IR 05000390/2023003

Integrated Inspection Report 05000390/2023003 and 05000391/2023003 and Apparent Violation

ML23313A002
Person / Time
Site:	Watts Bar
Issue date:	11/13/2023
From:	Louis Mckown Division Reactor Projects II
To:	Jim Barstow Tennessee Valley Authority
References
EA-23-129, EA-23-130, EA-23-131 IR 2023003
Download: ML23313A002 (32)

Text

November 13, 2023

SUBJECT:

WATTS BAR NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000390/2023003 AND 05000391/2023003 AND APPARENT VIOLATIONS

Dear Jim Barstow:

On September 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Watts Bar Nuclear Plant. On October 18, 2023, the NRC inspectors discussed the results of this inspection with Anthony Williams, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Section 71111.24 of the enclosed report documents a finding with an associated apparent violation for which the NRC has not yet reached a preliminary significance determination. This involved an inspector identified apparent violation (AV) of Technical Specifications Surveillance Requirement 3.6.12.2. associated with the failure to ensure that measuring and testing equipment provided reasonable assurance that required conditions were met.

Section 71152A of the enclosed report documents a finding with an associated apparent violation for which the NRC has not yet reached a preliminary significance determination. This involved a self-revealed AV of 10 CFR 50.155. associated with the licensees failure to maintain the equipment necessary to support deployment high pressure FLEX pumps.

Section 71153 of the enclosed report documents a finding with an associated apparent violation for which the NRC has not yet reached a preliminary significance determination. This involved a licensee identified AV of Technical Specification 5.7.1.1.d. associated with the licensees failure to implement the requirements Watts Bar Fire Protection Program, including, that each Emergency Diesel Generator (EDG) and its associated equipment are separated from each other by 3-hour fire barriers.

The NRCs significance determination process (SDP) is designed to encourage an open dialogue between your staff and the NRC; however, neither the dialogue nor the written information you provide should affect the timeliness of our final determination. We ask that you promptly provide any relevant information that you would like us to consider in making our determination. We are currently evaluating the significance of this finding and will notify you in a separate correspondence once we have completed our preliminary significance review. You will be given an additional opportunity to provide additional information prior to our final significance determination unless our review concludes that the finding has very low safety significance (i.e.,

Green).

Three findings of very low safety significance (Green) are documented in this report. Three of these findings involved violations of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.

If you contest the NCVs or the significance or severity of the NCVs documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:

Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Watts Bar Nuclear Plant.

If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Watts Bar Nuclear Plant.

This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.

Sincerely, Signed by McKown, Louis on 11/13/23 Louis J. McKown, II, Chief Reactor Projects Branch #5 Division of Reactor Projects Docket Nos. 05000390 and 05000391 License Nos. NPF-90 and NPF-96

Enclosure:

As stated

Inspection Report

Docket Numbers: 05000390 and 05000391 License Numbers: NPF-90 and NPF-96 Report Numbers: 05000390/2023003 and 05000391/2023003 Enterprise Identifier: I-2023-003-0030 Licensee: Tennessee Valley Authority Facility: Watts Bar Nuclear Plant Location: Spring City, TN 37381 Inspection Dates: July 01, 2023, to September 30, 2023 Inspectors: W. Deschaine, Senior Resident Inspector P. Gresh, Operations Engineer J. Griffis, Senior Health Physicist B. Kellner, Senior Health Physicist A. Nielsen, Senior Health Physicist D. Simpkins, Sr. Tech Training Program Specialist N. Smalley, Resident Inspector J. Tornow, Physical Security Inspector J. Walker, Sr Emergency Preparedness Inspector R. Wehrmann, Resident Inspector Approved By: Louis J. McKown, II, Chief Reactor Projects Branch #5 Division of Reactor Projects Enclosure

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Watts Bar Nuclear Plant, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.

List of Findings and Violations

Failure to maintain records of activities affecting Quality Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.13] - 71111.12 Systems NCV 05000390,05000391/2023003-01 Consistent Open/Closed Process The inspector identified a Green finding and associated Non-Cited Violation of 10 CFR Part 50 Appendix B, Criterion XVII, Quality Assurance Records for failure to maintain QA records.

Specifically, 354 safety related items were installed in the plant did not have the QA records establishing traceability.

Failure to implement the requirements of 1-SI-61-7, 18-Month Ice Condenser Intermediate Deck Doors Operational Check Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.12] - Avoid 71111.15 NCV 05000390/2023003-02 Complacency Open/Closed The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of Technical Specification 5.7.1.1a when the licensee failed to perform maintenance in accordance with written procedures. Specifically, the licensee failed to perform testing of the Unit 1 ice condenser intermediate deck doors in accordance with procedure 1-SI-61-7 when they used a force gauge unsuitable for the ambient temperature and without verifying the gauge was within calibration immediately after use. As a result, 8 intermediate deck doors were determined to be inoperable when discovered on September 21, 2023.

Failure to ensure bounding technical requirements were incorporated into a Procedurally Controlled Temporary Modification Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.12] - Avoid 71111.18 Systems NCV 05000391/2023003-03 Complacency Open/Closed The inspectors identified a Green finding and associated Non-Cited Violation of 10 CFR 50 Appendix B Criterion III for the failure of the licensee to ensure that bounding technical requirements were met prior to installation of a temporary modification. Specifically, the licensee failed to ensure that adequate technical requirements were incorporated into Procedurally Controlled Temporary Modification, 0-TI-66.004 Revision 001, prior to issuance.

Failure to obtain NRC approval prior to implementing a Technical Specification (TS) change for Watts Bar Unit 1 Mode 5 requirement Cornerstone Severity Cross-Cutting Report Aspect Section Not Applicable Severity Level IV Not Applicable 71111.18 NCV 05000390/2023003-04 Open/Closed An NRC-identified a Severity Level IV Non-Cited Violation of 10 CFR Part 50.59(c)(1) for failure to obtain NRC approval prior to implementing a Technical Specification (TS) change for Watts Bar Unit 1 Mode 5 requirement. Specifically, the licensee failed to perform adequate 50.59 Evaluations/Screenings for the proposed TS change requirement for Mode 5 with one reactor vessel closure bolt out of service from 2020 to 2023 for Watts Bar Unit 1 without NRC approval wherein Note b of TS Table 1.1-1, Modes is defined by TS for Mode 5 as All reactor vessel head closure bolts fully tensioned.

Failure to implement Surveillance Requirement 3.6.12.2, Ice Condenser Doors Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Pending [H.11] - 71111.24 AV 05000391/2023003-05 Challenge the Open Unknown EA-23-130 An NRC-identified finding with its safety significance as yet to be determined (TBD) and Apparent Violation (AV) of Technical Specification 3.6.12, Ice Condenser Doors, was identified for the licensees failure to implement the requirements of surveillance instruction 2-SI-61-6, Unit 2 Weekly Ice Condenser Intermediate Deck Doors Visual Inspection.

Specifically, the licensee failed to ensure that adequate Measuring and Test Equipment (M&TE) was utilized to provide reasonable assurance that the Intermediate Deck Doors remained operable.

Failure to Maintain N High Pressure Flex Capability Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Pending [H.4] - 71152A Systems AV 05000390,05000391/2023003-06 Teamwork Open EA-23-131 A self-revealed finding with its safety significance as yet to be determined (TBD) and Apparent Violation (AV) of 10 CFR 50.155, Mitigation of Beyond Design Basis Events, was identified for the licensees failure to implement the requirements of technical instruction 0-TI-446, Diverse and Flexible Coping Strategies (FLEX) Program Bases. Specifically, Watts Bar did not maintain the equipment (hose couplings) necessary to support the deployment of the High Pressure (HP) FLEX pumps as needed to support Phase 2 mitigation strategy defined in 0-TI-446.

Licensee identified Unanalyzed Condition for the 2A-A Emergency Diesel Generator Cornerstone Significance/Severity Cross-Cutting Report Aspect Section Mitigating Pending None (NPP) 71153 Systems Apparent Violation

AV 05000390,05000391/2023003-07 Open EA-23-129 A licensee-identified finding with its safety significance as yet to be determined (TBD) and Apparent Violation (AV) of Technical Specification (TS) 5.7.1.1 d, Fire Protection Program implementation, was identified for the licensees failure to implement the requirements Watts Bar Fire Protection Program, including, that each Emergency Diesel Generator (EDG) and its associated equipment are separated from each other by 3-hour fire barriers. Specifically, the alternate feeder common to the 480V Shutdown Boards 2A1-A and 2A2-A lacked the required fire barrier.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000391/2023-001-00 LER 2023-001-00 for Watts 71153 Closed Bar Nuclear Plant, Unit 2,

Unanalyzed Condition Related to Loss of the 2A Emergency Diesel Generator During a Postulated Appendix R Fire

PLANT STATUS

Unit 1 operated at or near rated thermal power for the entire inspection period.

Unit 2 began the inspection period at or near rated thermal power. Unit 2 experienced a malfunction of the #2 Main Feedwater Regulated Valve (MFRV) on August 4, 2023. This malfunction caused the MFRV to close which caused an automatic trip of the reactor. The unit was returned to rated thermal power on August 6, 2023, and remained at or near rated thermal power for the remainder of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

71111.01 - Adverse Weather Protection

Impending Severe Weather Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather, severe thunderstorms, and tornado watch, on August 7, 2023.

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (3 Samples)

The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:

(1) Unit 2 Safety Injection (SI) system on September 13, 2023

(2) Unit 2 Emergency Diesel Generator (EDG) system on September 14, 2023.

(3) Emergency Gas Treatment system (EGTS) on September 28, 2023.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:

(1) 480V Board Rooms (Auxiliary Building Elevation 772') on July 12, 2023

(2) Diesel Generator Building (Elevation 742' and 760.5') on July 21, 2023

(3) Mechanical Equipment and 480V Transformer Rooms (Auxiliary Building Elevation 772') on July 27, 2023

(4) Control building, elevation 708', Unit 1 and 2 auxiliary instrument rooms and computer room on August 30, 2023

(5) Aux Building Corridor, elevation 692' on September 12, 2023

Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the onsite fire brigade training and performance during an announced fire drill on September 20, 2023.

71111.06 - Flood Protection Measures

Flooding Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated internal flooding mitigation protections in the Unit 2 safety injection pump rooms on September 28, 2023.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)

(1 Sample)

(1) The inspectors observed and evaluated licensed operator performance in the control room during reactor startup on August 05, 2023.

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (2 Samples)

(1) The inspectors observed and evaluated Licensed Operator Requalification Training for Crew 5A and Crew 5B. The scenario involved isolation of an intermediate feedwater heater string, dropped control bank rod, and large break loss-of-coolant accident (LOCA) on September 12, 2023.

(2) The inspectors observed and evaluated Licensed Operator Requalification Training for the following:

• Crew 5B - The scenario involved failure of pressurizer pressure instrument, trip of the B main feed pump, and steam generator tube rupture on September 13, 2023.

• Crew 3A - The scenario involved failure of steam generator pressure channel low, trip of a condenser circulating water pump, a reactor trip with a failed EDG and a faulted steam generator on September 19, 2023.

These two observations count as a single sample.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (2 Samples)

The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:

(1) 1-RM-90-130 Containment Purge Air Exhaust Radiation Monitor failure and repair under CR1877031 and WO123958788 on August 27, 2023

(2) Repair of 1-FCO-030-0248A under CR1870666 due to discovery of failure on July 26, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (9 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) Elevated Emergency Diesel Generator Exhaust Fan Vibrations under CR1870179 and CR1711241

(2) Containment Building Lower Compartment Radiation Monitor Operability Evaluation under CR1873810

(3) 1B2 EDG Engine Jacket Water Temperature Switch Operability evaluation under CR1873989 on August 29, 2023

(4) 2B-B EDG Engine 2B1-B Low Lube Oil Level Operability evaluation under CR1875047 on August 28, 2023

(5) 1A-A ERCW Screen Wash Pump Failure Operability evaluation under CR1875059 on August 28, 2023

(6) 1B-B Centrifugal Charging Pump Failure to maintain Pressurizer level Operability evaluation under CR1877508 on August 30, 2023

(7) Unit 1 Intermediate Deck Doors under CR1881663 on September 20, 2023, through September 22, 2023

(8) 2B-B EDG Engine 2B2-B Low Lube Oil Level Operability evaluation under CR1882102 on September 26, 2023

(9) 1B-B EDG Bearing/Stator Temperature Recorder 1-ITR-82-1032 Operability evaluation under CR1882104 on September 29, 2023

71111.18 - Plant Modifications

Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02)

(2 Samples)

The inspectors evaluated the following temporary or permanent modifications:

(1) WBN-2-2022-067-01, Removal of Upper Compartment Coolers 2A & 2D, Rev. 0

(2) WBN-20-1371, Unit 1 Reactor Vessel Head Closure Bolts, Rev. 0, 1, & 2

71111.24 - Testing and Maintenance of Equipment Important to Risk

The inspectors evaluated the following testing and maintenance activities to verify system

operability and/or functionality: Post-Maintenance Testing (PMT) (IP Section 03.01)

(1) PMT of SR and IR NIs, on August 5, 2023, under WO122942390

(2) 1B-B EBR Chiller, under WO123119720 on August 30, 2023

(3) A-A Train Emergency Gas Treatment System under WOs 123119769, 123019529, and 122055927

(4) Unit 1 Source Range Nuclear Instrument Channel II new cable splice under WO133932510 implementing Engineering Change Package WBN-21-090, PMT under WO123220697 on September 13, 2023

(5) Unit 2 2A-A Boric Acid Transfer Pump Quarterly Pump Testing under WO\124022281 after seal replacement under WO124022281 on September 27, 2023.

Surveillance Testing (IP Section 03.01) (2 Samples)

(1) 1B Motor Driven AFW Pump Quarterly Surveillance on July 24, 2023, under WO123349890

(2) A-A MCR Chiller surveillance under WO123395204 on September 12, 2023

Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)

(1) Licensee entered Action Level III on July 6, 2023, and performed leak identification and correction per NETP-125 as documented in CR1866303.

71114.01 - Exercise Evaluation

Inspection Review (IP Section 02.01-02.11) (1 Sample)

(1) The inspectors evaluated the biennial emergency plan exercise during the week of August 14, 2023. The exercise scenario began with a dose equivalent iodine sample greater than the technical specification allowable limit. This met the conditions for declaring a Notice of Unusual Event. Afterwards, as radiation monitors indicate increasing levels, a reactor coolant system leak (RCS) develops. An Alert is declared when radiation monitor readings rise above values for loss of RCS barrier. Radiation levels continue to rise above threshold, and then main and auxiliary feedwater are lost. A Site Area Emergency for loss or potential loss of two barriers is declared. The RCS leak turns into a loss of coolant accident, driving containment pressure above threshold value. When a sequential trip of containment spray occurs, the conditions for a General Emergency are met allowing the Offsite Response Organizations to demonstrate their ability to implement emergency actions.

71114.04 - Emergency Action Level and Emergency Plan Changes

Inspection Review (IP Section 02.01-02.03) (1 Sample)

(1) The inspectors evaluated submitted Emergency Action Level, Emergency Plan, and Emergency Plan Implementing Procedure changes during the week of August 14, 2023. This evaluation does not constitute NRC approval.

71114.06 - Drill Evaluation

Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)

The inspectors evaluated:

(1) The inspectors observed the sites response to an emergency preparedness drill on July 19, 2023. This drill involved a dropped cask during transport, a primary to secondary leak that transitioned to a stream generator tube rupture, the ruptured steam generator later became faulted resulting in a release and a general emergency declaration.

71114.08 - Exercise Evaluation - Scenario Review

Inspection Review (IP Section 02.01 - 02.04) (1 Sample)

(1) The inspectors reviewed and evaluated in-office, the proposed scenario for the biennial emergency plan exercise prior to the day of the exercise.

71124.06 - Radioactive Gaseous and Liquid Effluent Treatment

Walkdowns and Observations (IP Section 03.01) (3 Samples)

The inspectors evaluated the following radioactive effluent systems during walkdowns:

(1) Auxiliary Building Vent

(2) Emergency Gas Treatment System

(3) Unit 1 Shield Building Vent

Sampling and Analysis (IP Section 03.02) (3 Samples)

Inspectors evaluated the following effluent samples, sampling processes and compensatory samples:

(1) Compensatory sampling records, Auxiliary Building effluent flow rate, 2/21/23 -

2/28/23

(2) Compensatory sampling records, Unit 2 Shield Building effluent monitor, 6/12/22 -

12/8/22

(3) Auxiliary Building effluent monitoring system sample lines

Dose Calculations (IP Section 03.03) (2 Samples)

The inspectors evaluated the following dose calculations:

(1) Liquid release permit L-20230615-1676-B

(2) Gaseous release permit G-20230616-1705-C

Abnormal Discharges (IP Section 03.04) (2 Samples)

The inspectors evaluated the following abnormal discharges:

(1) Unit 2 Steam Generator PORV release, 2/28/22 - 6/23/22

(2) Unit 2 Containment integrated leak rate test, 6/21/22 - 6/22/22

71124.07 - Radiological Environmental Monitoring Program

Environmental Monitoring Equipment and Sampling (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated environmental monitoring equipment and observed collection of environmental samples.

Radiological Environmental Monitoring Program (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated the implementation of the licensees radiological environmental monitoring program.

GPI Implementation (IP Section 03.03) (1 Sample)

(1) The inspectors evaluated the licensees implementation of the Groundwater Protection Initiative program to identify incomplete or discontinued program elements.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:

MS06: Emergency AC Power Systems (IP Section 02.05) ===

(1) Unit 1 (January 1, 2022, through June 30, 2023)

(2) Unit 2 (January 1, 2022, through June 30, 2023)

MS07: High Pressure Injection Systems (IP Section 02.06) (2 Samples)

(1) Unit 1 (July 1, 2022, through June 30, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023)

MS08: Heat Removal Systems (IP Section 02.07) (2 Samples)

(1) Unit 1 (July 1, 2022, through June 30, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023)

MS09: Residual Heat Removal Systems (IP Section 02.08) (2 Samples)

(1) Unit 1 (July 1, 2022, through June 30, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023)

MS10: Cooling Water Support Systems (IP Section 02.09) (2 Samples)

(1) Unit 1 (July 1, 2022, through June 30, 2023)

(2) Unit 2 (July 1, 2022, through June 30, 2023)

BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10) (2 Samples)

(1) Unit 1 (January 1, 2022, through June 30, 2023)

(2) Unit 2 (January 1, 2022, through June 30, 2023)

BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)

(1) Unit 1 (January 1, 2022, through June 30, 2023)

(2) Unit 2 (January 1, 2022, through June 30, 2023)

PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)

(1) April 16, 2022, through August 4, 2023

EP01: Drill/Exercise Performance (DEP) Sample (IP Section 02.12) (1 Sample)

(1) July 1, 2022, through March 31, 2023.

EP02: Emergency Response Organization (ERO) Drill Participation (IP Section 02.13) (1 Sample)

(1) July 1, 2022, through March 31, 2023.

EP03: Alert And Notification System (ANS) Reliability Sample (IP Section 02.14) (1 Sample)

(1) July 1, 2022, through March 31, 2023.

71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)

The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:

(1) Loss of High-Pressure Flex Pump Capability as documented under CR1862048 on June 12, 2023.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP section 03.02)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 391/2023-001-00 Unanalyzed Condition Related to the Loss of the 2A Emergency Diesel Generator During a Postulated Appendix R Fire. ADAMS Accession No. ML23201A109 The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71153. This LER is

===Closed.

Personnel Performance (IP section 03.03) (1 Sample)===

(1) The inspectors evaluated operator response to Unit 2 reactor scram due to feed regulating valve failure and licensees performance on August 4,

INSPECTION RESULTS

Failure to maintain records of activities affecting Quality Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.13] - 71111.12 Systems NCV 05000390,05000391/2023003-01 Consistent Open/Closed Process The inspectors identified an issue of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50 Appendix B, Criterion XVII, Quality Assurance Records for failure to maintain QA records. Specifically, 354 safety related items were installed in the plant did not have the QA records establishing traceability.

Description:

On July 26, 2023, inspectors identified that 480V Transformer Room 1B inlet damper, WBN-1-FCO-030-0248A a safety related component, was failed during an inspection. CR 1870666 was initiated to document the non-functional damper. During a review of repairs to the 480V Transformer Room 1B inlet damper, the inspectors discovered an apparent lack of traceability of the parts that were installed to repair the damper.

CR 1872072 was generated to document the questions from the inspectors regarding the traceability of components that were installed in the damper. The licensee identified that the parts were drawn from the warehouse under a material tracking activity 114914516. TVA-NQA-PLN89-A, Nuclear Quality Assurance Plan, incorporates the regulatory requirements associated with traceability for quality related items.

During the inspectors review of the material tracking activity, it was identified that 354 safety related items had been drawn from the warehouse under that single material tracking activity.

The inspectors questioned the licensee about the end use of the 354 items, and the licensee has not been able to establish traceability to any of the components except the items which were used to repair the damper.

Corrective Actions: CR 1872489 was initiated to document the apparent non-compliance and address the traceability issues.

Corrective Action References: CR 1872489, CR 1870666, CR 1872072.

Performance Assessment:

Performance Deficiency: The licensee's failure to maintain quality assurance records for safety related materials installed in the plant in accordance with 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records was the Performance Deficiency. Specifically, Work Orders and traceability information for 354 safety related components was not maintained as required by TVA-NQA-PLN89-A Rev. 41 and 10 CFR 50, Appendix B, Criterion XVII.

Screening: The inspectors determined the performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Inspection manual chapter (IMC) 0612 Appendix E provides examples of minor issues, and this finding most closely relates to example 1.a. when the licensee failed to document and evaluate test results in accordance with regulatory requirements. The example describes a performance deficiency affected the mitigating systems cornerstone attributes of equipment performance and procedure quality and adversely impacted the cornerstone objective. The example shows when a significant number of records associated with equipment are missing that reasonable assurance for operability is called into question. In this case, given a 10 CFR 21 notification against any of the 354 installed safety related components installed without traceability throughout the station, the licensee would be unable to determine the impact to safety.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The finding was routed through the process described in IMC 0609 Appendix A, Exhibit 2, Part A, Mitigating SSCs and PRA Functionality. All six questions were answered "no," therefore, the finding screens to green.

Cross-Cutting Aspect: H.13 - Consistent Process: Individuals use a consistent, systematic approach to make decisions. Risk insights are incorporated as appropriate. Specifically, TVA utilized a common material tracking activity for the installation of at least 354 safety related components installed throughout the station vice generating traceable records as required by regulations and station processes.

Enforcement:

Violation: Title 10 CFR 50, Appendix B, Criterion XVII, Quality Assurance Records, states, in part, that, Sufficient records shall be maintained to furnish evidence of activities affecting qualityThe records shall also include closely-related data such as qualifications of personnel, procedures and equipmentRecords shall be identifiable and retrievable.

TVA Nuclear Quality Assurance Plan Section 8.3.2 B. Traceability states in part, the traceability of materials, parts, or components to specific maintenance shall be provided either on the item or on records traceable to the item.

Contrary to the above, since July 26, 2013 the licensee has not maintained records related to the use of safety related components in an identifiable and retrievable manner. Specifically, the licensee failed to maintain records of the use of 354 safety related components that were drawn from the warehouse under material tracking activity 114914516.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to implement the requirements of 1-SI-61-7, 18-Month Ice Condenser Intermediate Deck Doors Operational Check Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.12] - Avoid 71111.15 NCV 05000390/2023003-02 Complacency Open/Closed The inspectors identified a Green finding and associated Non-Cited Violation (NCV) of Technical Specification 5.7.1.1a when the licensee failed to perform maintenance in accordance with written procedures. Specifically, the licensee failed to perform testing of the Unit 1 ice condenser intermediate deck doors in accordance with procedure 1-SI-61-7 when they used a force gauge unsuitable for the ambient temperature and without verifying the gauge was within calibration immediately after use. As a result, 8 intermediate deck doors were determined to be inoperable when discovered on September 21, 2023.

Description:

On September 21, 2023, during a routine six-month calibration check of force gauge E60686, the licensee discovered the gauge was out of the allowable calibration range of +/-0.5 lbf. The gauge measured -0.71 lbf and the licensee performed an out of tolerance investigation to determine the scope and impact of the use of the gauge since the last time it had been calibrated.

The licensee determined that E60686 was used with WO 122771580 on May 7, 2023, to determine the required lifting force as required by surveillance procedure 1-SI-61-7, 18-Month Ice Condenser Intermediate Deck Doors Operational Check, Revision 05. 1-SI-61-7 is performed every 18 months just prior to starting up from each refueling outage to demonstrate Ice Condenser Intermediate Deck Doors operable pursuant to Technical Specification Surveillance Requirements 3.6.12.6. This requirement visually verifies no structural deterioration, free movement of the vent assemblies, and ascertains free movement when lifted. The performance of this activity provides assurance that the intermediate deck doors are free to open in the event of a design basis accident.

The acceptance criteria in 1-SI-67-7 required that each intermediate door be free to move. As described in the bases for surveillance requirement 3.6.12.6, free movement of each door is verified when the lifting force for each door is less than or equal to a specific value included in the bases (which varies depending on the specific door). Upon reviewing the results from the test from WO 122771580 and including the additional out of tolerance amount of 0.21 lbf to the values obtained in the WO, the licensee discovered that the following eight doors failed to meet the established acceptance criteria for operability:

1. Bay 4, Door 3 had a measured lifting force of 31.1 lbs, a corrected lifting force of 31.81

lbs, and an acceptance criteria of <= 31.8 lbs.

2. Bay 4, Doors 8 had a measured lifting force of 30.3 lbs, a corrected lifting force of 31.01

lbs, and an acceptance criteria of <= 31.0 lbs.

3. Bay 6, 3 had a measured lifting force of 31.1 lbs, a corrected lifting force of 31.81 lbs, and

an acceptance criteria of <= 31.8 lbs.

4. Bay 6, Door 4 had a measured lifting force of 30.3 lbs, a corrected lifting force of 31.01

lbs, and an acceptance criteria of <= 31.0 lbs.

5. Bay 6, Door 6 had a measured lifting force of 33.1 lbs, a corrected lifting force of 33.81

lbs, and an acceptance criteria of <= 33.8 lbs.

6. Bay 9, Door 2 had a measured lifting force of 33.1 lbs, a corrected lifting force of 33.81

lbs, and an acceptance criteria of <= 33.8 lbs.

7. Bay 9, Door 3 had a measured lifting force of 31.2 lbs, a corrected lifting force of 31.91

lbs, and an acceptance criteria of <= 31.8 lbs.

8. Bay 10, Door 3 had a measured lifting force of 31.1 lbs, a corrected lifting force of 31.81

lbs, and an acceptance criteria of <= 31.8 lbs.

Upon review the licensees corrective actions, the inspectors identified additional concerns with the control of M&TE used for WO 122771580. The inspectors found per the vendor manual that the acceptable operating temperature for E60686 was between 40 °F and 110

°F. However, section 4.2 [2] of 1-SI-61-7 required that the M&TE used was suitable for use at 10°F which was the temperature normally experienced during performance of this surveillance. When questioned as to the suitability of E60686 to be used outside the range specified by the vendor, the licensee was unable to provide any technical justification for such a deviation.

Further, the inspectors determined that section 7.0 of 1-SI-61-7 required that the force gauge was returned to the M&TE tool room immediately after use for a post-use calibration check to ensure that the gauge remained within acceptable calibration limits before making a Mode change. When the inspectors questioned the licensee on the results of the post use check for WO 122771580, the licensee was unable to verify that the required post use check was performed.

The inspectors determined that the licensees failure to follow sections 4.2 [2] and 7.0 of 1-SI-61-7, as described above, directly led to the failure to identify that eight, ice condenser intermediate deck doors were inoperable from May 7, 2023 until September 21, 2023.

Corrective Actions: Licensee documented the issue in the corrective action program and preformed 1-SI-61-7 under WO 124005606 on September 21, 2023.

Corrective Action References: CR 1881873, CR 1881911, CR 1881663

Performance Assessment:

Performance Deficiency: The failure to perform maintenance in accordance with written procedures was a performance deficiency. Specifically, the licensee failed to use appropriate M&TE and failed to perform required post-use calibration checks. As a result, eight safety related ice condenser intermediate deck doors were later determined to be inoperable.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to implement the surveillance procedure resulted in eight intermediate deck doors being inoperable.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, Exhibit 3 C. Reactor Containment. Because the inspectors answered Yes to Question 1, the inspectors used IMC 0609 Appendix H, Containment Integrity SDP.

Screening of the performance deficiency resulted in no impact to CDF. Therefore, it was screened as a Type B Finding per Table 4.1 which lists ice condenser doors as SSCs considered for LERF Implications. This screened to Table 7.1 which indicates to perform a Phase 2 assessment per Table 7.2. As the blockage was less than 15 percent, the issue screens to Green.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, individuals did not question the procedural adherence due to previous successful performances.

Enforcement:

Violation: Watts Bar Unit 1 Technical Specification 5.7.1.1.a, Administrative Procedures, requires that written procedures shall be established, implemented, and maintained covering activities referenced in Regulatory Guide 1.33, Revision 2, dated February 1978, Appendix A.

Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance, requires, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures.

Contrary to the above, on May 7, 2023, the licensee failed to perform maintenance on safety-related equipment in accordance with written procedures. Specifically, the licensee failed to perform testing of the Unit 1 ice condenser intermediate deck doors in accordance with procedure 1-SI-61-7 when they used a force gauge unsuitable for the ambient temperature and without verifying the gauge was within calibration immediately after use. As a result, 8 intermediate deck doors were determined to be inoperable when the failure to follow 1-SI-61-7 was discovered on September 21, 2023.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to ensure bounding technical requirements were incorporated into a Procedurally Controlled Temporary Modification Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.12] - Avoid 71111.18 Systems NCV 05000391/2023003-03 Complacency Open/Closed The inspectors identified a Green finding and associated Non-Cited Violation of 10 CFR 50 Appendix B Criterion III for the failure of the licensee to ensure that bounding technical requirements were met prior to installation of a temporary modification. Specifically, the licensee failed to ensure that adequate technical requirements were incorporated into Procedurally Controlled Temporary Modification, 0-TI-66.004 Revision 001, prior to issuance.

Description:

During a review of station logs, the inspectors identified the installation of 7 temporary dehumidifiers in Unit 2 Upper Containment. Further inspection determined that 0-TI-66.004 was revised to allow for the increased number of dehumidifiers. However, inspectors identified that the revision to the technical evaluation did not take into account the actual plant configuration of the Upper Compartment Coolers. Specifically, the technical evaluation credited margin in the capacity of 2 Upper Compartment Coolers for acceptability of the Temporary Modification. Upper Compartment Coolers 2A and 2D were removed as part of TMOD WBN-2-2022-067-01 as a result of issues encountered during outage 2R4 close-out. Upper Compartment Cooler 2B was tagged out on October 31, 2022, due to repeated tripping of the supply breaker on thermal overloads. This resulted in a configuration of only a single Upper Compartment Cooler in service on Unit 2 since the October 31, 2022.

0-TI-66.004 Revision 1 Technical Evaluation was approved on August 04, 2023, following the installation of the additional dehumidifiers the Resident Inspectors questioned the adequacy of the evaluation which differed from the actual plant conditions. The licensee secured the additional dehumidifiers and documented the issue in CR 1873023 and Revision 0010 to 0-TI-66.004 was issued on August 09, 2023. The updated analysis required the utilization of GOTHIC heat transfer modeling to demonstrate that the additional heat load did not adversely impact the accident analysis initial conditions.

TVA-NQA-PLN89-A, TVA Nuclear Quality Assurance Plan, states that 10 CFR 50 Appendix B Criterion III is implemented through Section 7.0 Design Control. Section 7.2.5 Design Output A. states: "Engineering requirements on plant activities (e.g., operation, maintenance, installation, modification, surveillance) shall be identified in design output documents." TVA-NQA-PLN89-A Nuclear Quality Assurance Plan Section 7.2.5 Design Output B. requires that:

"Measures shall be established and documented to control the preparation, review, approval, issuance, and revision of design output documents. These measures shall include criteria and responsibilities to ensure that adequate technical and quality requirements are incorporated prior to issuance."

NPG-SPP-09.5, Temporary Modifications Temporary Configuration Changes implements the requirements of TVA-NQA-PLN89-A. NPG-SPP-09.5 Section 3.2.5 B requires in part that Procedurally Controlled Temporary Modification Technical Evaluations are prepared in accordance with Attachment 9.

NPG-SPP-09.

requires in part that the Responsible Engineer will use the Design Attribute Review to identify potential impacts. Additionally Block 5 requires that the Responsible Engineer Describe how/why the item evaluated is acceptable from a nuclear safety standpoint.

Corrective Actions: The issue was documented in the corrective action program and a revision to the technical evaluation was performed.

Corrective Action References: CR 1873023

Performance Assessment:

Performance Deficiency: Failure to implement the requirements NPP-SPP-09.5 Temporary Modifications Temporary Configuration Changes and 10 CFR 50 Appendix B Criterion III is the performance deficiency. Specifically, Attachment 9 Technical Evaluation Instructions Block 4 requires the Responsible Engineer will use the Design Attribute Review to identify potential impacts. Additionally Block 5 requires that the Responsible Engineer Describe how/why the item evaluated is acceptable from a nuclear safety standpoint. The technical evaluation for 0-TI-66.004 Revision 1 did not ensure that the requirements for installation of the temporary modification were incorporated into the precautions and limitations of TI-66.004 prior to issuance. Additionally, the technical evaluation did not demonstrate that acceptability of the design based on actual plant configuration and required additional analysis to demonstrate acceptability.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately evaluate the change to prior to implementation challenged the initial condition assumptions of containment in the accident analysis and required reanalysis to demonstrate acceptability.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The performance deficiency was more than minor due to the need to revise the technical evaluation, which required the utilization of the GOTHIC heat transfer modeling to demonstrate acceptability of the new configuration. This condition is similar to Example 3.a of IMC 0612 Appendix E. Screening this issue through 0609 Appendix A, Exhibit 3 C. all questions are answered NO which results in Screen to Green.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, individuals failed to ensure that the physical plant configuration was bounded by the analysis being performed for the procedurally controlled TMOD.

Enforcement:

Violation: 10 CFR 50 Appendix B Criterion III states in part that: measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into procedures and instructions.

Contrary to the above from August 04, 2023, to August 09, 2023, the licensee failed to ensure that adequate technical requirements were incorporated into Procedurally Controlled Temporary Modification, 0-TI-66.004 Revision 001, prior to issuance. This resulted in the installation of 7 temporary dehumidifiers into Unit 2 Upper Containment without bounding technical analysis to demonstrate the acceptability of the configuration.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to obtain NRC approval prior to implementing a Technical Specification (TS) change for Watts Bar Unit 1 Mode 5 requirement Cornerstone Severity Cross-Cutting Report Aspect Section Not Severity Level IV Not 71111.18 Applicable NCV 05000390/2023003-04 Applicable Open/Closed An NRC-identified a Severity Level IV Non-Cited Violation of 10 CFR Part 50.59(c)(1) for failure to obtain NRC approval prior to implementing a Technical Specification (TS) change for Watts Bar Unit 1 Mode 5 requirement. Specifically, the licensee failed to perform adequate 50.59 Evaluations/Screenings for the proposed TS change requirement for Mode 5 with one reactor vessel closure bolt out of service from 2020 to 2023 for Watts Bar Unit 1 without NRC approval wherein Note b of TS Table 1.1-1, Modes is defined by TS for Mode 5 as All reactor vessel head closure bolts fully tensioned.

Description:

On May 2, 2023, Watts Bar Unit 1 entered Mode 5, Cold Shutdown (Note b),wherein Note b is defined by Technical Specifications as All reactor vessel head closure bolts fully tensioned. At this time, only 53 of 54 reactor vessel head closure bolts were tensioned due to inability to tension location #34. On May 3, 2023, the Resident staff discovered the failure to tension all reactor vessel head closure bolt locations during review of condition reports and station logs. When the residents brought this to the Shift Manager, the licensee was not aware of the condition of the bolting or why less than all locations tensioned was acceptable. The regional Projects Branch consulted with the NRR Program and NRR TS Branch concerning the All verbiage in Note b. They confirmed Watts Bar would require an approved license amendment to change from Mode 6 to 5 with any less than 54 reactor vessel head closure bolts fully tensioned as any deviation from 54 bolts would be a change in the common interpretation of All. This position was consistent with IN 97-80, Licensee Technical Specifications Interpretations.

The NRC inspectors reviewed the licensees 50.59 evaluation/screening documents. During reactor vessel stud removal activities on May 13, 2020, the licensee identified thread damage on reactor vessel stud hole #34. On May 24, 2020, unable to make timely repair to the damaged threads, the licensee performed a 50.59 evaluation/screening (WBN-20-1371 Rev.

0) as a Temporary Modification (TMOD). The TMOD provided justification that WBN Unit 1 could operate with one reactor head vessel closure stud out-of-service for Cycle 17. The 50.59 screening stated, Since the FSAR is being revised to note only 53 studs are in service for Unit 1 Cycle 17 and the analysis [by Westinghouse] shows the ASME Code is still met with only 53 studs in service, the intent of T/S Table 1.1-1 is met and does not warrant revision by this Temporary Modification.

Over time, unable to affect repair to location #34, TVA with the help of Westinghouse analyses expanded the change to apply to any reactor vessel head stud location on either unit.

Corrective Actions: At 0237, on May 4, 2023, Watts Bar Unit 1 returned to Mode

6. Subsequently, TVA requested and received an exigent TS amendment from the NRC to

operate Unit 1 for the current fuel cycle with location #34 vacant.

Corrective Action References: CR 1854299 and 1607752

Performance Assessment:

The inspectors determined this violation was associated with a minor performance deficiency. The inspectors determined that the failure to perform adequate evaluation/screenings for one reactor vessel stud out of service for Mode 5 was contrary to 10 CFR 50.59

(c) 1 and was the performance deficiency (PD).

If possible, the underlying technical issue is evaluated under the SDP to determine the risk significance of the issue. In this case, the inspectors determined the finding could be evaluated using inspection Manual Chapter 0612, Issue Screening, effective date, 01/01/2020 and determined that this issue is considered minor under the ROP because the PD did not adversely affect the Barrier Integrity cornerstone objective. The Westinghouse engineering evaluation of one reactor vessel stud out of service shows that there is no adverse effect on the design functions of the reactor vessel, or any other component described in the updated final safety analysis report (UFSAR).

Enforcement:

The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance.

Severity: NRC Enforcement Policy dated January 15, 2020, Section 2.1.3, Enforcement of 10 CFR 50.59, the activity or change

(1) required prior Commission review and approval, and the licensee failed to obtain Commission approval as example of a SL IV Violation.

Violation: Title 10 CFR Part 50.59, Changes, Tests, and Experiments,Section I(1) states, in part, that a licensee may make changes in the facility as described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to Sec.

50.90 only if:

(i) A change to the technical specifications incorporated in the license is not required.

Note

(b) of Watt Bar Unit 1 TS Table 1.1-1, Modes states, in part, that All reactor vessel head closure bolts must be fully tensioned for Mode 5.

Question 5, Technical Specification Change of Attachment 2, Format and Content of Screening Review of NPG-SPP-09.4, 10 CFR 50.59 Evaluations of Changes, Tests, and Experiments, Revision 15, states, in part, that screening Reviews indicate a need for a Technical Specification change will require a license amendment be submitted and approved by NRC in accordance with licensing procedures, prior to implementation.

Contrary to the above, from May 24, 2020, to May 3, 2023, the licensee made changes to the facility as described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to Sec. 50.90 for a change to the technical specifications incorporated in the license. In accordance with the Enforcement Policy, the violation was classified as a Severity Level IV violation because this violation was not repetitive or willful and was entered into the licensees CAP as CR1854299 and subsequently, an exigent license amendment was submitted and approved by NRC for the current operating cycle.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to implement Surveillance Requirement 3.6.12.2, Ice Condenser Doors Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Pending [H.11] - 71111.24 AV 05000391/2023003-05 Challenge the Open Unknown EA-23-130 An NRC-identified finding with its safety significance as yet to be determined (TBD) and Apparent Violation (AV) of Technical Specification 3.6.12, Ice Condenser Doors, was identified for the licensees failure to implement the requirements of surveillance instruction 2-SI-61-6, Unit 2 Weekly Ice Condenser Intermediate Deck Doors Visual Inspection.

Specifically, the licensee failed to ensure that adequate Measuring and Test Equipment (M&TE) was utilized to provide reasonable assurance that the Intermediate Deck Doors remained operable.

Description:

On August 10, 2023, the resident office identified issues with the as left pull testing under 2-SI-61-6, Unit 2 Weekly Ice Condenser Intermediate Deck Doors Visual Inspection, specifically, the licensee failed to properly perform as-left testing of Ice Condenser Intermediate Deck Doors (IDD). During further review the inspectors identified that the Measuring and Test Equipment (MT&E) that was utilized for the surveillance did not meet the environmental requirements specified in 2-SI-61-6. Specifically, the force gauge was not rated for a 10°F environment. The inspectors found per the vendor manual that the acceptable operating temperature for the gauge was between 40 °F and 110 °F.

CR 1874466 on August 14, 2023, documented the inspectors challenge to the licensees MT&E selection. CR 1876542 on August 24, 2023, documented a calibration failure of the force gauge in a non-conservative direction for the application. Specifically, the force gauge was reading 0.71 lbf lower than the actual force applied. Licensee staff completed the Out-of-Tolerance investigation for the affected work orders and determined that there was impact to operability during the period under investigation at potentially greater than 15% of locations.

Corrective Actions: Licensee performed an immediate operability review and is performing a past operability evaluation.

Corrective Action References: 1874466, 1876542

Performance Assessment:

Performance Deficiency: Failure to implement the requirements of 2-SI-61-6, Unit 2 Weekly Ice Condenser Intermediate Deck Doors Visual Inspection, was a Performance Deficiency.

Specifically, the licensee failed to ensure that adequate Measuring and Test Equipment (M&TE) was utilized to provide reasonable assurance that the Intermediate Deck Doors remained operable.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee failed to ensure that adequate Measuring and Test Equipment (M&TE) was utilized to provide reasonable assurance that the Intermediate Deck Doors remained operable.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix H, Containment Integrity SDP. The performance deficiency is more that minor due to the adverse impact to the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone objective. Specifically, the Blockage of the intermediate deck doors challenges the ability of containment to prevent early containment failure. IMC 0609 Appendix A Exhibit 3 C. Reactor Containment Question 1 is answered YES--> Stop. Go to IMC 0609, Appendix H.

IMC 0609 Appendix H Table 4.1 List ice condenser doors as SSCs considered for LERF Implications. Processed as a Type B finding. Table 7.1 screens to Perform Phase 2. Table 7.2 screens to a detailed risk analysis and could not be screened to Green and is pending a significance determination.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, individuals did not ensure that the M&TE was suitable for a 10°F environment prior to use and did not question the M&TE issued.

Enforcement:

Violation: Watts Bar Unit 2 Technical Specification Surveillance Requirement SR 3.6.12.2 requires in part to verify by visual inspection each intermediate deck door is closed and not impaired by ice, frost, or debris.

Section 1.2.2 of Licensee Surveillance Instruction 2-SI-61-6, states in part that performance of this instruction satisfies the following Surveillance Requirement (SR) 3.6.12.2.

Section 4.2 [2] of Licensee Surveillance Instruction 2-SI-61-6 also states in part to ENSURE that the M&TE is available is suitable for use at 10 F with a range of 50 lbs and an accuracy of +/-1.0% (of range).

Contrary to the above, from April 06, 2023 to August 13, 2023, the M&TE utilized for the performance of 2-SI-61-6 was not suitable for use at 10°F. Specifically, the M&TE used at Watts Bar for the Surveillance testing is not rated for use in a 10°F environment and has subsequently failed post use calibration. The failure of the M&TE has called into question the operability of the ice condenser intermediate deck doors. Licensee initiated CR 1880146 on September 13, 2023, documented the work orders that were impacted in which intermediate deck doors did not meet acceptance criteria.

Enforcement Action: This violation is being treated as an apparent violation pending a final significance (enforcement) determination.

Failure to Maintain N High Pressure Flex Capability Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Pending [H.4] - 71152A Systems AV 05000390,05000391/2023003-06 Teamwork Open EA-23-131 A self-revealed finding with its safety significance as yet to be determined (TBD) and Apparent Violation (AV) of 10 CFR 50.155, Mitigation of Beyond Design Basis Events, was identified for the licensees failure to implement the requirements of technical instruction 0-TI-446, Diverse and Flexible Coping Strategies (FLEX) Program Bases. Specifically, Watts Bar did not maintain the equipment (hose couplings) necessary to support the deployment of the High Pressure (HP) FLEX pumps as needed to support Phase 2 mitigation strategy defined in 0-TI-446.

Description:

On June 12, 2023, during annual testing of the High Pressure (HP) Flex Pumps the licensee discovered that the HP FLEX pump hoses had the incorrect fittings. Specifically, the fittings would not allow for the connection of any of the HP Flex pumps to plant systems.

CR 1862048 was initiated to document the issue. The last successful performance of the surveillance on the HP FLEX pumps was on July 18, 2022.

The HP FLEX pumps are required by 0-TI-446, Diverse and Flexible Coping Strategies (FLEX) Program Bases and 10 CFR 50.155 Mitigation of Beyond Design Basis Events, to support Phase 2 mitigation strategy wherein they maintain RCS inventory and/or restore the reactivity control function.

Corrective Actions: Licensee documented the issue in the Corrective Action program.

Corrective Action References: CR 1862048

Performance Assessment:

Performance Deficiency: The licensees failure to implement the requirements of 0-TI-446, Diverse and Flexible Coping Strategies (FLEX) Program Bases and 10 CFR 50.155 Mitigation of Beyond Design Basis Events, was a performance deficiency within the ability to foresee and prevent. Specifically, Watts Bar did not maintain the equipment necessary to support the deployment of the HP FLEX pump as needed to support Phase 2 mitigation strategy.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loss of all HP Flex pump ability to inject for RCS inventory control results in the inability to implement the Phase 2 Flex Strategy.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. IMC 0609 Appendix A Exhibit 2 E. Flexible Coping Strategies (FLEX) Question 2 is answered YES--> Stop. Go to Detailed Risk Evaluation section. The finding could not be screened to Green and is pending a significance determination.

Cross-Cutting Aspect: H.4 - Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the failure of the licensee organization to effectively communicate the cause, interim compensatory actions, and required corrective actions to all impacted stake holders resulted in extending the out of service time by months for the High-Pressure Flex Pumps.

Enforcement:

Violation: 10 CFR 50.155(c)(1) states, in part, The equipment relied on for the mitigation strategies and guidelines required by paragraph (b)(1) of this section must have sufficient capacity and capability to perform the functions required by paragraph (b)(1) of this section.

10 CFR 50.155(b)(1) requires, in part, that these strategies and guidelines must include, maintaining or restoring core cooling.

0-TI-446 section 1.5.2 Phase 2 implements requirements of 10 CFR 50.155(b)(1) via the HP FLEX Pumps credited to maintain or restore core cooling.

Contrary to the above, since July 18, 2022, equipment, HP FLEX pumps, relied on for the mitigation strategies and guidelines required by 10 CFR 50.155(b)(1), as implemented by the licensee in accordance with 0-TI-446 section 1.5.2, have not had sufficient capability to perform the functions required. Specifically, the licensee has not maintained the capability to maintain or restore core cooling due to the inability to connect the HP Flex Pumps to plant systems.

Enforcement Action: This violation is being treated as an apparent violation pending a final significance (enforcement) determination.

Licensee identified Unanalyzed Condition for the 2A-A Emergency Diesel Generator Cornerstone Significance/Severity Cross-Cutting Report Aspect Section Mitigating Pending None (NPP) 71153 Systems AV 05000390,05000391/2023003-07 Open EA-23-129 A licensee-identified finding with its safety significance as yet to be determined (TBD) and Apparent Violation (AV) of Technical Specification (TS) 5.7.1.1 d, Fire Protection Program implementation, was identified for the licensees failure to implement the requirements Watts Bar Fire Protection Program, including, that each Emergency Diesel Generator (EDG) and its associated equipment are separated from each other by 3-hour fire barriers. Specifically, the alternate feeder common to the 480V Shutdown Boards 2A1-A and 2A2-A lacked the required fire barrier.

Description:

On May 25, 2023, at 1345 Eastern Daylight Time, TVA staff performing a Fire Probabilistic Risk Assessment for Watts Bar notified operations personnel of a missing (uninstalled) fire barrier (protective cable covering/coating) on the alternate feeder common to the 480V Shutdown Boards 2-BD-212-A1-A (2A1-A) and 2-BD-212-A2-A (2A2-A) within Fire Area 737-A1B. The normal and alternate feeders for these two shutdown boards share the same breaker on the associated 480V shutdown board.

Lacking this barrier on the alternate feeder, during a fire event within Fire Area 737-A1B, the common breaker on boards 2A1-A and 2A2-A would trip on a short to ground causing the loss of both boards. This would result in the loss of the 2A EDG, which is the credited power source for an Appendix R fire in this area. Specifically, the 2A EDG would lose cooling water and the inability to pump fuel oil from the 7-day tank to the day tank on loss of the 480V shutdown boards resulting in a run-time failure of the EDG. Without the credited source of Appendix R power, Watts Bar Unit 2 was in an unanalyzed condition (LER-2023-001). The licensee previously installed the fire barrier on the normal feeder in 2010 correcting the legacy design issue which identified both cables lacking a fire barrier required by the Watts Bar Fire Protection report. The licensee found that 2010 corrective maintenance for the alternate feeder failed to install the required fire barrier.

Corrective Actions: Licensee entered the issue in the corrective action program and has installed a plant modification to remedy the unanalyzed condition.

Corrective Action References: CR 1858921

Performance Assessment:

Performance Deficiency: Failure to implement the requirements of the Watts Bar Fire Protection program, including, that each EDG and its associated equipment are separated from each other by 3-hour fire barriers, was a performance deficiency. Specifically, the alternate feeder common to the 480V Shutdown Boards 2A1-A and 2A2-A lacked the required fire barrier.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, given a fire initiator in area 737-A1B, the 480V board that supports auxiliary loads for the 2A EDG would lose function due to cable fire damage resulting in loss of function of the 2A EDG which is the credited safe shutdown power supply for that scenario. Specifically, the damaged cabling would result in a loss of emergency raw cooling water to the EDG which would cause a failure to load-run under accident conditions.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix F, Fire Protection and Post - Fire Safe Shutdown SDP. IMC 0609 Appendix F 1 screens the finding as a high level of degradation due to a missing (not degraded) fire barrier that impacts function of credited safe shutdown power supply. The area of the missing fire barrier is not covered by detection or suppression systems and the licensee does not have a completed fire PRA. Therefore, the finding could not be screened to Green and is pending a significance determination.

Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.

Enforcement:

Violation: Watts Bar Operating License Condition 2.C(8) requires that TVA shall implement and maintain in effect all provisions of the approved fire protection program as described in the Fire Protection Report (FRP) for the facility, as described in NUREG-0847, Supplement 29. In the FPR Part VIII, TVA requires that each EDG and its associated equipment are separated from each other by 3-hour fire barriers.

Contrary to the above, from 2010 to June 2023, a fire barrier for area 737-A1B was not installed and would render the 2A EDG not operable in the event of a fire on the Unit 2 side of elevation 737 in the Auxiliary Building on 2010. The 2A EDG is the credited power source for fire safe shutdown for a fire located in this area. Without the credited source of power, this placed Watts Bar Nuclear Plant (WBN) Unit 2 in an unanalyzed condition.

Enforcement Action: This violation is being treated as an apparent violation pending a final significance (enforcement) determination.

EXIT MEETINGS AND DEBRIEFS

The inspectors verified no proprietary information was retained or documented in this report.

• On October 18, 2023, the inspectors presented the integrated inspection results to Mr.

Anthony Williams, Site Vice President and other members of the licensee staff.

• On August 3, 2023, the inspectors presented the Radiation Protection Exit Meeting inspection results to Anthony Williams, Site Vice President and other members of the licensee staff.

• On August 17, 2023, the inspectors presented the Emergency Preparedness Exercise Inspection Exit Meeting inspection results to B. Jenkins, Plant Manager and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Miscellaneous WBN-SDD-N3- Chemical and Volume Control system description Revision 40

2-4001

71111.04 Miscellaneous WBN-SDD-N3- Safety Injection system description Revision 39

63-4001

71111.04 Miscellaneous WBN-SDD-N3- Emergency Gas Treatment System 0014

65-4001

71111.04 Miscellaneous WBN-SDD-N3- Standby Diesel Generator system description Revision 29

2-4002

71111.05 Fire Plans AUX-0-692-01 WBN Prefrie Plan Auxiliary Building Elevation 692 Revision 4

71111.05 Fire Plans AUX-0-692-02 WBN-Prefire Plan Auxiliary Building Elevation 692 Revision 6

71111.05 Fire Plans AUX-0-692-03 WBN-PreFire Plan Auxiliary Building Elevation 692 Revision 2

71111.05 Fire Plans AUX-0-692-04 WBN Pre-Fire Plan - ERCW Tunnels A&B 692 elev. Revision 1

71111.05 Fire Plans AUX-0-772-01 Auxiliary Building Elevations 772 and 782 Revision 1

71111.05 Fire Plans CON-0-708-01 Pre-Fire Plan Control Building Elevation 708 Revision 3

71111.05 Fire Plans DGB-0-742-01 Diesel Generator Building Revision 4

71111.05 Fire Plans DGB-0-760-01 Diesel Generator Building Revision 2

71111.05 Fire Plans Fire Protection Fire Protection Report Revision 60

Report Part I

71111.05 Fire Plans FPIP-23-0312 The 1A-A Electrical BD RM CO2 timer not working as 07/06/2023

designed. (CR 1866410)

71111.12 Corrective Action CR 1870666 Damper Linkage may be broken 07/26/2023

Documents

Resulting from

Inspection

71111.12 Corrective Action CR 1872072 NRC Question: MMWO Documentation 08/02/2023

Documents

Resulting from

Inspection

71111.12 Corrective Action CR 1872489 NRC Question: Material Tracker 08/03/2023

Documents

Resulting from

Inspection

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.12 Work Orders WO 114914516 Material Tracking 08/02/2023

71111.15 Miscellaneous VTD-G063-0060 General Atomics Electromagnetics Radiation Monitoring 01

System Particulate and Gas Monitor System Technical

Manual

71111.24 Corrective Action CR 1866643 NRC Identified. Condition associated with CR 186303 has 07/06/2023

Documents reached NETP-125 Action Level III for RCS unidentified

Resulting from leakage was NRC identified and subsequently verified by

Inspection engineering.

71114.01 Corrective Action CR 1815824 Potential Adverse Trend Related to TEENS Tests

Documents

71114.01 Procedures EPIP-1 Emergency Plan Classification Logic Rev. 58

71114.01 Procedures EPIP-2 Notification of Unusual Event Rev. 48

71114.01 Procedures EPIP-3 Notification of Alert Rev. 51

71114.01 Procedures EPIP-4 Site Area Emergency Rev. 49

71114.01 Procedures EPIP-5 General Emergency Rev. 61

71114.04 Corrective Action CR 1824194 Procedure Change Request (PCR) EPIP-1 Revision 57 page

Documents 96 figure missing information

71114.04 Miscellaneous CECC 2022-046 Effectiveness Evaluation Form for EPIP-1 Rev. 57 12/01/2022

71114.04 Miscellaneous CECC 2022-047 Effectiveness Evaluation Form for EPIP-1 Rev. 57 12/01/2022

71114.04 Miscellaneous CECC 2023-008 Effectiveness Evaluation Form for EPIP-1 Rev. 58 02/06/2023

71114.04 Miscellaneous CECC 2023-022 Screening Evaluation Form for WBN EPIP-1, Emergency 05/25/2023

Plan Classification Logic, Rev. 59

71114.04 Procedures REP-Appendix C Watts Bar Nuclear Plant Radiological Emergency Plan Rev. 115

71114.04 Procedures REP-Generic Radiological Emergency Plan (Generic Part) Rev. 115

71114.06 Miscellaneous 2023 WBN July Integrated Training Drill (Team A) 07/19/2023

71114.08 Miscellaneous 2023 WBN August Graded Exercise (Team A) 06/12/2023

71114.08 Procedures EPIP-1, WBN Hot and Cold Condition ICs/EALs Rev. 58

71114.08 Procedures NP-REP, State Multijurisdictional Radiological Emergency Response Rev. 92

Appendix E, Page Plans

E-1

71114.08 Self-Assessments 2022 WBN June Quarterly Training Drill 06/22/2022

71114.08 Self-Assessments 2022 WBN August EMPE Drill Report 09/28/2022

71114.08 Self-Assessments 2021WBN Graded Exercise Report 11/22/2021

Inspection Type Designation Description or Title Revision or

Procedure Date

71124.06 Corrective Action CR 1872198

Documents

Resulting from

Inspection

71124.06 Corrective Action CR 1872342

Documents

Resulting from

Inspection

71124.06 Miscellaneous 2022 Annual

Radioactive

Effluent Release

Report

71124.06 Procedures NPG-SPP-18.3.5 Equipment Important to Emergency Response Revision 12

71124.07 Corrective Action CR1616608

Documents

71124.07 Corrective Action CR1736664

Documents

71124.07 Corrective Action CR1774165

Documents

71124.07 Corrective Action CR1828058

Documents

71124.07 Corrective Action CR1856857

Documents

71124.07 Corrective Action CR1861244

Documents

71124.07 Corrective Action CR1866579

Documents

71124.07 Miscellaneous 2021 Annual Radiological Environmental Operating Report May 2022

71124.07 Miscellaneous 2022 Annual Radiological Environmental Operating Report May 2023

71124.07 Miscellaneous 2021 Site Conceptual Model Update 10/26/2021

71124.07 Miscellaneous Meteorological Data Recovery Tables for Watts Bar Nuclear Years 2020 -

Plant 2022

71124.07 Procedures 0-CM-1.07 Strategic Plan for Groundwater Protection 09/08/2017

71124.07 Procedures 0-PI-CEM-11.0 Groundwater Monitoring 02/22/2023

Inspection Type Designation Description or Title Revision or

Procedure Date

71124.07 Procedures 0-TI-449 Watts Bar Tritium Management Strategic Plan 10/04/2021

71124.07 Radiation Watts Bar Chemistry Sample Results for Groundwater 01/01/2023 -

Surveys Monitoring Wells 07/12/2023

71151 Corrective Action CR 1802682 During the August EMPE REP drill controllers engaged in

Documents coaching of participants

71151 Corrective Action CR 1807287 Siren failed monthly activation test on 10/5/22

Documents

71151 Corrective Action CR 1832927 OSC PA System Non-functional

Documents

29