IR 05000321/2024001
| ML24107A927 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 04/22/2024 |
| From: | Alan Blamey Division Reactor Projects II |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| IR 2024001 | |
| Download: ML24107A927 (24) | |
Text
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANTS, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000321/2024001 AND 05000366/2024001
Dear Jamie Coleman:
On March 31, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Edwin I. Hatch Nuclear Plants, Units 1 and 2. On April 17, 2024, the NRC inspectors discussed the results of this inspection with Mr. Jason Butler, plant manager, 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. None of these findings involved a violation of NRC requirements.
If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement 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 Edwin I. Hatch Nuclear Plants, Units 1 and 2.
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, Alan J. Blamey, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos. 05000321 and 05000366 License Nos. DPR-57 and NPF-5
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000321 and 05000366 License Numbers:
DPR-57 and NPF-5 Report Numbers:
05000321/2024001 and 05000366/2024001 Enterprise Identifier:
I-2024-001-0024 Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Edwin I. Hatch Nuclear Plants, Units 1 and 2 Location:
Baxley, GA Inspection Dates:
January 1, 2024 to March 31, 2024 Inspectors:
J. Bell, Senior Health Physicist B. Bowker, Reactor Inspector R. Easter, Resident Inspector B. Kellner, Senior Health Physicist D. Neal, Health Physicist P. Niebaum, Senior Resident Inspector A. Rosebrook, Senior Reactor Analyst Approved By:
Alan J. Blamey, Chief Reactor Projects Branch 2 Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Edwin I. Hatch Nuclear Plants, Units 1 and 2, 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 Operate the Hatch Unit 2 Radwaste Area HVAC System as described in the FSAR Design Bases Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green FIN 05000366/2024001-01 Open/Closed
[P.4] - Trending 71124.08 The inspectors identified a Green finding (FIN) for failure to meet the Hatch Unit 2 Final Safety Analysis Report (FSAR) radwaste area heating, ventilation, and air conditioning (HVAC)system design bases as described in FSAR section 9.4.3.1. Specifically, the unit 2 radwaste ventilation was out of service (OOS) from April 2020 through June 2023, operated for a short period, and was not operating at the time of the inspection due to flow indication related issues that require troubleshooting and repair. Additionally, no reviews were completed to evaluate the significance of the OOS ventilation system, nor were any compensatory air sampling activities performed in the radwaste building during the extended OOS period.
Failure to Use Design Control Process Results in Unit 2 Loss of Feedwater and Manual Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000366/2024001-02 Open/Closed
[P.2] -
Evaluation 71152A A self-revealing FIN was identified when the licensee installed a pipe support for a 1-inch feedwater drain pipe outside of the design change process in 2007. Specifically, the pipe support was installed too close to the branch connection with a U-bolt configured as a 3-way restraint per maintenance work order (MWO) 2061756801, contrary to licensee procedure 40AC-ENG-003-0, Design Control.
Additional Tracking Items
None.
PLANT STATUS
Unit 1 began the inspection period at approximately 79 percent rated thermal power (RTP) in response to an electro-hydraulic control (EHC) fluid leak associated with the tubing for the number (no.) 2 turbine control valve (TCV). On January 2, 2024, power was raised to approximately 94 percent RTP. On January 3, the licensee established 93 percent RTP as the new power limit due to moisture carry over concerns during end-of-cycle operation. On January 3, power was lowered to approximately 82 percent RTP due to EHC pipe vibrations. On January 11, power was raised to approximately 86 percent RTP. On February 3, power was lowered in preparation to shut down for the upcoming unit 1 cycle 31 (1R31) refueling outage that began on February 4. Following the refueling outage, the unit was re-started on March 10 and began raising power. On March 11, the unit was manually scrammed from approximately 35 percent RTP due to the trip of the A reactor feed pump/turbine and was later cooled down to Mode 4 (cold shutdown). On March 17, the unit was re-started and began raising power. On March 22, the unit achieved approximately 97 percent RTP with extraction steam isolated to the fifth stage feedwater heaters. On March 25, power was lowered to approximately 70 percent RTP to isolate an EHC fluid leak associated with the no. 3 TCV. On March 26, power was lowered to 57 percent RTP to perform a modification to the instrument taps on the fifth stage feedwater heaters. On March 28, power was lowered further to approximately 30 percent RTP to better isolate the feedwater heater in support of the modification work. Following the completion of the work, unit 1 began raising power on March 29 and reached near 100 percent RTP on March 31, 2024 Unit 2 operated at or near rated thermal power for the entire 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 onsite 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)
The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of seasonal cold temperatures for the following systems:
- (1) Emergency diesel generators (EDG), and plant service water (PSW) pumps in the intake structure on January 16 and 17, 2024
Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the adequacy of the overall preparations to protect risk significant systems from impending severe weather (i.e., Tornado Watch/Warning) in the local area on January 9, 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 A standby gas treatment (SBGT) system on January 25, 2024
- (2) Unit 1 A residual heat removal (RHR) system on February 28-29, 2024
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Units 1 and 2 cable spreading room on January 3, 2024
- (2) Unit 1 steam chase on February 8, 2024
- (3) Unit 1 torus room on February 16, 2024
- (4) Unit 1 and U2 turbine building 164-foot elevation, east wall on February 26, 2024
- (5) Unit 1 drywell on February 8, 2024
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated the onsite fire brigade training and performance during an unannounced fire drill on March 21, 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 nondestructive testing by reviewing the following examinations from February 5 - 8, 2024:
- (1) Ultrasonic Examination
- 1B11\\STUD-1 THRU 52, closure head studs, ASME Class 1
- 1B21-1FW-12AB-1, feedwater (FW) reducer to pipe, ASME Class 1
- 1B21-1FW-12AB-2, FW pipe to elbow, ASME Class 1
- 1B21-1FW-12AB-3, FW elbow to elbow, ASME Class 1 Liquid Penetrant Examination (PT)
- Pre-and post-weld PTs for work order (WO) SNC781702, weld buildup associated with replacing RHR service water (SW) discharge check valve (1E11C001C), ASME Class 3
- Three post-weld PTs for WO SNC1368285, RHRSW strainer piping replacement, ASME Class 3 Visual Examination
- 1E11-S1, VT-3, RHR hydraulic snubber, ASME Class 1
- 1B21-SS1, VT-3, main steam (MS)/FW hydraulic snubber, ASME Class 1
- 1B21-RCIC-SS-41, VT-3, MS/FW hydraulic snubber, ASME Class 1
- 1B21-FDH-700, VT-3, MS/FW hydraulic snubber, ASME Class 1
- 1B11-STUD-1 THRU 52, closure head studs, ASME Class 1
- 1B11 I\\K-SD-1, VT-3 steam dryer upper vane banks 0 to 90 degrees (º), ASME Class 1
- 1B11 I\\K-SD-2, VT-3 steam dryer upper vane banks 90 to 180º, ASME Class 1
- 1B11 I\\K-SD-3, VT-3 steam dryer upper vane banks 180 to 270º, ASME Class 1
- 1B11 I\\K-SD-4, VT-3 steam dryer upper vane banks 270 to 360º, ASME Class 1 Welding Activities
- Gas tungsten arc welding o
WO SNC781702, weld buildup associated with replacing RHRSW discharge check valve (1E11C001C), ASME Class 3 o
WO SNC1368285, RHRSW strainer piping replacement, ASME Class 3
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01)
(1Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the main control room during the unit 1 final feedwater temperature reduction on January 10, 2024, during the unit 1 load reduction for refueling outage 1R31 on February 4, 2024, and during the unit 1 reactor startup on March 10, 2024.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated 'just-in-time-training (JITT) of licensed operators for the unit 1 startup following refueling outage 1R31 on March 4, 2024.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 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) Unit 1 A PSW maintenance rule (10 CFR 50.65) (a)(1) evaluation
- (2) Condensate/feedwater system due to a plant level event or PLE associated with unit 2 scram and loss of feedwater on November 1, 2023
- (3) Unit 1 RHR heat exchanger bypass valve (1E11-F048B) would not stroke from remote shutdown panel on February 12, 2024
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 1, risk informed completion time (RICT) evaluation for Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1.1.A (Function 9 of Table 3.3.1.1-1)entered on December 30, 2023 (and exited on January 10, 2024) due to an inoperable TCV no. 2 fast closure function.
- (2) Unit 1, elevated risk due to a B SBGT system outage on January 22, 2024.
- (3) Unit 1, elevated risk while unit 1 was in cold shutdown (Mode 4) with low reactor water level on February 26, 2024.
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (5 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Condition report (CR) 11042235. Swing 1B EDG jack water heat exchanger leak, reviewed on January 31, 2024.
- (2) CR 11048856. Unit 1 RHR valve motor pinion backing out, reviewed on March 1, 2024.
- (3) CR 11056947. Unit 1 emergency core cooling systems suction strainers impact due to loose insulation and duct tape, reviewed on March 8, 2024.
- (4) CR 11048534. Unit 1 high pressure coolant injection (HPCI) 10-inch pipe saddle detachment reviewed on March 19, 2024.
- (5) CR 11050267. Unit 1 PSW pipe saddle (1P41-SWH-90) for failed inservice inspection (ISI) VT examination (past operability), reviewed on March 21, 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)42EN-P41-001-1. Temporary routing of PSW from main control room environmental control system chillers to unit 2, completed on February 28, 2024.
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
- (1) The inspectors evaluated unit 1 refueling outage 1R31 activities from February 4 to March 11, 2024.
- (2) The inspectors evaluated unit 1 forced maintenance outage activities following the trip of the A reactor feedwater pump/turbine and manual reactor scram from March 11 to March 18, 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) (6 Samples)
(1)34SV-P41-003-2, standby diesel service water system operability on January 24, 2024 (2)34SV-C71-005-1, TCV fast closure instrument functional test on January 10, 2024 (3)52PM-B31-003-1, reactor recirculation pump motor maintenance on March 5, 2024 (4)34SV-E41-005-1, unit 1 HPCI testing at 165 pounds-per-square-inch gauge (psig) on March 10, 2024 (5)42IT-TET-006-1, ISI pressure test of the class 1 system and recirculation pump runback test following unit 1 reactor vessel reassembly on February, 25 and 26, 2024
- (6) WO SNC1728349, unit 1 local leak rate test (LLRT) of outboard feedwater isolation 18-inch check valve, reviewed on March 19, 2024
Surveillance Testing (IP Section 03.01) (3 Samples)
(1)34SV-R43-012-2, swing 1B EDG 24-month surveillance (24-hour run) from unit 2 on January 8, 2024 (2)34SV-R43-022-1, unit 1 1C EDG loss of coolant accident (LOCA)/loss of offsite power (LOSP) surveillance on February 12, 2024 (3)34SV-E41-002-1, unit 1 HPCI testing at 920 psig on March 11, 2024
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
(1)34SV-E21-001-2, unit 2 A core spray pump IST on January 4, 2024
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
(1)2SV-TET-001-0, unit 1 main steam isolation valve LLRT on February 7 and 8,
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) Workers doffing protective clothing at the unit 1 drywell
- (2) Workers exiting the radiologically controlled area (RCA) at unit 1 during refueling outage 1R31
- (3) Control of the clean island for the generator rewind project inside of the RCA
Radiological Hazards Control and Work Coverage (IP Section 03.04) (4 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Unit 1 torus diving activities
- (2) Unit 1 reactor vessel disassembly
- (3) Unit 1 fuel movement
- (4) Unit 1 reactor water cleanup (RWCU) heat exchanger room locked high radiation area (LHRA) entry for ISI activities High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (5 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
- (1) Unit 1 drywell HRA
- (3) Unit 1 fuel pool cooling heat exchanger LHRA
- (5) Unit 1 traversing in-core probe LHRA 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) Low level radioactive waste facility (LLRWF) and waste separation and temporary storage facility (WSTSF) (located within the same fenced area)
- (2) Interim radioactive waste storage facility (IRWSF) (Category 2 storage)
- (3) Various designated radioactive storage locations inside the protected area boundary
Radioactive Waste System Walkdown (IP Section 03.02) (3 Samples)
The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:
- (1) Unit 1 radwaste building (primarily liquid and gaseous radioactive waste processing)
- (2) Unit 2 radwaste building (primarily liquid and gaseous radioactive waste processing)
- (3) Unit 2 waste tanks (e.g. chemical waste sample tank and waste surge tank) and sample hoods
Waste Characterization and Classification (IP Section 03.03) (3 Samples)
The inspectors evaluated the following characterization and classification of radioactive waste:
- (1) Hatch 2022 scaling factor determination for dry active waste (DAW), condensate polishing resin (CPS) for units 1 and 2, and spent resin (SR) for units 1 and 2, March 3, 2024
- (2) Hatch 2023 scaling factor determination for DAW, CPS for units 1 and 2, and SR for units 1 and 2, May 30, 2023
- (3) Hatch 2023 scaling factor determination for unit 2 torus filters, May 30, 2023
Shipment Preparation (IP Section 03.04) (1 Sample)
- (1) The inspectors observed the preparation of radioactive shipment no. 24-RW-006, Radioactive Material, Scrap Metal, Surface Contaminated Objects (SCO-II), on February 7, 2024.
Shipping Records (IP Section 03.05) (4 Samples)
The inspectors evaluated the following non-excepted radioactive material shipments through a record review:
- (1) Shipment no. 22-RW-043, Radioactive Waste, Reactor Water Cleanup Pump (RWCU),
Low Specific Activity (LSA-II), November 30, 2022
- (2) Shipment no. 22-RM-011, Radioactive Material, Unit 1 Control Rod Drives (CRD), LSA-II, March 1, 2022
- (3) Shipment no. 23-RW-024, Radioactive DAW, Asbestos Waste, LSA-II, June 30, 2023
- (4) Shipment no. 23-RW-042, Radioactive Resin (gross dewatered for processing), LSA-II, December 12,
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 (January 1, 2023 through December 31, 2023)
- (2) Unit 2 (January 1, 2023 through December 31, 2023)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
- (1) Unit 1 (January 1, 2023 through December 31, 2023)
- (2) Unit 2 (January 1, 2023 through December 31, 2023)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
- (1) Unit 1 (January 1, 2023 through December 31, 2023)
- (2) Unit 2 (January 1, 2023 through December 31, 2023)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) February 1, 2023 to January 31, 2024 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) July 1, 2023 to January 31, 2024
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) Corrective action report (CAR) 547712. Root cause report for the unit 2 FW drain pipe break, loss of both reactor feed pumps, and manual scram on November 1, 2023.
- (2) CAR 560851. Equipment reliability checklist (ERC) for the unit 2 A EDG fire door that was discovered non-functional on December 19, 2023.
71153 - Follow-up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) The inspectors evaluated a unit 1 manual scram on lowering reactor water level and licensees response on March 11, 2024.
Event Report (IP Section 03.02) (1 Sample)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000366/2023-003-0, Unit 2 Manual Reactor Trip due to Loss of Feedwater (ADAMS Ascension No. ML ML23353A109). The inspection conclusions associated with this LER are documented in this report under Inspection Results Section 71152A. This LER is closed.
INSPECTION RESULTS
Failure to Operate the Hatch Unit 2 Radwaste Area HVAC System as described in the FSAR Design Bases Cornerstone Significance Cross-Cutting Aspect Report Section Occupational Radiation Safety Green FIN 05000366/2024001-01 Open/Closed
[P.4] - Trending 71124.08 The inspectors identified a Green finding for failure to meet the Hatch Unit 2 Final Safety Analysis Report (FSAR) radwaste area heating, ventilation, and air conditioning (HVAC)system design bases as described in FSAR section 9.4.3.1. Specifically, the unit 2 radwaste ventilation was out of service (OOS) from April 2020 through June 2023, operated for a short period, and was not operating at the time of the inspection due to flow indication related issues that require troubleshooting and repair. Additionally, no reviews were completed to evaluate the significance of the OOS ventilation system, nor were any compensatory air sampling activities performed in the radwaste building during the extended OOS period.
Description:
During a walkdown of the unit 2 radwaste building, the inspectors noted that building ventilation was not running. Upon questioning the licensee, it was revealed that the radwaste building ventilation system had been mostly OOS since April 2020 due to several mechanical issues. The inspectors noted FSAR section 9.4.3.1 states, in part, that the design basis for the radwaste area HVAC is to minimize the escape of potential airborne radioactivity to the outside atmosphere and to exhaust air, through a suitable filtration system, from the areas where a significant potential for radioactive particulates and radioactive iodine contamination exists and to provide for air movement from areas of lesser potential airborne radioactivity to areas of greater potential airborne radioactivity prior to final exhaust. The inspectors also noted that FSAR sections 11.3.2.2, 12.1, 12.3 states the need for airborne radioactivity air sampling in plant areas susceptible to airborne radioactivity, when conditions warrant, to protect operating personnel and to reduce onsite and offsite radiation levels from potentially radioactive air by exhausting it through a filter train consisting of roughing, charcoal, and high efficiency particulate air (HEPA) filters. However, discussions with licensee staff and a review of maintenance records and corrective action documents showed that the unit 2 radwaste ventilation was OOS from April 2020 through June 2023. Also, evaluations were not completed to determine the impact of the ventilation being inoperable for extended periods of time, nor were there any compensatory actions in place for the OOS ventilation system. Lastly, while several low-priority corrective action documents were created for individual issues with the system, no condition report had been created to address the extended inoperability of the overall system to drive timely return of the system to service.
Corrective Actions: As immediate compensatory actions, the licensee obtained baseline air samples, placed a portable continuous air monitor, with alarm function, on the 178-foot elevation, implemented routine monthly particulate and iodine grab air sampling, and will obtain job coverage samples as required for specific work being performed in the building.
Corrective Action References: Condition report (CR) 11048487
Performance Assessment:
Performance Deficiency: The inspectors determined that the failure to operate the unit 2 radwaste ventilation system to minimize exposure of operating personnel from airborne radioactivity and to minimize the escape of potential airborne radioactivity to the outside atmosphere to reduce onsite and offsite radiation levels from potentially radioactive air for an extended period of time was outside the design basis as described in Hatch Unit 2 FSAR section 9.4.3.1 and was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Plant Facilities/Equipment and Instrumentation 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.
Inadequate monitoring of areas with the potential for airborne radioactivity and not operating ventilation and filtration systems as designed can lead to worker contamination and increased radiation exposure.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process. The finding was determined to be of very low safety significance (Green) because this was not an ALARA planning issue, there was neither an overexposure or a substantial potential for overexposure, the licensees ability to assess worker dose was not compromised, and the dose impact to a member of the public from a radiological release was not likely to exceed the dose values in Appendix I to 10 CFR Part 50 and/or 10 CFR 20.1301(e).
Cross-Cutting Aspect: P.4 - Trending: The organization periodically analyzes information from the corrective action program and other assessments in the aggregate to identify programmatic and common cause issues. No active ownership, oversight, or trending of operation or failures of the radwaste ventilation system was noted by the inspectors.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Failure to Use Design Control Process Results in Unit 2 Loss of Feedwater and Manual Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000366/2024001-02 Open/Closed
[P.2] -
Evaluation 71152A A self-revealing finding (FIN) was identified when the licensee installed a pipe support for a 1-inch feedwater drain pipe outside of the design change process in 2007. Specifically, the pipe support was installed too close to the branch connection with a U-bolt configured as a 3-way restraint per maintenance work order (MWO) 2061756801, contrary to licensee procedure 40AC-ENG-003-0, Design Control.
Description:
On November 1, 2023, on unit 2, a 1-inch drain pipe branched from a 20-inch reactor feedwater suction pipe in the A reactor feed pump (RFP) room sheared from its half coupling connection and resulted in an estimated 400 gallon per minute (gpm) steam/water leak in the turbine building. In response, the licensee operations staff reduced reactor power to approximately 58 percent of rated thermal power and removed the A RFP from service.
Water from the leak entered electrical cabinets and impacted components associated with the B RFP and caused it to trip and no longer provide feedwater to the reactor vessel. With reactor vessel water level lowering, due to no feedwater, the operators inserted a manual scram to shut down the unit. The licensee determined that the failed drain pipe had a history of vibration, had multiple changes to its support configuration, and had multiple failures of supporting U-bolts leading up to the pipe failure on November 1, 2023.
According to the licensees investigation, the drain pipe sheared at the half coupling due to excessive stresses from low and high cycle fatigue. In 2007, an unanalyzed pipe support was installed approximately 3.5 feet from the branch connection using a maintenance work order (MWO 2061756801) and not the design change process as required by licensee procedure 40AC-ENG-003-0, Design Control, Revision 9.5. Specifically, paragraph 8.2.1.5 of the procedure required engineering support notification when design interpretations or alternate design configurations are needed during the conduct of routine operations or maintenance activities. The licensees root cause report determined the minimum design offset length for installation of a pipe support for the 1-inch drain pipe would be approximately 10 feet away from the branch connection. This accounts for the total thermal displacement, from the licensees piping stress calculation BH2-PD-0404, and information contained in drawing A42900, Design Guideline for Smallbore Piping (Non-Seismic). The maintenance work order (MWO 2061756801) that installed the pipe support in 2007 was not associated with any engineering notification or evaluation supporting the application of an alternate design configuration. Additionally, MWO 2061756801configured the U-bolt support as a 3-way restraint. This was contrary to industry guidance that recommended the use of a 3-way restraint on systems operating at temperatures less than 150 degrees Fahrenheit (°F) and on piping that does not exhibit low cycle fatigue from thermal cycles. The normal operating temperature and pressure of the reactor feedwater suction pipe and drain pipe is approximately 325 °F and 325 pounds-per-square-inch gauge (psig).
It appears that the U-bolt broke off and/or was removed sometime after its installation in 2007 and before it was reinstalled on June 30, 2023, per work order SNC1504851. Additionally, U-bolts were found broken at the same support stanchion several times following the installation on June 30, 2023. The U-bolts were discovered broken on September 18, 2023, October 10, 2023, and October 29, 2023, as documented in condition reports, and were replaced each time under a maintenance work order. With a U-bolt installed in this location and in the configuration that restrained the 1-inch drain pipe, excessive stresses were introduced to the weld at the drain pipe branch connection causing its eventual failure.
Corrective Actions: The licensee initiated a design change (SNC161440) that cut off and capped the 1-inch drain pipe at the sheared branch connection.
Corrective Action References: Root cause report (CAR 547712)
Performance Assessment:
Performance Deficiency: The failure to evaluate the adequacy of the 1-inch feedwater drain pipe support installed under MWO 2061756801 in accordance with licensee procedure 40AC-ENG-003-0, Design Control, was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency adversely affected the cornerstone objective in that it contributed to a feedwater drain line failure that resulted in a loss of reactor feedwater (i.e., heat sink) and a manual scram of unit 2 on November 1, 2023.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power. The inspectors screened this finding using Exhibit 1, Initiating Events Screening Questions of NRCs Inspection Manual Chapter (IMC) 0609, Appendix A. The finding could not be screen to Green and required a detailed risk evaluation because the finding resulted in a reactor scram and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of feedwater).
A regional senior reactor analyst (SRA) performed a detailed risk evaluation using the guidance in IMC 0609 Appendix A, The Significance Determination Process for Findings At-Power and the Risk Assessment of Operational Events Handbook, Volume 1 - Internal Events. The failure mechanism was high cycle fatigue failure due to stresses placed on the main feedwater pump drain line welds due to an improperly designed pipe support. Although the support was designed and installed originally in 2007, the support was broken for many years and was reinstalled on June 30, 2023. The exposure period (T) is the time from when the support was reinstalled until the failure of the weld on November 1, 2023, or 124 days. It was also noted that, during the exposure time, the support was found broken and was reinstalled on a few occasions. As such, the SRA assumed an exposure time for the SDP of T/2 or 62 days to account for the unknown periods of time the support was broken and not stressing the welds. The impact of the weld failure was a loss of one main feedwater (MFW)pump (i.e., the A RFP) due to operator action and the resulting steam release caused the other MFW pump (i.e., the B RFP) to trip due to an electrical short. The SRA modeled the condition using Saphire 8 version 8.2.9 and the Hatch Unit 1 and Unit 2 Standardized Plant Analysis Risk (SPAR) model version 8.82 dated August 13, 2023. The SRA increased the probability of the term IE-LOMFW (initiating event probability for a loss of MFW) by one order of magnitude for the exposure period. The SRA also performed a Bayesian update to the industry initiating event frequency data given the failure and the run time accumulated. This confirmed that the one order of magnitude estimate was conservative. The dominant accident sequence was a loss of MFW event, with a failure to run of the high pressure coolant injection pump and the reactor core isolation cooling pump, with a failure of the operators to manually depressurize the reactor coolant system.
Since the risk from internal events was greater than 1 E-7, internal flooding and seismic sequences were also considered. The licensees probabilistic risk assessment model was considered best available information for internal flooding effects from the MFW system since the SPAR model did not model flooding in the affected flood area. Non-protected electrical equipment in the turbine building was conservatively assumed to fail and modeled in a flag set.
The risk from internal events and internal flooding/seismic was added together and determined to be less than 1 E-6 event per year and therefore, is 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. The licensee failed to adequately evaluate the feedwater drain pipe vibrations in 2007 when the decision was made to install the pipe support without using the design change process. Similar behaviors were seen in 2023 when the licensee did not evaluate the cause(s) of the repeated drain pipe U-bolt failures before it led to a weld failure.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On April 17, 2024, the inspectors presented the integrated inspection results to Jason Butler, plant manager, and other members of the licensee staff.
- On February 8, 2024, the inspectors presented both the ISI and radiation safety inspection results to Matt Busch, site vice president, and other members of the licensee staff.
THIRD PARTY REVIEWS Inspectors reviewed an Institute of Nuclear Power Operations (INPO) report that was issued during the inspection period.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
Resulting from
Inspection
Condition Reports
(CRs)
11045385, 11048614
Naturally Occurring Phenomena
20.3
Cold Weather Checks
18.6
Procedures
NMP-OS-017
Severe Weather
3.1
Residual Heat Removal System
46.6
Procedures
21.7
Corrective Action
Documents
CR 11053941
2/26/2024
NMP-ES-035-
019-GL02-F03
Control Building El. 147
1.0
NMP-ES-035-
019-GL02-F12
U1 Reactor Building El. Below 130
1.0
Fire Plans
NMP-ES-035-
019-GL02-F13
U1 Reactor Building El. 130
1.0
NMP-ES-035-014
Fleet Transient Combustible Controls
7.0
Procedures
NMP-TR-425-F14
Fire Drill Package
03/21/2024
Miscellaneous
001
Hatch 1 Cycle 32 Beginning of Cycle Startup
03/07/2024
Plant Startup
46.5
Normal Plant Shutdown
2.9
NMP-RE-008-F01
Detailed Reactivity Management Plan
2/03/2024
Procedures
NMP-RE-008-F01
Detailed Reactivity Management Plan
01/10/2024
Corrective Action
Documents
CR 11049222
2/12/2024
Engineering
Evaluations
EVAL-H-P41-
05530, EVAL-H-
N21-06251
(a)(1) review
01/11/2024
Procedures
Plant Service Water Operability
15.4
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Unit 1 Plant Service Water Pump & Motor Major
Inspection/Overhaul
10.2
NMP-ES-021
Structural Monitoring Program for the Maintenance Rule
2.0
NMP-ES-027
11.1
SNC1023546
Replace PSW Pump 1P41C001A
Work Orders
SNC1719885
1E11F048B would not stroke from rsdp
2/13/2024
Corrective Action
Documents
CR 11035078
Miscellaneous
1-RICT-2023-02
Risk Informed Completion Time (RICT) worksheets
2/31/2023
Outage Safety Assessment
5.5
MSR, Extraction Steam and Shell Drain System
17.5
Turbine Control Valve Fast Closure Instrument Functional
Test
21.1
NMP-GM-031-
2
Calculation of RMAT and RICT for the RICT Program
6.2
NMP-GM-031-
003
Risk Management Actions for 10CFR50.65(a)(4) and the
Risk Informed Completion Time Program
9.2
NMP-OS-010
Protected Train/Division and Protected Equipment Program
Procedures
NMP-OS-010-002
Hatch Protected Equipment Logs
2.6
BH1-PD-2748
Stress Analysis of Plant Service Water System
6.0
Calculations
SMNH-97-017
Methodology for Sizing Emergency Core Cooling Suction
Strainers
Corrective Action
Documents
Condition Reports
(CRs)
11056813, 11048534, 11055757, 11048856
Engineering
Evaluations
Technical
Evaluations (TEs)
1148116
Miscellaneous
S54711
Unit 1 ECCS Suction Strainer Hydraulic Sizing Report, MPL
E11 and E21
09/11/1997
CR 11042235
Operability Determination Support Basis
01/23/2024
Operability
Evaluations
TE 1147382
Past Operability Review
03/15/2024
RHR Valve Operability
01/14/2024
Procedures
NMP-AD-012
16.1
Work Orders
SNC No.
1735315, 1736869, 1717396, 1734676
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
SNC1423943
1B RHR Loop Valve Operability 34SV-E11-002-1
01/15/2024
Corrective Action
Documents
Resulting from
Inspection
CR 11054239
Procedures
NMP-ES-095-002
Temporary Configuration Change Process
Work Orders
SNC1039107
Corrective Action
Documents
Condition Reports
(CRs)
11022000, 11057287, 1103862, 11046809, 11049455,
11050296
Corrective Action
Documents
Resulting from
Inspection
Condition Reports
(CRs)
11054342, 11054347, 11056923, 11057018, 11057024
H16983
Main Steam Piping Loops C & D
09/23/1981
S17181
Steam Pipe Suspension
Drawings
S17182
Steam Pipe Suspension
NMP-ES-035-
019-GL02-F12
U1 Reactor Building El. Below 130
1.0
Fire Plans
NMP-ES-035-
019-GL02-F13
U1 Reactor Building El. 130
1.0
1-DT-24-1E11-
03004(1)
Outage Maintenance on 1A Loop of RHR
2/20/2024
1-DT-24-1E21-
03102(1)
Isolates Core Spray 'A' for Maintenance and Testing
2/13/2024
1-DT-24-1T47-
03755
Drywell Cooling Unit Isolation
Outage Safety Assessment
03/13/2024
H1C32 Core
Verification Map
Hatch Unit 1 Core, Cycle 32
2/18/2024
2
Hatch 1 Cycle 32 March Startup
03/11/2024
Miscellaneous
TL 2024-0258
Temporary Lift for local leak rate test (LLRT)
Procedures
Outage Safety Assessment
5.5
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
RPV Water Level Corrections
5.16
Fuel Movement Operation
29.1
Plant Startup
46.6
Plant Startup
46.5
Normal Plant Shutdown
2.9
Maintaining Cold Shutdown or Refuel Condition
15.2
Torus and Torus Area Closeout
Control Rod Withdrawal in Shutdown or Refuel
10.16
Drywell Closeout
6.19
Diesel Generator 1C LOCA/LOSP LSFT
2.2
DI-OPS-37-0889
Fuel Movement Rules
9.0
001
Detailed Reactivity Management
2/03/2024
NMP-AD-016
Fatigue Management Program
11.2
NMP-AD-016-004
Reviews and Reporting
11.2
NMP-RE-007
Core Verification
6.0
Condition Reports
(CRs)
11043165, 11048909, 11055649, 11057636, 11049165,
11048909
Corrective Action
Reports (CARs)
608270, 606885
Corrective Action
Documents
Technical
Evaluation (TE)
1147460
1-007A-C1
F028A)
Volumetric leak rate monitor LLRT data sheet
2/07/2024
1-007B-C1
F028B)
Volumetric leak rate monitor LLRT data sheet
2/08/2024
1-007C-C1
F028C)
Volumetric leak rate monitor LLRT data sheet
2/08/2024
Engineering
Evaluations
1-007D-C1
F028D)
Volumetric leak rate monitor LLRT data sheet
2/08/2024
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Turbine Control Valve Fast Closure Instrument Functional
Test
1/10/2024
Core Spray Pump Operability
24.7
HPCI Pump Operability
03/11/2024
HPCI Pump Operability 165 Psig Test
03/10/2024
Diesel Generator 1B 24 Month Operability Test
8.3
Diesel Generator 1C LOCA/LSP LSFT
2.2
LLRT Testing Methodology
13.0
Procedures
Reactor Recirc. Pump Motor Maintenance
16.2
Work Orders
SNC No.
1094633, 1454523, 1718232, 1728349, 1701489, 1212787,
1734664
Corrective Action
Documents
Resulting from
Inspection
CR 11056256
NRC RP inspection re-exit - Green Finding for 2V41 system
03/05/2024
Miscellaneous
S-RP-430
Qualification and Curriculum Status Report for 'RP Shipping
Specialists' [Reviewed 4 records]
2/6/2024
Procedures
NMP-SE-018-002
Physical Protection of Category 1 and 2 Quantities of
Radioactive Material - Hatch Nuclear Plant Instruction
Version 4.0
Radiation
Surveys
Radis survey#
207537
Semi-Annual Inventory and Leak Testing Radioactive
Sources (per NMP-HP-400, Control and Accountability of
Radioactive Sources Inventory)
09/19/2023
IRIS Rpt. 554464
PI Report
03/11/2023
IRIS Rpt. 554609
PI Report
03/31/2023
IRIS Rpt. 559335
PI Report
05/12/2023
IRIS Rpt. 564812
PI Report
08/01/2023
71151
Miscellaneous
IRIS Rpt. 573509
PI Report
11/01/2023
Corrective Action
Documents
Condition Reports
(CRs)
11029735, 11029736, 11036286, 11036278, 11036280,
11036282, 11036283, 11036284, 11036285, 11036290,
11036291, 11037287, 11043293, 10984143, 11007736,
11014034, 11019672
71152A
Corrective Action
Documents
Resulting from
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Inspection
Engineering
Evaluations
Technical
Evaluations (TEs)
1142947, 1126840, 1011857, 1114414, 1116083
0S
Design Control
9.5
NMP-ES-035-
017-002
Hatch Fire Door Inspection Procedure
2.0
NMP-ES-035-
017-002-F03
Swinging Fire Door Inspection
2.0
Procedures
NMP-OS-007-007
Rounds
7.1
Work Orders
SNC No.
1644971, 1528237, 1504851
Corrective Action
Documents
Condition Reports
(CRs)
11058317, 11058287, 11058294, 11032506, 10256426
03/12/2024
Procedures
Scram Procedure
13.5