IR 05000390/2023004

Integrated Inspection Report 05000390/2023004 and 05000391/2023004 and Apparent Violation

ML24043A083
Person / Time
Site:	Watts Bar
Issue date:	02/13/2024
From:	Son Ninh Division Reactor Projects II
To:	Jim Barstow Tennessee Valley Authority
References
EA-23-104, EA-23-129, EA-23-130, EA-23-131 IR 2023004
Download: ML24043A083 (30)

Text

SUBJECT:

WATTS BAR NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000390/2023004 AND 05000391/2023004 AND APPARENT VIOLATION

Dear James Barstow:

On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Watts Bar Nuclear Plant. On January 31, 2024, the NRC inspectors discussed the results of this inspection with Mr. Anthony Williams, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.

Section 71111.20 of the enclosed report discusses 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 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, for failure to promptly identify and correct the impact of Unit 2 feedwater leakage on ice condenser lower inlet doors. This condition led to the accumulation of ice on the lower inlet doors challenging their ability to open.

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).

Four findings of very low safety significance (Green) are documented in this report. Four of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.February 13, 2024 Licensee-identified violations which were determined to be of very low safety significance are 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, 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: 05000390/2023004 and 05000391/2023004

Licensee: Tennessee Valley Authority

Facility: Watts Bar Nuclear Plant

Location: Spring City, TN 37381

Inspection Dates: October 01, 2023 to December 31, 2023

Inspectors: W. Deschaine, Senior Resident Inspector D. Lanyi, Senior Operations Engineer M. Magyar, Reactor Inspector A. Price, Resident 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. Licensee-identified non-cited violations are documented in report sections: 71152A and 7115

List of Findings and Violations

Failure to Implement Inservice Test Requirements in Accordance with ASME OM Code Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.14] - 71111.24 Systems NCV 05000390,05000391/2023004-01 Conservative Open/Closed Bias The inspectors identified a Green finding and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(f), for the licensee's failure to implement inservice testing (IST) requirements. The licensee failed to properly implement testing procedures for Essential Raw Cooling Water (ERCW) pump discharge check valve 0-CKV-67-503D, and did not take proper corrective action following discovery of unacceptably degraded valve internals in accordance with the 2004 edition through the 2006 addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), as incorporated by reference in 10 CFR 50.55a, which is the current IST Program Code of Record for Watts Bar.

Failure to Identify Condition Adverse to Quality Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Pending [P.2] - 71111.20 AV 05000391/2023004-02 Evaluation Open The inspectors identified a finding with its safety significance as yet to be determined and associated apparent violation (AV) of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, for failure to promptly identify and correct the impact of Unit 2 feedwater leakage on ice condenser lower inlet doors. This condition led to the accumulation of ice on the lower inlet doors challenging their ability to open.

Failure to implement Surveillance Requirement 3.6.12.2 Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.11] - 71111.24 NCV 05000391/2023003-05 Challenge the Closed Unknown EA-23-130 The inspectors identified a Green finding and an associated NCV of Technical Specification (TS) 3.6.12, Ice Condenser Doors, for the licensees failure to implement the requirements of surveillance instruction 2-SI-61-6, Weekly Ice Condenser Intermediate Deck Doors (IDD)

Visual Inspection. Specifically, the licensee failed to ensure that adequate Measuring and Test

Equipment (M&TE) was utilized to provide reasonable assurance that the IDD remained operable.

Failure to Maintain N High Pressure Flex Capability Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.14] - 71152A Systems NCV 05000390,05000391/2023003-06 Conservative Closed Bias EA-23-131 A self-revealed Green finding and associated NCV of 10 CFR 50.155, Mitigation of Beyond Design Basis Events, was identified for the licensees failure to implement the requirements of 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.

Failure of Surveillance Instruction 1/2-SI-61-3, Unit 1/2 Ice Condenser Flow Passage Inspection, to include appropriate acceptance criteria for determining that a Surveillance Requirement was met Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.11] - 71152A NCV 05000390,05000391/2023004-03 Challenge the Open/Closed Unknown The inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B,

Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee s failure to include appropriate quantitative or qualitative acceptance criteria in Surveillance Instruction 1/2-SI-61-3, Unit 1/2 Ice Condenser Flow Passages Inspection could result in TS 3.6.11 not being met.

Additional Tracking Items

None.

PLANT STATUS

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

Unit 2 began the inspection period at RTP. On November 3, 2023, the unit was shut down for a planned refueling outage (U2R5). The unit was returned to 100 percent RTP on December 4, 2023, and remained there 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

Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures on November 27, 2023, for the following systems:

• Chemical Volume Control System

• Emergency Raw Cooling Water

• Auxiliary Feedwater (AFW)

71111.04 - Equipment Alignment

Partial Walkdown Sample (IP Section 03.01) (2 Samples)

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

(1) Unit 1 and Unit 2 spent fuel pool cooling system during Unit 2 refueling outage while the core is off-loaded on November 16, 2023

(2) Unit 2 residual heat removal system during the first week of the outage on November 10, 2023

Complete Walkdown Sample (IP Section 03.02) (1 Sample)

(1) The inspectors evaluated system configurations during a complete walkdown of the Unit 2 AFW system on November 24, 2023.

71111.05 - Fire Protection

Fire Area Walkdown and Inspection Sample (IP Section 03.01) (4 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) Unit 2 control rod drive equipment room, pressurizer heater transformer room and emergency gas treatment filter room on November 21, 2023

(2) Auxiliary building elevation 757' shutdown board rooms on November 22, 2023

(3) Unit 2 containment on November 23, 2023

(4) Diesel generator building board rooms, exhaust rooms and intake rooms on December 15, 2023

===71111.08P - Inservice Inspection Activities (PWR)

The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk-significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined and accepted by reviewing the following activities from November 13 to November 28.

PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===

The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:

(1) Ultrasonic Examination (UT)

• Reactor Vessel Closure Head (RVCH) Control Rod Drive Mechanism (CRDM)

Penetration Nozzle 1

• RVCH CRDM Penetration Nozzle 26

• RVCH CRDM Penetration Nozzle 52

• Auxiliary Head Adapters/Upper Head Injection (AHA/UHI) penetration B

• CRDM Nozzle 63 Category B-O Dissimilar Metal Weld

PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection

Activities (IP Section 03.02) (1 Sample)

The inspectors verified that the license conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:

(1)

• Ultrasonic Examination:

o Reactor Vessel Closure Head (RVCH) Control Rod Drive Mechanism (CRDM) Penetration Nozzle 1 o RVCH CRDM Penetration Nozzle 26 o RVCH CRDM Penetration Nozzle 52

PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)

The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:

(1)

• Boric Acid Walkdown - November 14, 2023

• CR 1866772

• CR 1866775

PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04) (1 Sample)

The inspectors verified that the licensee is monitoring the steam generator tube integrity appropriately through a review of the following examinations:

(1)

• Eddy Current Examination (ECT):

o Steam Generator 2 - ECT for tubes R80C50, R81C56, R84C95, R92C93, R105C62

71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance

Requalification Examination Results (IP Section 03.03) (1 Sample)

The licensee completed the annual requalification operating examinations and biennial written examinations required to be administered to all licensed operators in accordance with Title 10 of the Code of Federal Regulations 55.59(a)(2), "Requalification Requirements," of the NRC's "Operator's Licenses." The inspector performed an in-office review of the overall pass/fail results of the individual operating examinations, the crew simulator operating examinations, and the biennial written examinations in accordance with Inspection Procedure (IP) 71111.11, "Licensed Operator Requalification Program and Licensed Operator Performance." These results were compared to the thresholds established in Section 3.03, "Requalification Examination Results," of IP 71111.11.

(1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam completed on October 6, 2023 and the biennial written examinations completed on August 29, 2023.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (4 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) Discovery of an unplanned breach between the auxiliary building and the annulus on Unit 2 due to degraded seal on elevation 757' containment air lock as identified on CR 1894289, on November 25, 2023.

(2) Unit 2 turbine driven AFW steam leak on AFW turbine stop valve as documented under CR 1894442 and repaired under Work Order (WO) 124137309, on November 27, 2023.

(3) 2-RM-90-130 Containment purge radiation monitor instrument malfunction documented under Condition Report (CR) 1896682 on December 06, 2023

(4) Ice condenser maintenance rule scoping and evaluation of performance criteria December 12, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (3 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) Unit 2 shutdown activities on November 4, 2023

(2) First week of U2R5 the week of November 6, 2023

(3) Unit 1 Risk during ERCW A Header work from November 13, 2023 through November 15, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (7 Samples)

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

(1) 1A centrifugal charging pump balancing drum flow high under CR 1884541, on October 05, 2023

(2) Unit 2 entry into 2-AOI-44 for 2-R-6 failure under CR 1885009, on October 10, 2023

(3) FCV-0-32-85 failed to meet stroke time during 0-SI-32-902-B under CR 1884776, on October 10, 2023

(4) Lower containment exhaust isolation valve failed LLRT as documented in CR

===1889590, on November 01, 2023

(5) WBN-2-ISV-078-0600 fuel transfer tube Isolation failed during stroke open under CR 1890684, on November 6, 2023

(6) WBN-2-MVOP-003-0033-A SG #1 Feedwater Isolation Valve under CR 1891597, on November 10, 2023

(7) Containment spray heat exchanger 2A-A ERCW return valve exceeded torque and running load limits as documented under CR 1892908, on November 16, 2023

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01)===

(1) The inspectors evaluated Unit 2 refueling outage five activities from November 3, 2023, to November 28, 2023.

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) (7 Samples)

(1) Repairs to Door C036 as documented under CR 1891904 and the associated minor maintenance WO, on November 12, 2023

(2) N44 power range nuclear instrument after replacement of gain potentiometer under WO 123468573 and WO 123181003, on November 15, 2023

(3) D-A ERCW Check Valve replacement PMT under WO 123572239, on November 15, 2023

(4) RHR full flow testing for work done under WO 123166303, on November 16, 2023.

(5) Flow balancing of 2A-A safety injection pump loop flows per 2-SI-63-903 under WO

===123180639, on November 19, 2023

(6) PMT of SI Check Valves 2-FCV-63-157 and 2-FCV-63-165 per 2-SI-63-915-B under WO 123180769, on November 22, 2023

(7) Turbine Driven AFW Pump testing per 2-SI-3-902 under WO 123179843, on November 27, 2023

Surveillance Testing (IP Section 03.01)===

(1) 18-month channel calibration containment pressure channel II under WO 123425815, on October 11, 2023

(2) 0-SI-82-5 Loss of Offsite Power with Safety Injection Test DG 2A-A under WO

===123180873, on November 07, 2023

(3) 2-SI-63-907 Residual Heat Removal HL and CL Injection Check Valve Testing During Refueling Outages under WO 123180653, November 9, 2023

(4) 2-SI-63-905 Boron Injection Check Valve Flow Test During Refueling Outages under WO 12380636, on November 19, 2023

(5) 2-SI-62-906 Check Valve Testing During Refueling Outages - Chemical Volume Control System under WO 123180626, on November 21, 2023

Inservice Testing (IST) (IP Section 03.01)===

(1) 2-CKV-67-575D failed as left LLRT as documented in CR 1892759, was repaired and retested under WO 124122038

Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)

(1) 2-CKV-26-1260 gross failure as documented under CR 1893100, was repaired and retested under WO 124129181

Ice Condenser Testing (IP Section 03.01) (1 Sample)

(1) 2-SI-61-9 Ice Condenser Drain Check Valve Testing under WO 123180158, on November 21, 2023

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 October 5, 2023. This drill involved elevated reactor coolant system (RCS) vibrations resulting in a small break loss-of-coolant accident, failure of the 1B emergency diesel generator (EDG), and failure of the A train of containment spray. These conditions lead to a release and a general emergency declaration.

OTHER ACTIVITIES - BASELINE

===71151 - Performance Indicator Verification

The inspectors verified licensee performance indicators submittals listed below:

IE01: Unplanned Scrams per 7000 Critical Hours Sample (IP Section 02.01)===

(1) Unit 1 (April 1, 2022, through September 30, 2023)

(2) Unit 2 (April 1, 2022, through September 30, 2023)

IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)

(1) Unit 1 (April 1, 2022, through September 30, 2023)

(2) Unit 2 (April 1, 2022, through September 30, 2023)

IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)

(1) Unit 1 (April 1, 2022, through September 30, 2023)

(2) Unit 2 (April 1, 2022, through September 30, 2023)

MS05: Safety System Functional Failures (SSFFs) Sample (IP Section 02.04) (2 Samples)

(1) Unit 1 (April 1, 2022, through September 30, 2023)

(2) Unit 2 (April 1, 2022, through September 30, 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) Part 21 Screening and evaluations as documented in CR 1884685, 1880481,

===1884702, 1888156, and 1844718. The licensee documented gaps in the Part 21 screening process; these gaps are being addressed under CR 1888590 and CR 1888574.

(2) CR 1867666 documented a temporary modification that was not controlled per licensee processes. Evaluations of the impacts were performed, and WO 119685686 is scheduled to remove the temporary modification.

(3) CRs 1855720 & 1885751 documented questions the NRC had regarding Surveillance Instructions 1/2-SI-61-3, Unit 1/2 Ice Condenser Flow Passages Inspection.

71152S - Semiannual Trend Problem Identification and Resolution

Semiannual Trend Review (Section 03.02)===

Perform a semiannual review of licensee PI&R program documents to identify potential trends that might indicate the existence of a more significant safety issue.

(1) During the period of June to December, 2023, the inspectors reviewed the licensees corrective action program, which included inspector-identified issues, for potential adverse trends in the implementation of the Watts Bar fire protection program that might be indicative of a more significant safety issue.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Follow up (IP Section 03.01)

(1) The inspectors evaluated licensee response to notification of unusual event on October 21, 2023, EN 56809, CR

INSPECTION RESULTS

Failure to Implement Inservice Test Requirements in Accordance with ASME OM Code Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.14] - 71111.24 Systems NCV 05000390,05000391/2023004-01 Conservative Open/Closed Bias The inspectors identified a Green finding and associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(f), for the licensee's failure to implement inservice testing (IST) requirements. The licensee failed to properly implement testing procedures for Essential Raw Cooling Water (ERCW) pump discharge check valve 0-CKV-67-503D, and did not take proper corrective action following discovery of unacceptably degraded valve internals in accordance with the 2004 edition through the 2006 addenda of the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), as incorporated by reference in 10 CFR 50.55a, which is the current IST Program Code of Record for Watts Bar.

Description:

During a review of Train A ERCW pump comprehensive testing, the NRC inspectors identified failures of the licensee to follow the ASME OM Code IST requirements for ERCW pump discharge check valve 0-CKV-67-503D.

The Watts Bar Updated Final Safety Analysis Report (UFSAR), section 9.2.1.1, describes the design bases for the ERCW system. Section 9.2.1.1 states, The ERCW system is safety-related because it provides essential auxiliary support functions to the engineered safety features of the plant. The system is designed to supply cooling water to safety and non-safety-related equipment. Provisions are made to ensure a continuous flow of cooling water to those systems necessary for plant safety either during normal operation or under accident conditions.

On May 9, 2022, the licensee deferred completion of preventative maintenance on check valve 0-CKV-67-503D, for ERCW pump D-A. The preventative maintenance change request (PMCR 1510346) documenting the deferral identified that leakage through check valve 0-CKV-67-503D, was affecting ERCW pump performance. Additionally, the PMCR stated that leakage through check valve 0-CKV-67-503D, could result in other pumps on ERCW Train A (A-A, B-A, C-A) being declared inoperable and unable to provide sufficient flow to components cooled by ERCW.

On May 23, 2023, ERCW pumps B-A and C-A failed to meet the pump flow test acceptance criteria of 0-SI-67-917-A, and 0-SI-67-918-A, respectively. On May 24, 2023, leakage across check valve 0-CKV-67-503D, was determined to be the primary contributor to the failure of the pump tests. These test results confirmed that the D-A ERCW discharge check valve was degraded to the point that the system performance was challenged. This was documented in CR 1858712.

The 2004 Edition through the 2006 Addenda of the ASME OM Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraph ISTC-5221, Valve Obturator Movement, states, in part, Check valves that have a safety function in both the open and close directions shall be exercised by initiating flow and observing that the obturator has traveled to either the full open position or to the position required to perform its intended function(s), and verify that on cessation or reversal of flow, the obturator has traveled to the seat.

ASME OM Code, Subsection ISTC, paragraph ISTC-5224, Corrective Action, states, in part, that if a check valve fails to exhibit the required change of obturator position, it shall be declared inoperable. Paragraph ISTC-5224 also states that a retest showing acceptable performance shall be run following any required corrective action before the valve is returned to service.

During troubleshooting efforts, the licensee failed to implement ASME OM Code (2004 Edition through 2006 Addenda), Subsection ISTC, paragraph ISTC-5224 after determining that the D-A ERCW check valve did not meet acceptance criteria for demonstrating proper obturator movement. The licensee documented these observations in the Train A ERCW Testing Sequence Summary. A train ERCW performance was evaluated by the licensee under CR 1858712, and determined that no loss of safety function occurred.

Corrective Actions: Licensee documented system impact and evaluation under CR 1858712, and the Check Valve was replaced under WO 123572239 on November 15, 2023.

Corrective Action References: CR 1858712

Performance Assessment:

Performance Deficiency: The failure to recognize tested component failure of the D-A ERCW discharge check valve 0-CKV-67-503D and take corrective action per ASME OM Code, Subsection ISTC, paragraph ISTC-5224, was a performance deficiency.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance 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 follow testing requirements for check valve 0-CKV-67-503D adversely impacted the ability of the ERCW system to provide cooling to systems needed for safe shutdown. Additionally, further degradation of the leakage through check valve 0-CKV-67-503D would result in the ERCW Train A (A-A, B-A, C-A) being declared inoperable and unable to provide sufficient flow to components cooled by ERCW.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors assessed the significance of the finding using Inspection Manual Chapter 0609, 04, Initial Characterization of Findings, and determined that the finding was associated with the Mitigating Systems cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, Section A, the screening questions, were all answered a NO; therefore, the inspectors determined the finding was of very low safety significance (Green).

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Previously failed pump tests were reperformed without correcting issues related to degraded valve performance.

Enforcement:

Violation: 10 CFR 50.55a(f)(4), Inservice testing standards requirement for operating plants, requires in part that, Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements (except design and access provisions) set forth in the ASME OM Code and addenda that become effective subsequent to editions and addenda specified in paragraphs (f)(2) and

(3) of this section and that are incorporated by reference in paragraph (a)(1)(iv) of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components. The current IST Program Code of Record for Watts Bar is the 2004 Edition through the 2006 Addenda of the ASME OM Code, as incorporated by reference in 10 CFR 50.55a, which requires in Subsection ISTC, paragraph ISTC-5224, that, If a check valve fails to exhibit the required change of obturator position, it shall be declared inoperable. A retest showing acceptable performance shall be run following any required corrective action before the valve is returned to service.

Contrary to the above, from June 23, 2023, to November 15, 2023, when pump tests revealed that significant leakage past check valve 0-CKV-67-503D could result in an inoperable ERCW train, the licensee failed to take action in accordance with ASME OM Code, Subsection ISTC, paragraph ISTC-5224, and failed to follow the provisions of paragraph ISTC-5224 to correct the degraded valve condition.

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

Failure to Identify Condition Adverse to Quality Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Pending [P.2] - 71111.20 AV 05000391/2023004-02 Evaluation Open The inspectors identified a finding with its safety significance as yet to be determined and associated apparent violation (AV) of 10 CFR 50 Appendix B Criterion XVI, Corrective Actions, for failure to promptly identify and correct the impact of Unit 2 feedwater leakage on ice condenser lower inlet doors. This condition led to the accumulation of ice on the lower inlet doors challenging their ability to open.

Description:

From January 19, 2023, to November 4, 2023, the licensee was operating under an Adverse Condition Monitoring Plan due to excessive ice buildup on the intermediate deck doors caused by a feedwater leak in lower containment. The licensee recognized that continued operations with the feedwater leak was challenging the aspects of operation of the Unit 2 ice condenser; however, the licensee did not identify any impacts to the opening capability of the lower inlet doors.

On November 5, 2023, licensee staff performed walkdowns of the lower ice condenser and identified adverse conditions (glycol leakage and flow passage ice buildup), but they failed to identify ice accumulation on the lower inlet doors.

On November 5, 2023, after the licensee walkdown, the resident office performed an independent walkdown and identified excessive ice buildup in the lower ice condenser impacting the ability of at least eight lower inlet doors to open.

UFSAR Section 6.7.8.1 4. A. Design Criteria - Accident Conditions states, in part, that all doors open to allow venting of energy to the ice condenser for any leak rate which results in a divider deck differential pressure in excess of the ice condenser cold head.

TS SR 3.6.12.3 requires, in part, to verify by visual inspection that the ice condenser inlet door is not impaired by ice, frost, or debris. Watts Bar staff perform this surveillance as part of unit startup activities coming out of a refueling outage.

Corrective Actions: The inspectors immediately informed the licensee staff of the apparent, non-conforming observations on November 5, 2023. The licensee documented the concern in CR 1893269 on November 18, 2023.

Corrective Action References: CR 1893269

Performance Assessment:

Performance Deficiency: Failure of the licensee to identify a condition adverse to quality is the performance deficiency reasonably within the licensees ability to foresee and correct.

Specifically, the licensee failed to identify buildup of ice on the lower inlet doors challenging their safety function. Further, the licensee failed to ensure that the adverse condition was fully evaluated in a timely manner to ensure that there were no questions regarding the operability of the lower inlet doors.

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 Blockage of the lower inlet 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.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix H, Containment Integrity SDP. 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 Blockage of more than 15% of the flow passages into or through the ice bed for an exposure period of >3 days results in an initial Risk Significance of Greater than Green and requires Detailed Risk Evaluation.

Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. The licensee failed to perform a thorough evaluation of the as found condition, resulting in a required second evaluation that failed to provide technical evaluation of the as found condition.

Enforcement:

Violation: 10 CFR 50 Appendix B Criterion XVI requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected.

From January 19, 2023, to November 5, 2023, TVA at Watts Bar failed to establish measures to assure prompt identification of conditions adverse to quality associated with the impact of feedwater leakage on ice condenser lower inlet doors. This condition leads to accumulation of ice on the lower inlet doors challenging their ability to open.

TS 3.6.12 Limiting Condition for Operation, requires that the ice condenser inlet doors, intermediate deck doors, and top deck doors shall be operable and closed.

Contrary to the above, from January 19, 2023, to November 5, 2023, the licensee failed to ensure that the lower inlet doors remained operable. Specifically, the licensee allowed the accumulation of ice on and about the ice condenser lower inlet doors challenging their capability to open during a design basis event.

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

Failure to implement Surveillance Requirement 3.6.12.2 Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.11] - 71111.24 NCV 05000391/2023003-05 Challenge the Closed Unknown EA-23-130 The inspectors identified a Green finding and an associated NCV of Technical Specification (TS) 3.6.12, Ice Condenser Doors, for the licensees failure to implement the requirements of surveillance instruction 2-SI-61-6, Weekly Ice Condenser Intermediate Deck Doors (IDD)

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

Description:

On August 10, 2023, the resident office identified issues with the as left pull testing under 2-SI-61-6. Specifically, the licensee failed to properly perform as-left testing of the ice condenser IDD. During further review the inspectors identified that the MT&E that was utilized for the surveillance did not meet the environmental requirements specified in 1-SI-61-6 and 2-SI-61-6. Moreover, the force gauge was not rated for the 10°F environment of the Ice Condensers.

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; this CR was screened as a non-corrective action program item. Licensee staff completed the out-of-tolerance investigation for the affected WOs and determined that there was a substantive challenge to operability during the period under investigation.

Corrective Actions: Licensee performed an immediate operability review and performed a past operability evaluation, the result of the review required additional evaluation by Westinghouse to provide additional analysis under WAT-D-13092 on October 17, 2023 as documented in CR 1880146.

Corrective Action References: 1874466, 1876542, 1880146

Performance Assessment:

Performance Deficiency: Failure to implement the requirements of 2-SI-61-6, Weekly Intermediate Deck Doors Visual Inspection, was a performance deficiency. Specifically, the licensee failed to ensure that adequate M&TE was utilized to provide reasonable assurance that the IDDs 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 M&TE was utilized to provide reasonable assurance that the IDDs remained operable.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix H, Containment Integrity SDP. Screens to Green on Table 7.2 due to the blockage being less than 15% of the flow passage into or through the ice bed.

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 TS 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 2-SI-61-6, states, in part, that performance of this instruction satisfies the following: SR 3.6.12.2.

Section 4.2 [2] of 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 was not rated for use in a 10°F environment and has subsequently failed post-use calibration. The failure of the M&TE has substantively challenged operability of the ice condenser intermediate deck doors. Licensee initiated CR 1880146 on September 13, 2023, documented the WOs that were impacted in which intermediate deck doors did not meet acceptance criteria.

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

Failure to Maintain N High Pressure Flex Capability Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.14] - 71152A Systems NCV 05000390,05000391/2023003-06 Conservative Closed Bias EA-23-131 A self-revealed Green finding and associated NCV of 10 CFR 50.155, Mitigation of Beyond Design Basis Events, was identified for the licensees failure to implement the requirements of 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 HP FLEX Pumps the licensee discovered that the HP FLEX pump hoses had the incorrect fittings. Specifically, the fittings available onsite 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. 0-TI-446, defines N capability as 2 High Pressure Flex Pumps, hoses, and fittings to connect the pumps to plant systems.

On June 22, 2023, CR 1863956 documented the need to develop compensatory measures for the inability to implement the Phase 2 mitigation strategy. Beyond the initial compensatory measures which were found to be unable to be implemented. The initial projected receipt of the fittings was September 6, 2023; however, delays in the receipt of the fittings were not communicated to site personnel, resulting in further delays. Fittings were received on site on October 16, 2023.

Corrective Actions: Licensee documented the issue in the Corrective Action program, and the fittings were procured and tested under WO 122993416, and WO 122993368.

Corrective Action References: CR 1862048

Performance Assessment:

Performance Deficiency: The licensees failure to implement the requirements of 0-TI-446 and 10 CFR 50.155, 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 licensee failed to maintain the capability to implement the bey

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. 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. A detailed risk evaluation was performed by a regional Senior Reactor Analyst using Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.2.9 and NRC Watts Bar Standardized Plant Analysis Risk (SPAR) model Version 8.80. The SPAR model was adjusted to reflect Watts Bar Flexible Coping Strategy (FLEX) implementation for both external flood and non-flood scenarios. A conditional analysis was performed to evaluate the risk increase due to the unavailability of appropriate fittings for the high-pressure FLEX Reactor Coolant System (RCS) makeup pumps. A condition exposure period of 0.627 years was used, and no credit was provided in the analysis for post-failure recovery of equipment impacted by performance deficiency. The dominant sequences involved external flood initiators followed by failure of the FLEX high pressure RCS makeup pumps which was accompanied by the failure of operator action to successfully align the auxiliary charging pumps. The performance deficiency did not impact the availability of the auxiliary charging pumps which mitigated the overall risk of the finding. The analysis determined that the estimated increase in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) was less than 1E-06/year for delta-CDF and less than 1E-07/year for delta-LERF, representing a finding of very low safety significance (GREEN) for Unit 1 and Unit 2.

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. 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 10CFR50.155 (b)(1), as implemented by the licensee in accordance with 0-TI-446 section 1.5.2, did not have sufficient capability to perform the functions required. Specifically, following loss of the required couplings, the licensee lost 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 a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure of Surveillance Instruction 1/2-SI-61-3, Unit 1/2 Ice Condenser Flow Passage Inspection, to include appropriate acceptance criteria for determining that a Surveillance Requirement was met Cornerstone Significance Cross-Cutting Report Aspect Section Barrier Integrity Green [H.11] - 71152A NCV 05000390,05000391/2023004-03 Challenge the Open/Closed Unknown The inspectors identified a Green finding and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the licensee s failure to include appropriate quantitative or qualitative acceptance criteria in Surveillance Instruction 1/2-SI-61-3, Unit 1/2 Ice Condenser Flow Passages Inspection could result in TS 3.6.11 not being met.

Description:

On April 28, 2023, Work Order (WO) 122771578 documented surveillance instruction 1-SI-61-3, Unit 1 Ice Condenser Flow Passages Inspection for the Unit 1 cycle 18 refueling outage and was signed complete on May 6, 2023.

Section 6.1, Notes 3) of 1-SI-61-3 states in part, "for evaluation purposes, the blockage in each flow passage shall be graded to be 0%,

25%, 50%, 75%, or 100%." These percentages are then used to calculate total flow passage blockage which is ultimately compared to Surveillance Requirement (SR) 3.6.11.4.

SR 3.6.11.4 states, "Verify, by visual inspection, accumulation of ice on structural members comprising flow channels through the ice bed is less than or equal to 15 percent blockage of the total flow area for each safety analysis section."

During testing, the performer positions themselves at one end of a single flow passage. They, then, cast their flashlight, the only source of light, into the 40-foot-long irregular void created by the hanging ice baskets, their structural supports, and any ice accumulation. The inspectors observed that both 1-SI-61-3 and 2-SI-61-3, Unit 2 Ice Condenser Flow Passages Inspection, provided no standard, example, training, or qualification to ensure that the skill of the craft performing the task was adequate to quantitatively determine how much blockage was present in each flow passage on a scale from 0% to 100% at 25% increments.

Therefore, the inspectors determined that the surveillance instructions lacked appropriate quantitative or qualitative acceptance criteria represented a substantive challenge to determining that the unit 1 and 2 Surveillance Requirement (SR) 3.6.11.4 were met.

Corrective Actions: The licensee documented the issue in the corrective action program and established a required pre-test briefing to both 1-SI-61-3 and 2-SI-61-3. The pre-test briefing in addition to further procedure improvements ensure adequate skill of the craft performing 1-SI-61-3 and 2-SI-61-3.

Corrective Action References: CR 1855720, 1885751

Performance Assessment:

Performance Deficiency: Failure of 1-SI-61-3 and 2-SI-61-3 to include appropriate quantitative or qualitative acceptance criteria for determining that unit 1 and 2 Surveillance Requirement (SR) 3.6.11.4 were met, was a performance deficiency reasonably within the licensee's ability to foresee and prevent.

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. Lacking any standard, example, training, or qualification to ensure that the skill of the craft could determine compliance with SR 3.6.11.4, 1-SI-61-3 and 2-SI-61-3 represented a substantive challenge to the operability of the ice condenser.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix H, Containment Integrity SDP. 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 Blockage was less than 15 percent therefore the issue screens to Green.

Cross-Cutting Aspect: H.11 - Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. If a procedure or work document is unclear or cannot be performed as written, individuals stop work until the issue is resolved. Specifically, licensee craft performing 1-SI-61-3 and 2-SI-61-3 should have stopped work up recognizing the surveillance instruction lacked proper acceptance criteria.

Enforcement:

Violation: 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Watts Bar Unit 1 & 2 Technical Specification 3.6.11, Ice Bed, states, The ice bed shall be OPERABLE.

SR 3.6.11.4 states, Verify, by visual inspection, accumulation of ice on structural members comprising flow channels through the ice bed is less than or equal to 15 percent blockage of the total flow area for each safety analysis section.

Section 1.2.2 of Licensee Surveillance Instruction 1-SI-61-3 & 2-SI-61-3, states, in part, that:

performance of this instruction satisfies Surveillance Requirement (SR) 3.6.11.4.

Section 6.1, Notes 3) of Licensee Surveillance Instruction 1-SI-61-3 states in part, for evaluation purposes, the blockage in each flow passage shall be graded to be 0%, 25%,

50%, 75%, or 100%.

Contrary to the above, until October 2023, the licensee failed to ensure that Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.

Specifically, the licensee failed to include appropriate quantitative or qualitative acceptance criteria for determining that important activities such as if Surveillance Requirements had been met. 1-SI-61-3 and 2-SI-61-3 instructs the craft to grade the blockage for each flow passage but does ensure by direction or experience that the craft has the skill to make this determination.

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

Licensee-Identified Non-Cited Violation 71152A This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: EA-23-104 Unit 2 TS 5.7.1.1 a., requires, in part, that written procedures shall be implemented for the applicable procedures recommended in Regulatory Guide (RG) 1.33 Revision 2, Appendix A February 1978. Section 3.f of Appendix A of RG 1.33 required in part, procedures for maintaining containment integrity. Procedure 2-TI-68.002, Containment Closure Control implements penetration closure requirements to establish containment integrity. 2-TI-68.002, requires in part that when breaching the containment during modes 5 and 6, the breach is evaluated to ensure the breach can be closed within an allowed closure time and an owner is assigned to close the breach if containment closure is required.

Contrary to the above, from March 4, 2022, until June 20, 2022, 2-TI-68.002 was not implemented in that a breach of containment penetration X-36 existed and was not evaluated and no owner responsible to close the penetration was assigned.

Significance/Severity: Green.

The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. Using Attachment 1 "Phase 1 Initial Screening and Characterization of Findings," Exhibit 4 "Barrier Integrity Screening Questions" the inspectors were directed to IMC 0609 Appendix H "Containment Integrity Significance Determination Process, because the finding degraded the physical integrity of reactor containment (valves, penetrations, containment isolation components). Using IMC 0609 Appendix H, since the finding only affects LERF, and not CDF, it is considered a Type B shutdown finding and is evaluated per Section 07.02, Approach for Assessing Type B Findings at Shutdown. Per Step 2.1, the finding occurred within eight days of the outage in POS 2, so the inspectors continued to Step 2.2. Containment status was determined to be intact, based on Note 1 of Table 7.3, Phase 1 Screening-Type B Findings at Shutdown. (An intact containment is one in which, the licensee intends to:

(1) close all containment penetrations with a single barrier or can be closed in time to control the release of radioactive material, and

(2) maintain the containment differential pressure capability necessary to stay intact following a severe accident at shutdown. A Type B performance deficiency results when a licensee intends to have an intact containment but cannot maintain that capability due to a performance deficiency). Step 2.2.A directs the inspectors to use Table 7.3. Table 7.3 directs a phase II evaluation to be performed.

Because the issue did not screen to green in phase II, a phase III detailed risk assessment (DRE) using plant specific values, actual exposure times for this case, potential release paths from this penetration, and shutdown risk models is required. A regional Senior Reactor Analyst (SRA) coordinated with Idaho National Laboratories (INEL) and NRCs NRR Division of Risk Assessment (DRA) to create a temporary limited use shutdown PRA model for Watts Barr Units 1 and 2 dated 12/06/23. The DRE was performed for the time the plants was considered to be in Plant Operating State (POS) 2. The exposure period was broken into three segments 1) From when penetration x-36 was opened until the reactor vessel head was removed (27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />). 2) From the reactor vessel head being removed unless the refueling cavity was flooded up to greater than 21 feet above the reactor vessel flange (10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />), and 3) from the time the refueling cavity was flooded up to 21 feet above the reactor vessel head until the reactor vessel upper internals were removed and the plant is considered to be in POS 3 (11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />). Major assumptions include that in exposure period 1 gravity feed was not available; however, FLEX mitigation strategies were credited for Station Blackout sequences.

For exposure period 2, gravity feed was credited since the procedures directed the line up to be used. For exposure period 3, on Loss of Inventory shutdown events were considered due to the excess volume of water in the refueling cavity. This finding was considered to be a Type B finding only affecting Large Early Release Frequency (LERF) and delta LERF was calculated in accordance with IMC 03 appendix H, Containment Integrity Significance Determination Process Technical Basis. LERF = FB x FR x CDF x (multiplier for the duration of degraded condition) where: FB = increase in the conditional large early release probability resulting from the finding, and FR = fraction of the CDF for which the containment failure mode affected by the finding impacts LERF. The multiplier is the fraction of time in a shutdown year (IE exposure time). Long term accident sequences. FB was approximated to be 0.8. FR was approximated by removing all the cutsets which would result in a late onset of core damage. To do this the SRA ran a conditional case for each exposure period, setting term 2-SD-XHE-XL-SDC-3D LATE RECOVERY OF SDC - 3 DAYS to false.

This resultant CCDP is approximately equivalent to baseline CDF x FR. Removing sequences involving long term cooling permitted in Appendix H.

The dominant accident sequence was a Mode 5 Loss Of Offsite Power event in Exposure Period 1, common cause failure of the EDGs, and common cause failure of all FLEX EDGs.

The resulting change in LERF was less than 1 E-7 corresponding to a finding of very low safety significance (GREEN).

Corrective Action References: CR 1784613, WO 122641445

This LIV closed the EA-23-104 associated with AV 05000391/2023010-02, Licensee -

Identified Uncontrol opening of penetration X-36 results in Unrecognized Loss of Containment Closure.

Assessment 71152S Assessment of Semiannual Trend of Watts Bar Fire Protection Program

The inspectors identified an adverse trend in the licensees control of transient combustible materials as a result of identifying multiple examples over a short period where personnel failed to follow the requirements of procedure NPG-SPP-18.4.7, Control Transient Combustibles. These issues are documented in the minor violation results section of this report. The inspectors considered it noteworthy that all the issues were NRC-identified, and there were only four similar licensee-identified transient combustible issues that were found during the inspectors search of the licensees corrective action program during this six-month period of review. The following CRs associated with the identified issues are referenced in the minor violation results section of this report: CRs 1893014, 1884434, 1870684, 1867666, 1860713, and 1861737.

The inspectors noted that the licensees corrective actions for these items included the immediate correction of the identified discrepancies as well as disseminating crew learnings to site services as documented in CR 1885568 to reinforce the need for personnel to be more sensitive to adhering to proper transient combustible and housekeeping practices. The licensee acknowledged the adverse trend and entered it into their corrective action program as CR 1885568.

Minor Violation 71152S Minor Violations for Transient Combustible Controls

Minor Violation: The inspectors reviewed the licensees corrective action program, which included inspector-identified issues, between the period of June 2023 to December 2023, for potential adverse trends in the implementation of the Watts Bar fire protection program that might be indicative of a more significant safety issue.

The inspectors identified an adverse trend in the licensees control of transient combustible materials as a result of identifying multiple examples, where personnel failed to follow the requirements of procedure NPG-SPP-18.4.7, Control Transient Combustibles.

The inspectors considered it noteworthy that the majority of the issues were NRC-identified during this six-month period of review. The following CRs associated with the identified Minor Violations are: CRs 1893014, 1884434, 1870684, 1867666, 1860713.

Per NPG-SPP-18.4.7 defines Fire Risk Areas requirements applicable to all fire areas.

CR 1860713 documents an inspector identification of transient combustibles exceeding the allowance in the common Cable Spreading Room without and evaluation. The licensee performed an evaluation of the transient combustibles and with the updated evaluation determined that the transient combustibles were acceptable.

CR 1867666 documents an inspector identification of transient combustibles in the Shutdown Board Rooms without an evaluation. These transient combustibles did not meet the spacing/separation requirements of NPG-SPP-18.4.7. The licensee performed an evaluation of the transient combustibles and determined that the additional materials were acceptable.

CR 1870684 documents an inspector identification of transient combustibles in the 480V transformer rooms and the Mechanical Equipment Rooms on elevation 772 that did not meet the spacing/separation requirements of NPG-SPP-18.4.7 without an evaluation. Licensee evaluation of the transient combustibles determined that the materials were acceptable.

CR 1884434 documents an inspector identification of transient combustibles exceeding the limits in the Emergency Gas Treatment Filter Room without an evaluation. The licensee performed an evaluation of the transient combustibles and determined that the additional materials were acceptable.

CR 1893014 documents an inspector identification of transient combustibles that did not meet the spacing/separation requirements of NPG-SPP-18.4.7 without an evaluation. The licensee performed an evaluation of the transient combustibles and determined that the additional materials were acceptable.

In the above examples the licensee failed to implement the requirements of NPG-SPP-18.4.7 and the non-compliance was discovered by the inspectors. The licensee documented each of these issues in the corrective action program and most of the issues were screened as conditions adverse to quality. The licensee misclassified two of the issues initially, the licensee documented the misclassifications in follow-up CRs 1870930 and 1883420.

Screening: The inspectors determined the performance deficiency was minor. Unit 1 and Unit 2 Technical Specifications 5.7.1.1 d., requires, in part, that written procedures shall be implemented for the Fire Protection Program implementation. Specifically, NPG-SPP-18.4.7 Control of Transient Combustibles, a quality related procedure, requires that the limits for transient combustibles be controlled in accordance with Sections 3.2.1, 3.2.2, 3.3.1 and 3.4.

These controls included limiting the transient combustibles to amounts specified based on fire risk and ensuring adequate spacing and separation between transient combustibles and plant equipment. or to obtain a Transient Combustible Permit/Evaluation form with appropriate compensatory measures.

Contrary to the above, from June 2023 to December 2023, the licensee failed to implement all provisions of the approved fire protection program. Specifically, the licensee failed to control transient combustibles in accordance with Control of Transient Combustibles procedure Sections 3.2.1, 3.2.2, 3.3.1, and 3.4; or to obtain a Transient Combustible Permit/Evaluation form with appropriate compensatory measures.

Enforcement:

This failure to comply with Unit 1 and Unit 2 Technical Specification 5.7.1.1 d.

constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.

Licensee-Identified Non-Cited Violation 71153 This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Violation: EA-23-129 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.

Significance/Severity: Green. 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 Attachment 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 finding could not be screened to Green and a DRE is required.

The inspectors assessed the significance of the finding using IMC 0609, Appendix F, Fire Protection Significance Determination Process. A Phase 2 quantitative screening approach was necessary because the finding represented a high level of degradation due to the missing fire barrier, which impacted the function of the credited safe shutdown power supply, and the area of the missing fire barrier was not fully covered by detection or suppression systems. A regional Senior Reactor Analyst used Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8.2.9 and NRC Watts Bar Standardized Plant Analysis Risk (SPAR) model Version 8.80 to perform a Phase 2 screening and walked down the affected fire area the week of September 11, 2023. The SPAR model was adjusted to reflect Watts Bar Flexible Coping Strategy (FLEX) implementation for Station Blackout (SBO) scenarios. The Phase 2 conditional core damage probability associated with the performance deficiency was estimated to potentially be greater than 1E-06, therefore, a detailed risk evaluation was required. The analyst noted that the licensee had a fire Probabilistic Risk Assessment (PRA) model of record that was pending an external peer review and compared the results of the licensees analysis of the condition using the model of record with the preliminary screening SPAR model results. The analyst observed that prior walkdowns had identified credible ignition sources and physical arrangement of the affected cables and suppression systems that were consistent with those considered in the licensees fire PRA analysis. The estimated risk was mitigated by the relatively large size of the room, the low number of ignition sources that were near the affected cables, and the by consideration of the suppression systems capable of mitigating fires from those ignition sources (which were not credited in the screening analysis). A Significance Determination Process (SDP) maximum condition exposure period of one year was used, and no credit was provided for post-failure repair of equipment impacted by performance deficiency. The SPAR model dominant sequences involved fire initiators with random failures of FLEX equipment including the 6.9kV FLEX diesel generators and pumps necessary for decay heat removal.

The detailed risk analysis determined that the estimated increase in Core Damage Frequency (CDF) and Large Early Release Frequency (LERF) was less than 1E-06/year for delta-CDF and less than 1E-07/year for delta-LERF, representing a finding of very low safety significance (GREEN) for Unit 2.

Corrective Action References: CR 1858921

This LIV closed the EA-23-129 associated with AV 05000390,05000391/2023003-07, Licensee identified Unanalyzed Condition for the 2A-A Emergency Diesel Generator.

EXIT MEETINGS AND DEBRIEFS

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

• On January 31, 2024, the inspectors presented the integrated inspection results to Anthony Williams, Site VP and other members of the licensee staff.

• On November 29, 2023, the inspectors presented the Exit Meeting inspection results to Anthony Williams, Site VP and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.04 Drawings 2-47W810-1-Flow Diagram Residual Heat Removal System

ISI

71111.05 Corrective Action 1893709 NRC identified several issues which require resolution November 22,

Documents Resulting 2023

from Inspection

71111.05 Corrective Action 1898069 NRC identified unrestrained gang box in B ERCW pump December 13,

Documents Resulting room 2023

from Inspection

71111.05 Corrective Action 1898398 NRC identified during DG building walkdown that 0-DRV-26-December 15,

Documents Resulting 1079 is leaking water on the floor. 2023

from Inspection

71111.05 Corrective Action 1898402 NRC identified during DG building walkdown that Door D021 December 15,

Documents Resulting is not closing on its own 2023

from Inspection

71111.05 Corrective Action 1898411 NRC identified during DG building walkdown that there were December 15,

Documents Resulting ladders stored in front of and blocking emergency light 2023

from Inspection

71111.05 Fire Plans AUX-0-757-02 Pre-fire Plan Auxiliary Building Elevation 757 B Shutdown 005

Board Rooms

71111.05 Fire Plans AUX-0-757-03 Pre-fire Plan Auxiliary Building Elevation 757 Shutdown 004

Board Rooms A

71111.05 Fire Plans AUX-0-772-02 Pre-fire Plan Auxiliary Building Elevation 772 Motor 005

Generator Set Room and Pressurizer Heater Transformer

Room 2

71111.05 Fire Plans AUX-0-772-03 Pre-fire Plan Auxiliary Building Elevation 772 Pressurizer 005

Heater Transformer Room 1, and Motor Generator Set

Room

71111.05 Fire Plans AUX-757-04 Pre-fire Plan Auxiliary Building Emergency Gas Treatment 007

Room

71111.05 Fire Plans DGB-0-760-1 Pre-fire Plan Diesel Generator Building Elevation 760.5 004

71111.12 Corrective Action WBPER980424 Closure Report 08/06/1998

Documents

Inspection Type Designation Description or Title Revision or

Procedure Date

71111.12 Miscellaneous 1768 CDE Record 12/01/2021

71111.12 Miscellaneous WBPER980057 Meeting Minutes 04/07/1998

71111.12 Miscellaneous Westinghouse WAT-D-10547, WAT-D-10549, WAT-D-10564, WAT-D-

Letters 10686, WAT-D-10930, WAT-D-11822, WAT-D-11826, WAT-

D-11878, WAT-D-13092

71111.12 Procedures B 3.6 Tech Spec Bases

71111.12 Procedures Meeting No. 98-16, No. 98-22, No. 98-31

Minutes

71111.12 Procedures Surveillance 1-SI-61-1 thru 1-SI-61-9 & 2-SI-61-1 thru 2-SI-61-9

Instructions

71111.12 Procedures Technical TI-119 R4, TI-119 R6

Instructions

71111.12 Work Orders Work Order 124138483

71111.20 Corrective Action 1893269 NRC concern during walkdown of lower ice November 5, November 18,

Documents Resulting 2023 2023

from Inspection

71111.24 Procedures 0-SI-67-918-A Essential Raw Cooling Water Pump C-A and Pump D-A May 24, 2023

Comprehensive Pump Test

71111.24 Procedures 2-SI-63-907 Residual Heat Removal HL and CL Injection Check Valve Revision 18

Testing During Refueling Outages

71111.24 Procedures 2-SI-63-917 Testing of Cold Leg Accumulator Check Valves Revision 3

71111.24 Work Orders Work Orders 123452957, 123180636, 123180873, 123180769,

24080989, 123180639, 123180158, 119156701,

23180653, 123109038

71114.06 Miscellaneous 2023 WBN October Training Drill (Team D) 10/04/2023

71152A Procedures 1-SI-61-3 U1 Ice Condenser Flow Passages Inspection Revisions 13,

14, 15

71152A Procedures 2-SI-61-3 18 Month Ice Condenser Flow Passages Inspection Revisions 1,

2, 3

71152A Procedures NPG-SPP-Condition Report Initiation Revision 6

01.16

71152A Procedures NPG-SPP-01.2 Administration of Site Technical Procedures Revision 25

27