS and Condition for Unit 1 ii.
TS 3.8.9, Condition F for Unit 2 is not shown, only the one for same TS and Condition for Unit 1
, table E1-2, shows the same 14 TS changes as in attachment 2.1 including their individual wording.
, table E1-3, shows the same 14 TS changes as in Attachment 2.2 including their individual wording except for the following:
- i.
TS 3.8.9.A should be for Unit 2 AC [alternating current] shutdown boards ii.
TS 3.8.9.F should be for Unit 1 AC shutdown boards QUESTION 14 - TS LCO 3.8.1, Condition B, C, E, and F LCO 3.8.1.b requires four diesel generators (DGs) capable of supplying the onsite Class 1E AC electrical power distribution system to be operable. TS 3.8.1, Condition B, is for one required DG inoperable. The Required Action for that LCO Condition requires restoration of the DG.
Updated Final Safety Analysis Report (UFSAR) section 8.3.1 indicates that the onsite AC power system which includes the standby power system serving each unit is divided into two redundant load groups power trains. These power trains (Train A and Train B for each unit) provide power to all safety-related equipment. Safety-related loads are arranged electrically into four power trains. Trains 1A and 2A comprise load group A, and Trains 1B and 2B comprise load group B. Trains 1A and 1B are in Unit 1, and Trains 2A and 2B are in Unit 2. Additionally, TS Bases B3.8.1, Background, states that the onsite power system consists of four power trains shared between the two units and the load groups.
Design success criteria (DSC) in table E1-1 for TS LCO 3.8.1, Condition B appears inconsistent with column two of table E1-1 for the same LCO condition.
DSC indicates success is two DGs of the same load group but column two of table E1-1 for TS 3.8.1, Condition B refers to trains not to load groups. Please clarify the inconsistency by possibly noting the composition of a load group (e.g., DGs for train A from both Units 1 and 2 or for Train B from both Units 1 and 2.). This is also applicable to TS 3.8.1, Conditions C, E, and F in table E1-1.
QUESTION 15 - TS LCO 3.8.1, Condition E LCO 3.8.1.b requires four diesel generators (DGs) capable of supplying the onsite Class 1E
[alternating current] AC Electrical Power Distribution System to be operable. TS 3.8.1, Condition B, is for one required DG inoperable. The Required Action for that LCO condition requires restoration of DG.
UFSAR section 8.3.1 indicates onsite power system for either unit consists of the (1) standby AC power system, and (2) the 120 volts (V) vital AC system. The standby power ac system meets the requirements of GDC 17 for Class 1E onsite ac power system. The standby power system serving each unit is divided into two redundant load groups power trains. These power trains (Train A and Train B for each unit) provide power to all safety related equipment. The UFSAR further states that the four 6.9 kilovolt (kV) shutdown boards are arranged into four power trains (two per unit). Each 6.9 kV shutdown board is one train. UFSAR section 8.1.5.3 states that a single failure (loss of battery or loss of a DG) in the plant and an assumed loss of offsite power, sufficient engineered safety features (ESF) loads are still automatically available to the accident unit and to safely shut down the remaining unit. The shared safety systems are designed so that one load group (Train 1A & 2A or Train 1B & 2B) can mitigate a design-basis accident in one unit and accomplish an orderly shutdown of the other unit.
DSC in table E1-1 for TS LCO 3.8.1, Condition E appears inconsistent with the purpose of DSC of providing minimum SSCs for safe shutdown.
DSC indicates success is one qualified circuit between transmission network and the onsite 1E AC electrical power distribution systems or two DGs of a load group. The worst case is DGs of a load group available for safe shutdown with offsite power unavailable. Please clarify or explain why DSC, which should be the minimum SSCs to achieve a safe shutdown, does not address the worst case.
QUESTION 16 - TS LCO 3.8.4, Conditions A and B LCO 3.8.4 requires Train A and Train B vital direct current (DC) electrical power subsystems to be operable. UFSAR section 8.3.2.1.1 indicates that 125 VDC vital power system has four redundant channels (designated as Channels I, II, III, and IV) and consists of four lead-acid-calcium batteries, eight battery chargers (including two pairs of spare chargers), four distribution boards, battery racks, and the required cabling, instrumentation and protective features. A channel has a battery charger to supply normal DC power, a battery for emergency DC power, and a battery board which facilitates load grouping and provides circuit protection. The UFSAR also indicates that each channel supplies the control circuits for a shutdown board. USFAR section 8.1.5.3 indicates that the Class1E DC power system has four redundant divisions. TS Bases B3.8.4, DC Sources - Operating - Background, indicates that the 125 VDC vital power system is composed of the four redundant channels (Channels I and III are associated with Train A and Channels II and IV are associated with Train B) and channel composition.
DSC appears inconsistent with the LCO, the UFSAR and TS Bases, and column 2 of TS 3.8.4, Condition A or B which refer to trains. Please clarify or explain the inconsistencies by DSC stating the composition of a Power Train (e.g., Train A of Unit 1 and Train A of Unit 2 as a Power Train).
QUESTION 17 - TS LCO 3.8.4, Conditions D and E USFAR section 8.3.1.1 states that there is a DG battery subsystem for each DG. Each subsystem is comprised of a battery, battery charger, distribution center, and cabling. The battery provides control and field-flash power when the charger is unavailable. The charger, if 480 VAC is available, supplies the normal DC loads, maintains the battery in a fully charged condition, and recharges the battery while supplying the required loads regardless of the status of the plant. The batteries are physically and electrically independent. DSC appears inconsistent with the LCO which differentiates between vital DC and DG DC electrical power subsystems.
Please clarify or explain why table E1-1, column 2 and column 5, for TS 3.8.4 Conditions D and E apply to DC electrical power trains, but the LCO is for DG DC trains or subsystems.
QUESTION 18 - TS LCO 3.8.7, Condition A UFSAR section 8.3.1.1 indicates 125 VDC vital system supplies power for the 120 VAC system with there being four uninterruptible power systems (UPS) for ESF loads and four UPS for reactor protection system (RPS) loads. The UFSAR also indicates that the 120 V vital AC system consists of four identical power channels per unit (designated as Channels I, II, III and IV). Each channel for both units consists of an inverter and a distribution panel which facilitates load grouping and provides circuit protection.
DSC appears inconsistent with the LCO 3.8.7 which addresses inverters, and the DSC only requires ESF and RPS power divisions for safe shutdown. Please clarify or explain the inconsistency by, for instance, identifying number of inverters for those ESF and RPS power divisions.
Instrumentation and Controls Branch (EICB) Audit Questions QUESTION 19 - Defense-In-Depth Evaluation for Protective Functions TSTF-505, Revision 2 (ML18183A493), states:
The description of proposed changes to the protective instrumentation and control features in TS Section 3.3, Instrumentation, should confirm that at least one redundant or diverse means (other automatic features or manual action) to accomplish the safety functions (for example, reactor trip, SI [safety injection],
containment isolation, etc.) remains available during use of the RICT, consistent with the defense-in-depth philosophy as specified in RG 1.174 (Note that for each application, the staff may selectively audit the licensing basis of the most risk-significant functions with proposed RICTs to verify that such diverse means exist).
The LAR includes an evaluation to address the above; however, this evaluation is missing the table (typically supplied in TSTF-505 LARs) which systematically demonstrates, for each event in UFSAR Chapter 15 (that is mitigated by I&C [instrumentation and control] subject to RICT),
the credited means and the diverse means to initiate the protective function. Please provide this table for discussion during the audit.
Technical Specifications Branch (STSB) Audit Questions QUESTION 20 - Technical Specification 5.7.2.24 In attachment 4, cross reference table for TSTF-505, for the RICT program, page A4-10 of 10, it identifies it as Section 5.5.20 in the column of the table for WBN TS/RA. NRC staff notes that the RICT program for WBN is TS 5.7.2.24. NRC staff also notes that the Programs and Manuals heading for WBN is listed as TS Section 5.5 in the table, when it is TS Section 5.7. Confirm and correct.
QUESTION 21 - Technical Specification 3.8.9 In attachment 5, cross reference table for TSTF-439, page A5-2 of 2, for TS 3.8.9 RA A.1, staff notes in the variation reference column that it refers to The WBN Condition B instead of The WBN Condition A. Confirm and correct, if needed.
QUESTION 22 - Technical Specification 3.6.8 In table E1-1, for 3.6.8 Condition B, for one containment region with no OPERABLE hydrogen ignitor, the DSC states one region without an operable ignitor. Should it say, one region with an operable ignitor?
QUESTION 23 - Technical Specification Bases B.1 In attachment 3, bases markup, B 3.3.2.B.1, B.2.1, and B.2.2 on page B 3.3-84, it shows removing actions B.2.1 and B.2.2 but the heading doesnt reflect this removal. Consider revising the heading to match the removal of these actions and reflecting only B.1 in the heading.
List of Documents Audited During Audit The following supporting documents (e.g., analyses, calculations, reports, drawings, and procedures) were provided by TVA on the internet-based document portal available during the audit period.
Application-Specific Documents License Amendment Request to Revise Technical Specifications to Adopt Risk-informed Completion Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (ML24362A110)
Supplement to License Amendment Request to Revise Technical Specifications to Adopt Risk-informed Completions Times TSTF-505, Revision 2, Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b (ML25232A156)
Technical Documents Calculation NDN0000002022000002, Revision 1, WBN Fire PRA - Human Reliability Analysis, dated June 6, 2024 WBN-0-24-054, May 21, 2024, PWROG-23027-P, Revision 0, Peer Review of the Watts Bar Units 1 & 2 Fire Probabilistic Risk Assessment, dated December 2023 WBN-0-24-056, May 21, 2024, Watts Bar Nuclear Plant, Focused Scope Per Review -
High Energy Arcing Fault, NUREG-2262 Methodology, dated May 17, 2024 WBN-0-23-101, July 26, 2023, contains evaluation of closure of all Finding Level Facts and Observations associated with WBN Internal Events and Internal Flooding PRA Westinghouse Letter from David E. McCoy to Bradford Grimmel, RG 1.200 PRA Peer Review Against the ASEM/ANS PRA Standard Requirements for the Watts Bar Nuclear Plant Probabilistic Risk Assessment, dated December 17, 2009 Report 06044-RPT-01, Revision R1, Watts Bar Nuclear Plant Seismic PRA Finding Level F&O Independent Technical Review, dated July 6, 2017 PWROG-16011-P, Revision 0, Peer Review of the Watts Bar Seismic Probabilistic Risk Assessment, dated June 2016 Report No. 1NWL06127-RPT-01, Watts Bar Nuclear Plant PRA Finding Level Fact and Observation Independent Assessment, dated April 4, 2025 Calculation MDN00099920080137, Revision 7, WBN Probabilistic Risk Assessment -
Electric Power System, dated July 6, 2023 Calculation MDN-000-001-2008-0122, Revision 2, WBN Probabilistic Risk Assessment
- Main Steam System, dated July 8, 2015 Calculation MDN-000-003-2008-0124, Revision 5, WBN Probabilistic Risk Assessment
- Auxiliary Feedwater System, dated August 26, 2015 Letter from Farideh E. Saba, NRC, to Joseph W. Shea, TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3; Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Units 1 and 2 - Issuance of Amendment Nos. 309, 332, 292, 345, 339, 128, and 31 Regarding Unbalanced Voltage Protection, dated August 27, 2019 (ML18277A110)
Letter from Michael J. Wentzel, NRC, to James Barstow, TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3; Sequoyah Nuclear Plant, Units 1 and 2; Watts Bar Nuclear Plant, Units 1 and 2 - Closeout of Bulletin 2012-01, Design Vulnerability in Electric Power System, dated May 1, 2020 Response regarding Open Phase Condition Calculation MDN-000-024-2011-0176, Revision 3, WBN Probabilistic Risk Assessment
- Raw Cooling Water, dated February 8, 2021 Calculation MDN-000-032-2008-0126, Revision 2 - WBN Probabilistic Risk Assessment
- Plant Compressed Air, dated July 8, 2015 Calculation MDN-000-067-2008-0130, Revision 6 - WBN Probabilistic Risk Assessment
- Essential Raw Cooling Water System, dated August 11, 2015 Calculation MDN-000-070-2008-0132, Revision 3 - WBN Probabilistic Risk Assessment
- Component Cooling System, dated September 20, 2016 Calculation MDN-000-999-2008-0137, Revision 5, WBN Probabilistic Risk Assessment
- Electric Power System, dated March 27, 2017 Calculation NDN0000002022000011, Revision 2, WBN Fire PRA - Fire Risk Quantification, dated June 6, 2024 Calculation MDN0009992021001756, Revision 2, Watts Bar Fire PRA Plant Boundary and Partitions, dated June 6, 2024 Calculation MDN0009992015000717, Revision 2, WBN Probabilistic Risk Assessment
- Seismic Quantification (SQU) Notebook, dated May 22, 2017 NEDP-26, Revision 14, Probabilistic Risk Assessment, dated February 6, 2023 NPG-SPP-09.11, Revision 6, Probabilistic Risk Assessment Program, dated August 7, 2023 Calculation 06132-RPT-01, Revision 000, WBN RICT Estimates for TSTF-505 (RICT)
Program LAR Submittal, dated October 9, 2024 WBN-0-24-066, Acceptance of One Top Multi-Hazard Model (OTMHM), dated August 29, 2024 NPG-SPP-09.11.4, Revision 2, Risk-Managed Technical Specifications Program, dated September 24, 2024 NPG-SPP-09.11.5, Revision 000, Risk Managed Tech Specs Program Cumulative Risk, dated August 24, 2023 Calculation MDN-000-999-2009-0162, Revision 4, WBN Probabilistic Risk Assessment
- Sensitivity and Uncertainty Notebook, dated October 7, 2021 Report WBN-0-24-094, Plant Evaluation Response, dated October 15, 2024 Report 1NWL06127-RPT-01, PRA Finding Level Fact and Observation Independent Assessment, dated April 4, 2025
ML25268A010 NRR-106 OFFICE NRR/DORL/LPL2-2/PM NRR/DORL/LPL2-2/LA NRR/DRA/APLA/BC NRR/DRA/APLB/BC NAME KGreen ABaxter BPascarelli EDavidson DATE 9/26/2025 9/30/2025 11/13/2025 10/21/2025 OFFICE NRR/DRA/APLC/BC(A) NRR/DEX/EEEB/BC NRR/DEX/EICB/BC(A)
NRR/DSS/STSB/BC NAME SAlferink WMorton (SRay for)
SDarbali SMehta DATE 9/30/2025 11/19/2025 11/14/2025 11/13/2025 OFFICE NRR/DORL/LPL2-2/BC NRR/DORL/LPL2-2/PM NAME DWrona KGreen DATE 11/21/2025 11/21/2025