IR 05000259/2024001
| ML24131A141 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/14/2024 |
| From: | Louis Mckown NRC/RGN-II/DRP/RPB5 |
| To: | Jim Barstow Tennessee Valley Authority |
| References | |
| IR 2024001 | |
| Download: ML24131A141 (24) | |
Text
SUBJECT:
BROWNS FERRY NUCLEAR PLANT - INTEGRATED INSPECTION REPORT 05000259/2024001, 05000260/2024001, AND 05000296/2024001
Dear Jim Barstow:
On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Browns Ferry Nuclear Plant. On April 4, 2024, the NRC inspectors discussed the results of this inspection with Manu Sivaraman, site vice president, and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. The 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.
If you contest the violations or the significance or severity of the violations 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 Browns Ferry 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 Browns Ferry Nuclear Plant.
May 14, 2024 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. 05000259, 05000260, and 05000296 License Nos. DPR-33, DPR-52, and DPR-68
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000259, 05000260 and 05000296 License Numbers:
DPR-33, DPR-52, and DPR-68 Report Numbers:
05000259/2024001, 05000260/2024001 and 05000296/2024001 Enterprise Identifier:
I-2024-001-0019 Licensee:
Tennessee Valley Authority Facility:
Browns Ferry Nuclear Plant Location:
Athens, Alabama Inspection Dates:
January 01, 2024 to March 31, 2024 Inspectors:
S. Billups, Resident Inspector M. Magyar, Reactor Inspector A. Nielsen, Senior Health Physicist K. Pfeil, Resident Inspector J. Rivera, Health Physicist T. Steadham, Senior Resident Inspector Approved By:
Louis J. McKown, II, Chief Reactor Projects Branch 5 Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Browns Ferry 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 Promptly Repair a Failed Relay in the Standby Diesel Generator Start Circuit Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000296/2024001-01 Open/Closed
[P.2] -
Evaluation 71111.24 A self-revealed Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, was identified when the 3D standby diesel generator failed to start during testing due to a failed relay. Specifically, the relay was identified as degraded in January 2023, however, the licensee failed to properly evaluate the significance of the failure and consequently failed to take prompt corrective actions.
Failure to Evaluate Radiological Conditions Following the Transfer of Radioactive Sludge to a Shipping Liner Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000259,05000260,05000296/2024001-02 Open/Closed
[H.4] -
Teamwork 71124.01 A self-revealed Green finding and associated NCV of 10 CFR 20.1501, Surveys and Monitoring, was identified when the licensee failed to perform adequate surveys following a significant change in radiological conditions. Specifically, the licensee failed to evaluate radiological conditions in the 546-foot elevation condensate sludge pump room following the transfer of radioactive sludge into a shipping liner. This resulted in an individual receiving an electronic dosimeter alarm when they were not briefed on current dose rates in the area prior to entering.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000259/
2023-003-00 LER 2023-003-00 for Browns Ferry Nuclear Plant,
Unit 1, Standby Liquid Control Inoperable due to Demineralized Water In-Leakage 71153 Closed LER 05000259/
2023-002-00 LER 2023-002-00 for Browns Ferry Nuclear Plant,
Unit 1, Full Reactor Scram due to an Oscillation Power Range Monitor (OPRM) Confirmation Density Algorithm (CDA) Trip 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at rated thermal power (RTP). On January 24, 2024, operators shutdown the unit for forced outage F111 to repair a failing reactor recirculation pump seal. After replacing the seal, operators restarted the unit on January 30, 2024. On February 3, 2024, the unit was returned to RTP. On February 3, 2024, operators lowered reactor power to 65 percent RTP for a control rod sequence exchange. On February 4, 2024, the unit was returned to RTP. On February 5, 2024, operators lowered reactor power to 68 percent RTP for a control rod sequence exchange. On February 6, 2024, the unit was returned to RTP. On March 8, 2024, operators lowered reactor power to 68 percent RTP for a control rod sequence exchange. On March 10, 2024, the unit was returned to RTP. The unit operated at or near RTP for the remainder of the inspection period.
Unit 2 began the inspection period at RTP. On February 10, 2024, operators lowered reactor power to 65 percent RTP for a control rod sequence exchange. On February 11, 2024, the unit was returned to RTP. On March 29, 2024, operators lowered reactor power to 65 percent RTP for a control rod sequence exchange. On March 30, 2024, the unit was returned to RTP. The unit operated at or near RTP for the remainder of the inspection period.
Unit 3 began the inspection period at RTP. On February 16, 2024, operators shutdown the unit for refueling outage R321. On March 22, 2024, operators restarted the unit following the completion of outage work. On March 29, 2024, the unit was returned to RTP. On March 30, 2024, operators lowered reactor power to 60 percent RTP for a control rod sequence exchange. On March 31, 2024, the unit was returned to RTP.
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.
REACTOR SAFETY
71111.01 - Adverse Weather Protection
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) Seasonal extreme cold weather impact for the intake structure, firefighting, and emergency preparedness capabilities on January 17, 2024
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) Licensee preparations for expected local intense precipitation on January 23, 2024
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 core spray loop I walkdown after core spray flow rate test on January 23, 2024
- (2) Unit common auxiliary decay heat removal system walkdown during a refueling outage with residual heat removal system in a planned maintenance outage on February 26, 2024
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
- (1) Unit 3 core spray on March 29, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 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 1, Fire Area 1-1, northwest quad, reactor building 519, 541, and 565-foot elevations on January 23, 2024
- (2) Unit 1, Fire Area 26, reactor building main steam valve vault, 565-foot elevation on January 26, 2024
- (3) Unit 2, Fire Area 2-5, reactor building 621-foot elevation on February 01, 2024
- (4) Unit 1, Fire Area 1-4, 1B shutdown board room on February 05, 2024
- (5) Unit 3, Fire Area 16, auxiliary room and 1C control bay hallway 593-foot elevation on February 22, 2024
- (6) Unit 3, primary containment on February 28, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (2 Samples)
The inspectors evaluated onsite fire brigade training and performance by observing portions of the following fire drills; or follow-up to an actual event.
- (1) Fire brigade drill in response to a simulated fire in the 1B shutdown board room on January 23, 2024
- (2) Fire brigade response to smoke at the 1A reactor recirculation pump motor heater breaker on January 25, 2024
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated flooding mitigation protections for the underground cables associated with the residual heat removal service water system in manhole covers 15 and 26 on January 22, 2024
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding
Activities (IP Section 03.01)
The inspectors evaluated boiling water reactor non-destructive testing by reviewing the following examinations from February 19 - February 22, 2024:
(1)03.01.a - Nondestructive Examination and Welding Activities.
1. Penetrant Testing (PT)
a.
RWR-3-022-024-C0, pipe to valve, ASME Class 2 b.
RWR-3-022-024-C0, pipe to tee, ASME Class 2
2. Ultrasonic Testing (UT)
a.
KMS-3-089, pipe to pipe, ASME Class 2 b.
KMS-3-090, pipe to tee, ASME Class 2 c.
KMS-3-091, pipe to tee, ASME Class 2
3. Visual Testing (VT)
a.
3-47B455-629, snubber, ASME Class 2
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) Operator performance in the Unit 1 main control room during main generator synchronization on January 31, 2024
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) Shutdown just-in-time training for Unit 3 refueling outage on February 14, 2024
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (7 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) Maintenance risk during control bay chiller crosstie modification work on January 08, 2024
- (2) Emergent work control with Unit 1 reactor recirculation pump seal degradation on January 10, 2024
- (3) Emergent work control with Unit 3 residual heat removal shutdown cooling secured on February 28, 2024
- (4) Emergent work control for Unit 3 planned vessel drain-down for reactor vessel reassembly on March 08, 2024
- (5) Emergent work control for Unit 1 500 kilovolt (kV) switchyard distribution cabinets fed from 4kV common board A are de-energized on March 14, 2024
- (6) Maintenance risk during 3ED 4kV shutdown board maintenance on March 19, 2024
- (7) Emergent work control for Unit 2 high pressure coolant injection isolation on March 21, 2024
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) Condition report (CR) 1915212, loose mounting bolts for the Unit 3B outboard main steam isolation valve air accumulator on March 07, 2024
- (2) CR 1916505, foreign material found on fuel assembly FCK684 on March 14, 2024
- (3) CR 1912331, standby diesel generator 3B auto-start on bus undervoltage on March 14, 2024
- (4) CR 1910087, standby diesel generator 3D failure to start on March 18, 2024
- (5) CR 1911934, high pressure coolant injection room temperature switch as-found out of tolerance on March 18, 2024
- (6) CR 1912704, indications on Unit 3 feedwater sparger end-brackets on March 25, 2024
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Installation of cross tie between the Unit 3 control bay chilled water system and the Unit 1 and 2 control bay chilled water system on March 27, 2024
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
- (1) The inspectors evaluated various Unit 1 forced outage F111 activities from January 25, through January 31, 2024
- (2) The inspectors evaluated various Unit 3 refueling outage 3R21 activities from February 16, through March 24, 2024
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) (8 Samples)
- (1) Final core verification for Unit 3 cycle 22 reload on March 14, 2024
- (2) Work order (WO) 123233074, Unit 3 primary system pressure test, 3-SI-3.3.1A on March 15, 2024
- (3) WO 121335467, Unit 3 reactor core isolation cooling turbine overspeed test on March 18, 2024
- (4) WO 122901848, Unit 3 main generator hydrogen seal repairs on March 25, 2024
- (5) WO 124373923, Unit 2 high pressure coolant injection rupture disc replacement on March 25, 2024
- (6) WO 123888118, Unit 3 standby diesel generator 3D SUDR relay replacement on March 29, 2024
- (7) WO 124224394, Unit 1 recirculation pump 1B seal replacement on March 29, 2024
- (8) WO 123233229, Unit 3 high pressure coolant injection main and booster pump set and flow rate test at 150 psig reactor pressure, 3-SR-3.5.1.8 on March 29, 2024
Surveillance Testing (IP Section 03.01) (7 Samples)
- (1) WO 123351094, Unit 1 core spray flow rate loop I test 1-SR-3.5.1.6 (CS I) on January 16, 2024
- (2) WO 123233827, Unit 3 standby diesel generator 3D emergency load acceptance test on February 14, 2024
- (3) WO 121339305, non-destructive inspection of the critical structural welds on the reactor building crane, BFN-0-CRN-111-0010 on February 14, 2024
- (4) WO 123144738, Unit 3 suppression chamber water temperature monitoring on March 07, 2024
- (5) WO 121335467, inspection of the Unit 3 reactor core isolation cooling governor on March 08, 2024
- (6) WO 123453256, Unit common residual heat removal service water pump C2 comprehensive pump test on March 08, 2024
- (7) WO 123233124, Unit 3 primary containment isolation system low reactor water level instrumental channel A2 calibration test 3-SR-3.3.6.1.5 (A1/A2) on March 12, 2024
Inservice Testing (IST) (IP Section 03.01) (2 Samples)
- (1) WO 123183058, Unit 2 core spray flow rate loop I on March 05, 2024
- (2) WO 123468905, Unit 1 reactor core isolation cooling flow at normal operating pressure on March 08, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) WO 123240519, primary containment local leak rate test containment inerting system: penetration X-229N on March 04, 2024
Reactor Coolant System Leakage Detection Testing (IP Section 03.01) (1 Sample)
- (1) Unidentified drywell leakage monitoring for Unit 1 due to increase leak rate on January 10, 2024
71114.06 - Drill Evaluation
Additional Drill and/or Training Evolution (1 Sample)
- (1) The inspectors observed and evaluated an emergency preparedness drill on February 12, 2024. Drill events included failure of a condenser circulating pump, loss of bus duct cooling leading to a manual scram, and a reactor coolant system leak in the drywell.
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
- (1) Licensee surveys of potentially contaminated material leaving the radiologically controlled area (RCA) during Unit 3 refueling outage.
- (2) Workers exiting the RCA during Unit 3 refueling outage.
- (3) Workers exiting the drywell during Unit 3 refueling outage.
Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Radiation Work Permit (RWP) no. 24300142, 3R21 under vessel activities
- (2) RWP no. 24300072, 3R21 reactor building general activities
- (3) RWP no. 24300092, 3R21 outage refueling floor maintenance activities High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (4 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
- (1) Unit 3 drywell 550-foot through 628-foot elevations
- (2) Radwaste building condensate sludge pump room
- (3) Unit 3 turbine building
- (4) Unit 3 reactor building and refueling floor Radiation Worker Performance and Radiation Protection Technician Proficiency (IP Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &
Transportation
Radioactive Material Storage (IP Section 03.01)
The inspectors evaluated the licensees performance in controlling, labeling, and securing the following radioactive materials:
- (1) Radioactive material stored at the low-level radwaste facility
- (2) Radioactive material stored in the Unit 3 reactor building
Radioactive Waste System Walkdown (IP Section 03.02) (2 Samples)
The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:
- (1) Resin fill station in waste packaging area
- (2) Solid waste handling from reverse osmosis system
Waste Characterization and Classification (IP Section 03.03) (2 Samples)
The inspectors evaluated the following characterization and classification of radioactive waste: (1)2023 reactor water cleanup resin (2)2023 dry active waste
Shipment Preparation (IP Section 03.04) (1 Sample)
- (1) Shipment 1047-C-0211, Type B, resin
Shipping Records (IP Section 03.05) (4 Samples)
The inspectors evaluated the following non-excepted radioactive material shipments through a record review:
- (1) Shipment 231006, Type B, irradiated hardware
- (2) Shipment 1047-C-0195, low specific activity, resin
- (3) Shipment 230812, Type B, irradiated hardware
- (4) Shipment 1047-C-0186, Type B, resin
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10)===
- (1) Unit 1 (January 1, 2023, through December 31, 2023)
- (2) Unit 2 (January 1, 2023, through December 31, 2023)
- (3) Unit 3 (January 1, 2023, through December 31, 2023)
BI02: RCS Leak Rate Sample (IP Section 02.11) (3 Samples)
- (1) Unit 1 (January 1, 2023, through December 31, 2023)
- (2) Unit 2 (January 1, 2023, through December 31, 2023)
- (3) Unit 3 (January 1, 2023, through December 31, 2023)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) March 18, 2023, through February 22, 2024
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program to identify potential trends in foreign material exclusion (FME) that might be indicative of a more significant safety issue.
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 2023-002-00 for Browns Ferry Nuclear Plant, Unit 1, Full Reactor Scram due to an Oscillation Power Range Monitor (OPRM) Confirmation Density Algorithm (CDA)
Trip (ADAMS Accession No. ML23200A238). The inspectors determined that it was not reasonable for the licensee to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified. No violations were identified as a result of this review. This LER is Closed.
- (2) LER 2023-003-00 for Browns Ferry Nuclear Plant, Unit 1, Standby Liquid Control Inoperable due to Demineralized Water In-Leakage (ADAMS Accession No.
ML24029A241). The inspectors determined that this LER was related to a finding previously dispositioned as NCV 05000259/2023004-01 in Integrated Inspection Report 05000259/2023004, 05000260/2023004 AND 05000296/2023004 (ADAMS Accession No. ML24037A104). A minor violation is documented in the findings section of this report. This LER is Closed.
INSPECTION RESULTS
Very Low Safety Significance Issue Resolution Process: Scope expansion of supports in accordance with ASME Section XI, Subsection IWF 71111.08G This issue is a current licensing basis question and inspection effort is being discontinued in accordance with the Very Low Safety Significance Issue Resolution (VLSSIR) process. No further evaluation is required.
Description:
Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g)(4)(ii) states, in part, that Inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the ASME Code incorporated by reference in paragraph
- (a) of this section. The licensees inservice inspection (ISI) program was prepared in accordance with the 2007 Edition of the ASME Section XI Code, with addenda through 2009, as modified by 10 CFR 50.55a.
ASME Section XI, Subsection IWF Requirements for Class 1, 2, 3, and MC Components Supports of Light-Water Cooled Plants (IWF), Section IWF-3400 states, in part, that Component support conditions which are unacceptable for continued service shall include missing, detached, or loosened support items.
During Browns Ferry Unit 3 outage U3R20, the licensee performed a visual examination (VT-3) of high pressure coolant injection (HPCI) system mechanical snubber support, 3-47B455-629, in accordance with ASME Section XI, Table IWF-2500-1, Category F-A, Item No. F1.20. The VT-3 examination identified a relevant condition of two loose nuts on the wedge style concrete anchor bolts that exceeded the acceptance standards in IWF-3400.
ASME Section XI, IWF-3122.3(a) states, in part, that a component support that does not meet the acceptance standards of IWF 3410 shall be acceptable for service without corrective actions if an evaluation or test demonstrates that the component support is acceptable for service.
The licensee performed an engineering evaluation for the nonconforming condition of two loose nuts on concrete anchor bolts. The licensees evaluation states, in part, that the remaining pipe support anchors would be overstressed in tension by approximately 18% and 21%, exceeding their design criteria, but that the component support would still be able to perform its design function and withstand the imposed loading without inducing additional stress into the piping system given the design having a safety factor of 4. The evaluation concluded that two loose nuts needed to be torqued in accordance with their general construction specification. The inspector identified that the design criteria associated with the tensile service allowable capacity was exceeded as described in the licensees engineering design guide/standard DS-C1.7.1, General Anchorage to Concrete. Further communications with the licensee revealed that by exceeding their service allowable capacity, the factor of safety for the component support was reduced to 3.31.
ASME Section XI, IWF-3122.3(b) states, If a component support or a portion of a component support has been evaluated or tested and determined to be acceptable for service in accordance with IWF-3122.3(a), the Owner may perform corrective measures to restore the component support to its original design condition. The requirements of IWF-2220 are not applicable after corrective measures of IWF-3122.2(a) are performed.
The licensee concluded that their engineering evaluation demonstrated that component support 3-47B455-629 was acceptable for service in accordance with IWF-3122.3 even though it exceeded its design requirement. In addition, the licensee voluntarily restored the support back to its original design condition by torquing the loose nuts. The cascading effect of the engineering evaluations conclusion that the support was acceptable for service and subsequent voluntary restoration of the support to its original design condition would result in not having to perform other provisions of ASME Section XI code, specifically, IWF-2430, Additional Examinations. This provision of the code requires additional inspections of adjacent supports when a support requires corrective measures.
Without further substantial research, the inspectors, in consultation with NRR, are unable to determine if the licensees conclusion that the component support 3-47B455-629 was acceptable for service in accordance with IWF-3122.3 even though it exceeded its design requirement with a reduced factor of safety is within the current licensing basis for the facility.
Licensing Basis: NRC Bulletin 79-02, issued March 8, 1979, requested licensees to verify wedge style concrete expansion anchor bolts maintain a minimum design factor of safety of
4. The licensees response to the bulletin, issued June 19, 1979, states, in part, that a
minimum factor of safety of 4 for wedge bolts was used in their designs. The Browns Ferry Unit 3 UFSAR, Appendix C.3.6.1.1 states, in part, that the tensile allowable loads for all support loading conditions for wedge bolts is limited to one-fourth of the anchor concrete pullout capacity. The licensees design specification DS-C1.7.1, General Anchorage to Concrete, section 7.3.1, Expansion Anchor Allowable Capacity: Service Loading Conditions, states, in part, that the allowable tension and shear capacities are less than or equal to one-fourth of the manufacturers published ultimate capacities.
The September 10, 2008, statements of considerations for amendment of the 10 CFR 50.55a regulation stated, In the GALL report, Sections XI.M1, ASME Section XI Inservice Inspection, Subsections IWB, IWC, and IWD, XI.S1, ASME Section XI, Subsection IWE, XI.S2, ASME Section XI, Subsection IWL, and XI.S3, ASME Section XI, Subsection IWF, describe the evaluation and technical bases for determining the adequacy of Subsections IWB, IWC, IWD, IWE, IWF, and IWL, respectively. The inspectors were uncertain if this reference to the GALL report established a method of compliance with subsection IWF.
Without further substantial research, the inspectors could not conclude if the language in the September 10, 2008, statements of considerations for the amendment of the 10 CFR 50.55a regulation conflict with the Browns Ferry BL 79-02 commitments.
Significance: In response to the identified concern, the licensee performed inspections of the upstream and downstream supports and found that those supports were acceptable for service. Based on the inspection results, the inspectors assessed the significance of this concern to establish that, in this instance, the concern if developed into a finding involves very low safety significance. Using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems, Section A, Mitigating SSCs and PRA Functionality (except Reactivity Control Systems)," and determined the violation to be Green because the design or qualification of the mitigating SSC (HPCI Piping System) does maintain its operability.
Technical Assistance Request: As the current licensing basis for this facility cannot be determined without substantial further research and the significance for the associated issue of concern is not greater than very low level the inspection effort is being discontinued in accordance with the VLSSIR process and a technical assistance request (TAR) has not been initiated.
Corrective Action Reference: CR 1913906 Failure to Promptly Repair a Failed Relay in the Standby Diesel Generator Start Circuit Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000296/2024001-01 Open/Closed
[P.2] -
Evaluation 71111.24 A self-revealed Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified when the 3D standby diesel generator (SDG) failed to start during testing due to a failed relay. Specifically, the relay was identified as degraded in January 2023, however, the licensee failed to properly evaluate the significance of the failure and consequently failed to take prompt corrective actions.
Description:
While performing a routine 6-month fast-start surveillance of the Unit 3 3D SDG on December 14, 2022, the licensee identified that the 3D SDG took longer than normal to reach rated conditions. At 9.5 seconds, the 3D SDG still met the respective acceptance criteria of less than 10 seconds; however, the licensee documented this slower-than-normal response in condition report (CR) 1823305. Troubleshooting performed in January 2023, identified that relay BFN-3-RLY-082-D/SUDR was not functioning properly. In researching the function of this relay, the licensee incorrectly concluded that this relay only had a seal-in function in the start sequence of the SDG during a fast-start surveillance. However, the relay had a safety function to actuate and seal-in to allow the SDG to start during a bus undervoltage. As such, failure of the relay would prevent the SDG from starting during all design basis conditions, including a loss of offsite power.
The licensee created work order 123888118 to repair/replace this relay during the next 3D SDG maintenance outage in May 2023. This activity was later re-scheduled to 2025 when the licensee failed to identify any deficiencies in the field during this maintenance outage combined with the continued lack of understanding of the safety function.
On February 14, 2024, during the performance of surveillance procedure 3-SR-3.8.1.9(3D OL) Diesel Generator 3D Emergency Load Acceptance Test with Unit 3 Operating, the 3D SDG failed to start on demand. Troubleshooting identified BFN-3-RLY-082-D/SUDR had misaligned fingers and could not actuate as required. Further troubleshooting identified that if this relay did not actuate, then the 3D SDG could not start on an undervoltage condition. This was the direct cause of the 3D SDG failure to start during the load acceptance test. The licensee determined that the failure of this relay would have prevented the 3D SDG from automatically starting on an actual loss of offsite power and that the previous conclusion that the relay only had to function during the fast start test was incorrect. The licensee repaired the relay, re-performed the 3D SDG load acceptance test, and returned the SDG to service on February 15, 2024.
The licensee performed a past operability evaluation and found that the 3D SDG was inoperable from December 14, 2022, until February 15, 2024. The inspectors concluded that the failure to promptly identify and correct the condition adverse to quality associated with the failed relay directly led to the failure to restore the 3D SDG to operable status within the allowed outage time. Therefore, the corrective action aspect of this issue represented a more fundamental deficiency than the failure to comply with the Technical Specifications.
The licensee informed the inspectors that if they had properly evaluated the safety function of the relay in CR 1823305 then they would have performed more intrusive relay testing which would have identified that the relay was degraded to the point where it would not have performed its safety function. Consequently, the licensee likely would have either repaired or replaced the relay prior to the load acceptance test failure on February 14, 2024.
Corrective Actions: Work order 124301137 aligned BFN-3-RLY-082-D/SUDR contacts to restore functionality and surveillance procedure 3-SR-3.8.1.9(3D OL) was satisfactorily reperformed to restore operability. No common cause was identified with the other seven SDGs. Work order 123888118 later replaced BFN-3-RLY-082-D/SUDR with a new relay during the Unit 3 spring 2024 refueling outage.
Corrective Action References: CRs 1823305 and 1910087
Performance Assessment:
Performance Deficiency: The failure to identify and correct a condition adverse to quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI was a performance deficiency reasonably within the licensees ability to foresee and prevent. Specifically, following troubleshooting conducted in January 2023 (CR 1823305), the licensee had sufficient information to determine that relay BFN-3-RLY-082-D/SUDR needed to be either repaired or replaced. However, the licensee did not do so until failure in February 2024 despite a maintenance outage during the first available window on the affected SDG in May 2023.
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.
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 screened the performance deficiency using Exhibit 2 of Appendix A and determined that the issue needed a detailed risk evaluation due to the degraded condition representing a loss of the probabilistic risk assessment (PRA) function of one train of a multi-train technical specification for greater than its technical specification allowed outage time with 3D SDG being inoperable from December 14, 2022, until February 15, 2024.
A regional senior reactor analyst (SRA) conducted a detailed risk assessment for the degraded condition. The SRA assumed a maximum exposure time of 1 year to conservatively bound the condition. The SRA modeled the condition using Saphire 8 version 8.2.9 and the Browns Ferry Unit 3 Standardized Plant Analysis Risk (SPAR) model version 8.82 dated August 14, 2023. The SRA set EPS-DGN-FS-DG3D, Diesel Generator 3D Fails to Start to True which ensures the common cause failure term for the station emergency standby diesel generators is adjusted appropriately. The dominant accident sequence is a Weather-Related Loss of Offsite Power event with operators failing to establish suppression pool cooling when required, failing to vent containment when required, and a failure of late injection with the residual heat removal system. In this bounding analysis, risk was less than 1 E-7 events per year; therefore, external events were not required to be analyzed and risk was characterized as very low safety significance (Green).
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. Specifically, the licensee identified and documented a deficiency with relay BFN-3-RLY-082-D/SUDR but failed to properly evaluate the significance of a failure of the relay. Having failed to evaluate the safety significance of the failure, TVA did not restore the relay to proper condition at the first available maintenance opportunity.
Enforcement:
Violation: As required by 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, measures shall be established to assure that conditions adverse to quality, such as malfunctions, are promptly identified and corrected. TVA Nuclear Quality Assurance Plan TVA-NQA-PLN-89-A, dated April 16, 2021, section 10.2.2 states, in part, that TVA Nuclear organizations performing quality-related activities at nuclear facilities shall promptly identify and resolve adverse conditions.
Contrary to the above, from May 13, 2023, until February 15, 2024, the licensee failed to promptly identify and correct a condition adverse to quality. Specifically, from the end of the 3D SDG maintenance outage until the relay was ultimately repaired, the licensee failed to promptly identify and correct the malfunction with BFN-3-RLY-082-D/SUDR, a safety-related component.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Evaluate Radiological Conditions Following the Transfer of Radioactive Sludge to a Shipping Liner Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green NCV 05000259,05000260,05000296/20240 01-02 Open/Closed
[H.4] -
Teamwork 71124.01 A self-revealed Green finding and associated NCV of 10 CFR 20.1501, Surveys and Monitoring, was identified when the licensee failed to perform adequate surveys following a significant change in radiological conditions. Specifically, the licensee failed to evaluate radiological conditions in the 546-foot elevation condensate sludge pump room following the transfer of radioactive sludge into a shipping liner. This resulted in an individual receiving an electronic dosimeter alarm when they were not briefed on current dose rates in the area prior to entering.
Description:
On May 31, 2023, an operator was performing activities in the 546-foot elevation condensate sludge pump room, a posted HRA located in the radwaste building. This worker was on a RWP that had a self-reading dosimeter (SRD) alarm dose rate setpoint of 150 mrem/hr. The operator was briefed on radiological conditions based on a routine survey performed on May 2, 2023, which indicated dose rates of 22 mrem/hr at 30cm in the room. However, radwaste Operations had recently performed a transfer of radioactive sludge into a shipping liner, which required the condensate sludge pumps to be turned on. As a result, dose rates in the room had increased substantially compared to the previous readings taken on May 2. Unaware of the increased dose rates, the worker entered the room and received an SRD dose rate alarm, at which point they exited the area and reported to RP. A follow-up survey indicated dose rates had risen to 800 mrem/hr at 30cm.
The licensee indicated that communications between radwaste Operations and the radiation protection (RP) department regarding movement of the sludge were ineffective in this case. Although operating instruction 0-OI-77E, Solid Waste, Revision 50, contains several warnings and prompts to communicate with RP when evolutions are performed that could change radiological conditions, it is unclear whether those steps were performed. Neither radwaste Operations nor RP maintained records or logs of communications regarding potentially changing conditions in the radwaste building.
Corrective Actions: The licensees immediate corrective actions included flushing associated piping in the 546-foot condensate sludge pump room and conservatively posting the room as a locked high radiation area.
Corrective Action References: The licensee placed this event into their corrective action program under CRs 1859935 and 1912194.
Performance Assessment:
Performance Deficiency: The failure to evaluate radiological conditions in the 546-foot condensate sludge pump room following the transfer of sludge into a liner in accordance with 10 CFR 20.1501 was a performance deficiency and was reasonably within the licensees ability to foresee and correct.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Program & Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, workers who are not made aware of elevated radiological conditions in a HRA could receive unintended exposure.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety SDP. The finding was determined to be of very low safety significance (Green) because it was not related to as low as reasonably achievable planning, did not result in an overexposure beyond regulatory limits, there was no substantial potential for overexposure, and the ability to assess dose was not compromised.
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. This cross-cutting aspect was assigned to this performance deficiency due to inadequate communication between radwaste Operations and RP when sludge transfer activities occurred.
Enforcement:
Violation: As required by 10 CFR 20.1501, each licensee shall make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Contrary to this, on May 31, 2023, the licensee failed to evaluate the radiological hazard present in the 546-foot elevation condensate sludge pump room following the transfer of radioactive sludge into a shipping liner.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Observation: Semi-Annual Trend in Foreign Material Exclusion Practices 71152S The inspectors noted a number of issues related to FME controls and actual foreign material intrusion which did not meet licensee expectations. The inspectors observed that these instances all represented challenges to the licensee's defense-in-depth FME program that, if taken collectively, could indicate a more fundamental challenge to the licensee meeting their overall standards in this area.
The inspectors discussed these concerns with the licensee who acknowledged them.
Examples of these concerns included the identification of clear plastic on the refueling floor during fuel movements, the lack of clear standards with respect to tool and equipment tie-offs in a Zone 1 area, and foreign material identified on top of two fuel assemblies during a review of the final core verification video. Additional observations on FME practices will continue to be a focus item, as the inspectors deem appropriate, for future inspections.
Minor Violation: Failure to Follow Operability Determination Process 71153 Minor Violation: As required by 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, activities affecting quality shall be accomplished in accordance with procedures. Contrary to the above, the licensee failed to accomplish an activity affecting quality in accordance with procedures. Specifically, on January 19, 2023, operators identified an inadvertent level rise in the Unit 1 standby liquid control tank, a condition that could affect Unit 1 compliance with the applicable Technical Specification requirement. Procedure NPP-SPP-22.300, Corrective Action Program, requires, in part, that the license review the condition to determine if the condition affects operability. The licensee did not review the level rise to determine if it affected operability until the inspectors questioned the operability during a review of a similar event approximately ten months later as documented in Inspection Report 05000259/2023004, 05000260/2023004 AND 05000296/2023004 (ADAMS Accession Number ML24037A104). As a result of the inspectors' questions, the Unit 1 standby liquid control system was determined to be inoperable for greater than its allowed outage time and was subsequently reported as LER 2023-003-00 on January 29, 2024. The inspectors concluded that the failure to identify the inoperability on or around January 19, 2023, and thus report this event on or before March 20, 2023, was due to the failure to follow the immediate operability process in NPP-SPP-22.300.
Screening: The inspectors determined the performance deficiency was minor. The inspectors concluded that this violation was of minor significance because it did not contribute to the actual inoperability of the system and thus did not impact any cornerstone objective.
Enforcement:
This failure to comply with 10 CFR 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 4, 2024, the inspectors presented the integrated inspection results to Manu Sivaraman, site vice president (SVP), and other members of the licensee staff.
- On February 23, 2024, the inspectors presented the radiation protection inspection results to Daniel Komm, plant manager, and other members of the licensee staff.
- On February 29, 2024, the inspectors presented the ISI inspection results to Manu Sivaraman, SVP, and other members of the licensee staff.
- On April 25, 2024, the inspectors presented (re-exit) the ISI inspection results to Brad Bruce, plant support director, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
Operating
Instruction 0-OI-
2/ATT-1
Auxiliary Decay Heat Removal System Valve Lineup
Checklist
Rev. 55
Work Orders
Work Orders
Calculations
CD-Q3073-
910866
High-Pressure Coolant Injection System Qualification of Pipe
Support No. 3-47B455-629
Rev. 2
Corrective Action
Documents
CR 1759027
Condition Report
03/07/2022
Engineering
Evaluations
NOI U3R20-001
Notice of Indication Form
03/21/2022
Procedures
DS-C1.7.1
General Anchorage to Concrete
Rev. 13
Corrective Action
Documents
Condition Report
(CR)
1915212, 1910087, 1911934, 1911953, 1911955, 1911956,
1916505, 1916811, 1912331, 1912475, 1912704, 1907937
Drawings
0-47B435-8D,
3-47B600-1431,
3-47B600-1432
Mechanical General Notes, Pipe Supports,
Mechanical Control Air System I&C Pipe Support,
Mechanical Control Air System I&C Pipe Support
05/08/1992,
Rev. 0 and
Rev. 1
Work Orders
WO 24316626, 124364967
Work Orders
Corrective Action
Documents
CR 1922156
Condition Report
Unit 1 F111 startup sequence control rod movement data
sheet
Miscellaneous
Unit 1 Planned F111 outage safety plan
01/24/2024
Corrective Action
Documents
Condition Report
(CR)
1913028, 1911505, 1916505, 1922153
Drawings
2-47E812-1-ISI
ASME Section XI HPCI Code Class Boundaries
Dated
10/3/23
Calibration
Certificate No.
267645
Certificate of Calibration for Torque Wrench E51498
Dated
9/12/23
Miscellaneous
Calibration
Certificate of Calibration for Torque Wrench E45140
Dated
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Certificate No.
268824
10/3/23
Procedures
Procedure 3-SR-
3.8.1.9(3D OL)
Diesel Generator 3D Emergency Load Acceptance Test with
Unit 3 Operating
Rev. 22
21339305, 123240519, 123183058, 121335467,
23468905, 123453256, 123233124, 124308777,
21335467, 123233074, 124373923, 123233827,
24224394, 123233229, 123351094
Work Orders
Corrective Action
Documents
CR 1859935
Condition Report
Corrective Action
Documents
Resulting from
Inspection
CR 1912194
Condition Report
Procedures
0-OI-77E
Solid Radwaste
Rev. 50
M-20230502-12
Radwaste 546' General Area
05/02/2023
Radiation
Surveys
M-20230531-13
RW 546' Condensate Sludge Pump Room
05/31/2023
Radiation Work
Permits (RWPs)
23040552
Unit 0 Radwaste Operations Activities
Rev. 1
CR 1912203
Condition Report
CR 1912204
Condition Report
Corrective Action
Documents
Resulting from
Inspection
CR 1912205
Condition Report
Engineering
Evaluations
WMG Report 23-
395-RE-298
Packaging and Disposal of Irradiated Hardware at Browns
Ferry Nuclear Plant During 2023
November
23
CS-FP-PR-010
Preparation of Loaded 3-60B Shipping Cask for
Transportation
Rev. 4
Procedures
NPP-SPG-05.6.1
Radiation Protection Implementation of 10 CFR 37 Category
and Category 2 Quantities of Radioactive Material
Rev. 3
Corrective Action
Documents
CR 1922159
Condition Report