ML22129A145
| ML22129A145 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/10/2022 |
| From: | Alan Blamey NRC/RGN-II/DRP/RPB2 |
| To: | Gayheart C Southern Nuclear Operating Co |
| Mas-Penaranda D | |
| References | |
| IR 2022001 | |
| Download: ML22129A145 (44) | |
Text
May 10, 2022 Ms. Cheryl Gayheart Regulatory Affairs Director Southern Nuclear Operating Company, INC.
3535 Colonnade Parkway Birmingham, AL 35243
SUBJECT:
EDWIN I. HATCH - INTEGRATED INSPECTION REPORT 05000321/2022001 AND 05000366/2022001
Dear Ms. Gayheart:
On March 31, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Edwin I. Hatch. On April 11, 2022, the NRC inspectors discussed the results of this inspection with Sonny Dean and other members of your staff. The results of this inspection are documented in the enclosed report.
Four findings of very low safety significance (Green) are documented in this report. Three of these findings involved violations of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
A licensee-identified violation which was determined to be of very low safety significance is documented in this report. We are treating this violation as a non-cited violation (NCV) 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 Edwin I. Hatch.
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.
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.
C. Gayheart 2
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 cc w/ encl: Distribution via LISTSERV Signed by Blamey, Alan on 05/10/22
SUNSI Review x
Non-Sensitive
Sensitive x
Publicly Available
Non-Publicly Available OFFICE RII/DRP RII/DRP RII/DRP NAME R. Smith D. Mas-Penaranda A. Blamey DATE 5/5/2022 5/5/2022 5/10/2022
Enclosure U.S. NUCLEAR REGULATORY COMMISSION Inspection Report Docket Numbers:
05000321 and 05000366 License Numbers:
DPR-57 and NPF-5 Report Numbers:
05000321/2022001 and 05000366/2022001 Enterprise Identifier:
I-2022-001-0022 Licensee:
Southern Nuclear Operating Company, INC.
Facility:
Edwin I. Hatch Location:
Baxley, GA Inspection Dates:
January 01, 2022, to March 31, 2022 Inspectors:
J. Bell, Health Physicist J. Diaz-Velez, Senior Health Physicist J. Hickman, Resident Inspector M. Magyar, Reactor Inspector R. Patterson, Senior Reactor Inspector W. Pursley, Health Physicist R. Smith, Senior Resident Inspector Approved By:
Alan J. Blamey, Chief Reactor Projects Branch 2 Division of Reactor Projects
2
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Edwin I. Hatch, 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. A licensee-identified non-cited violation is documented in report section: 71152A.
List of Findings and Violations Incorrect Emergency Action Level Threshold Values for Spent Fuel Pool Level Instrumentation Cornerstone Significance Cross-Cutting Aspect Report Section Emergency Preparedness Green NCV 05000321,05000366/2022001-01 Open/Closed
[H.1] -
Resources 71152A The inspectors identified a Green NCV of Title 10 Code of Federal Regulations (CFR) Part 50.54(q)(2), Title 10 CFR Part 50.47(b)(4), and Title 10 CFR Part 50, Appendix E, Section IV.B for failure to maintain the effectiveness of the Plant Hatch emergency plan and a standard emergency classification scheme based on facility system parameters. Specifically, from October 10, 2017, until November 30, 2021, the licensees emergency classification scheme during an irradiated/spent fuel event RS2 and RG2, contained spent fuel pool (SFP) level threshold values which were outside the indicating range of the level instrumentation. Given these incorrect values, the inspectors determined that the licensee would still have declared a Site Area Emergency (SAE) or General Emergency (GE) in a timely and accurate manner given other emergency action levels (EALs) would also be in effect under these certain conditions.
Failure to Implement the NRC-Approved Emergency Action Levels Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000321,05000366/2022001-02 Open/Closed Not Applicable 71152A The inspectors identified a SL-IV non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(4), for changes made to the Plant Hatch Radiological Emergency Plan (E-Plan), and associated implementing procedure, that required prior NRC approval because the changes were determined to be a reduction in effectiveness of the E-Plan. Specifically, the licensee did not implement the NRC-approved changes prior to making changes to the emergency action levels (EAL) and associated implementing procedure.
Inoperability of Plant Service Water (PSW) and Residual Heat Removal Service Water (RHRSW) Pumps Due to Inadequate Maintenance Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321/2022001-03 Open/Closed
[H.1] -
Resources 71152S
3 The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a for the licensees failure to incorporate adequate maintenance procedural instructions for pump & motor maintenance and overhaul activities as recommended by Regulatory Guide (RG) 1.33, section 9.a. Specifically, preventative maintenance (PM) procedures 52PM-E11-005-1,2 Plant Service Water Pump & Motor Major Inspection/Overhaul and 52PM-P41-036-1,2 RHRSW Pump And Motor Maintenance did not contain instructions to ensure service water pump subcomponents were properly aligned and maintained. This resulted in both PSW and RHRSW pump and motor failures.
Automatic Reactor Shutdown on Low Reactor Water Level Due the Implementation of a Design Change with Single Point Vulnerability Resulting in a Total Loss of Reactor Feed Pumps Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000321/2022001-04 Open/Closed None (NPP) 71153 A self-revealed Green finding was identified when the licensee failed to perform a single point vulnerability (SPV) analysis per licensee procedure NMP-ES-044, Preparation of Design Change Packages. Specifically, the licensee failed to perform this analysis that resulted in implementation of the General Electric (GE) Mark VIe Digital Feedwater system being powered from a single power supply. This resulted in Unit 1 automatic reactor shutdown on low reactor water level (RWL) due to a loss of both running feed pumps and the inability to recover the feed pumps.
Inoperability of 1B Low Pressure Coolant Injection (LPCI) Loop Beyond the Technical Specification Allowed Completion Time Due to Inadequate Maintenance Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321/2022001-05 Open/Closed
[H.14] -
Conservative Bias 71153 A self-revealing Green non-cited violation of Unit 1 Technical Specification (TS) 3.5.1 Emergency Core Cooling Systems (ECCS), RPV water Inventory Control, and Reactor Core Isolation cooling (RCIC) System Conditions A and B, and TS 5.4.1.a Procedures, was identified when the licensee failed to incorporate adequate maintenance procedural instructions for electrical components. Specifically, maintenance procedure did not contain adequate information for fuse installation and component retesting after fuse replacement.
This resulted in the inoperability of B loop of LPCI for approximately 30 days, greater than the TS allowed outage time.
Additional Tracking Items Type Issue Number Title Report Section Status LER 05000321/2021-001-00 LER 2021-001-00 for Edwin L Hatch Nuclear Plant, Unit 1 re Automatic Reactor Scram on Low Reactor Water Level due to Loss of Reactor Feed Pumps 71153 Closed
4 LER 05000321/2022-001-00 LER 2022-001-00 for Edwin I. Hatch Nuclear Plant, Unit 1, Manual Reactor Scram due to Reactor Pressure Perturbations 71153 Closed LER 05000321/2021-002-00 LER 2021-002-00 for Edwin I. Hatch, Unit 1, Low Pressure Coolant Injection Inoperable Longer Than The Allowed Technical Specification Completion Time 71153 Closed
5 PLANT STATUS Unit 1 began the inspection period shutdown from the manual scram on December 29, 2021, due to pressure oscillations caused by a failed main turbine pressure transmitter. The licensee performed repairs by replacing the failed main turbine pressure transmitter. The unit was started up and on January 3, 2022, the unit was returned to 100 percent rated thermal power (RTP). On January 5, 2022, the unit reached maximum core flow and began the coast down in power for the 2022, winter refueling outage. On February 6, 2022, the operators shut down the unit for the scheduled refueling outage. The unit commenced restarted on March 1, 2022, and reached approximately 100 percent RTP on March 8, 2022. On March 26, 2022, the unit was down powered to 90 percent RTP to perform a sequence exchange, the unit was returned to 100 percent RTP the same day. The unit operated there for the remainder of the inspection period.
Unit 2 began the inspection period at 100 percent RTP. On March 18, 2022, the unit was down powered to 48 percent RTP to perform a sequence exchange and a temporary leak repair on the sixth stage B feed water heater manway cover. On March 21, 2022, the unit was returned to 100 percent RTP and operated there for the remainder of the inspection period.
INSPECTION SCOPES Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY 71111.01 - Adverse Weather Protection Impending Severe Weather Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated the adequacy of the overall preparations to protect risk-significant systems from impending severe weather for a tornado watch in the area on January 2, 2022.
External Flooding Sample (IP Section 03.03) (1 Sample)
(1)
The inspectors evaluated that flood protection barriers, mitigation plans, procedures, and equipment are consistent with the licensees design requirements and risk analysis assumptions for coping with external flooding on January 26,2022.
71111.04 - Equipment Alignment
6 Partial Walkdown Sample (IP Section 03.01) (4 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
(1)
Alignment of the Unit 2 reactor core isolation cooling (RCIC) system during repairs of high-pressure coolant injection (HPCI) system, using 34SO-E51-001-2, on February 7, 2022.
(2)
Alignment of the Unit 1 residual heat removal (RHR) system in a shutdown cooling mode lineup following the reactor shutdown for the refueling outage, using 34SO-E11-010-1, on February 8, 2022.
(3)
Alignment of the Unit 1 core spray (CS) system division A during refueling outage 30 due to the division being relied upon for inventory control if needed, using 34SO-E21-001-1, on February 15, 2022.
(4)
Alignment of the Unit 1 decay heat removal (DHR) system during the refueling outage 30 due to being the primary means of decay heat removal, using 34SO-G71-001-0, on February 17-18, 2022.
Complete Walkdown Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated system configurations during a complete walkdown of the Unit 1 standby liquid control system (SBLC) on March 25, 2022.
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)
Center alarm station security building on January 5, 2022.
(2)
Unit 2 reactor building below 130-foot elevation on January 7, 2022.
(3)
Unit 1 reactor building 130-foot elevation on January 25, 2022.
(4)
Unit 1 reactor building main steam chase on February 15, 2022.
(5)
Unit 1 reactor building torus area 87 foot and 114-foot elevations on February 24, 2022.
(6)
Unit 1 radwaste building 156-foot elevation on March 25, 2022.
Fire Brigade Drill Performance Sample (IP Section 03.02) (1 Sample)
(1)
The inspectors evaluated the onsite fire brigade training and performance during an announced fire drill on March 16, 2022.
71111.06 - Flood Protection Measures Inspection Activities - Internal Flooding (IP Section 03.01) (1 Sample)
The inspectors evaluated internal flooding mitigation protections in the:
(1)
Unit 1 and 2 control building 112' level
7
Unit 1 radiological waste system
Unit 1 condensate pump room
Unit 1 and 2 station battery rooms 71111.07A - Heat Exchanger/Sink Performance Annual Review (IP Section 03.01) (1 Sample)
The inspectors evaluated readiness and performance of:
(1)
Unit 1 residual heat removal (RHR) heat exchanger 1B.
71111.08G - Inservice Inspection Activities (BWR)
BWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01) (1 Sample)
(1)
The inspectors evaluated boiling water reactor non-destructive testing by reviewing the following examinations from February 14 - 18, 2022:
1.
Ultrasonic Testing (UT) a.
1B11\\HC-1-H, Meridional Weld Closure Head, ASME Class 1 (reviewed) b.
1B21-1MS-3-23, Pipe to Elbow, ASME Class 1 (reviewed) 2.
Magnetic Particle Testing (MT) a.
1E21-1CS-10A-5PS, E21-CSH-39, ASME Class 1 (reviewed) 3.
Penetrant Testing (PT) a.
1B31-1RC-28A-8BC, Pipe to BC, ASME Class 1 (reviewed) 71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
(1)
The inspectors observed and evaluated licensed operator performance in the control room during plant shutdown, they observed power decrease by reducing reactor recirculation flow to minimum, inserting control rods to 20 percent RTP, followed by a main turbine trip, and the insertion of a manual reactor trip to complete the shutdown.
They then observed the operating crew establish a controlled cooldown rate within technical specification (TS) limits, for the Unit 1 refueling outage, on February 5 and 6, 2022.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
(1)
The inspectors observed and evaluated License Operator Requalification Program (LORP) for the following Scenarios H-LTC-SE-50322, "Shutdown Cooling (SDC)
Operation / Loss of SDC / SOER 09-1," and H-LT-SG-50472, " Security Threat with LOSP or Loss of Ultimate Heat Sink," for one team during simulator training on January 6, 2022.
8 71111.12 - Maintenance Effectiveness Maintenance Effectiveness (IP Section 03.01) (1 Sample)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
(1)
Review of the Unit 1 and 2 pumps and motors failures for the PSW and residual heat removal service water (RHRSW) systems, on January 27, 2022.
Quality Control (IP Section 03.02) (1 Sample)
The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:
(1)
Installation of the Unit 1 main steam line leak detection temperature switch replacement modification, on February 28, 2022.
71111.13 - Maintenance Risk Assessments and Emergent Work Control Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:
(1)
Unit 2 elevated risk due HPCI being made inoperability to repair a steam leak on the steam admission valve 2E41-F001, on February 2, 2022.
(2)
Unit 1 elevated risk as the unit entered refueling outage 30, on February 7, 2022.
(3)
Unit 1 elevated risk due to inventory control being yellow status from February 13 through 18, 2022.
(4)
Unit 1 elevated risk due to placing 1A shutdown cooling in service following refueling activities, on February 24, 2022.
(5)
Unit 1 and 2 elevated risk due to 1B emergency diesel generator aligned to unit 2 during 2C emergency diesel generator maintenance outage, on March 24, 2022.
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) 10850131, Unit 2 HPCI loose nut for seismic support, operability determination.
(2)
CR 10846881, Unit 1 RHRSW with an ineffective seismic restraint, operability determination.
(3)
CR 10852559, Unit 2 PSW pump 2D, dP ratio falling in the required action range, operability determination.
9 (4)
CR 10852218, Unit 1 and 2 failure rate of 120 VAC and 125 VDC molded case circuit breakers (MCCB) and the nonconforming condition presented through lack of MCCB testing, operability determination.
(5)
CR 10855148 Unit 2 HPCI steam leak on the valve body of the 2E41-F095, steam line drain to main condenser, operability determination.
(6)
CR 10858662 Unit 1E 4160 VAC degraded voltage panel wiring error resulting in one of three under voltage relays being jumper out in error during installation.
(7)
CR 10864884, Unit 2 RCIC system room cooler tube sheet damage, operability determination.
71111.18 - Plant Modifications Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
(1)
Temporary modification to install jumper, for the runback #1 circuit on the 1A adjustable speed drive (ASD), on the limit switch for the recirculation discharge valve full open position, to clear the runback #1 signal due to the degraded limit switch.
(2)
Design change to add an uninterruptable power supply (UPS) for the Unit 1 reactor feed pump turbines (RFPTs) digital feedwater control.
Severe Accident Management Guidelines (SAMG) Update (IP Section 03.03) (1 Sample)
(1)
The inspectors verified the site Severe Accident Management Guidelines were updated in accordance with the BWR generic severe accident technical guidelines and validated in accordance with NEI 14-01, Emergency Response Procedures and Guidelines for Beyond Design Basis Events and Severe Accidents, Revision 1.
71111.19 - Post-Maintenance Testing Post-Maintenance Test Sample (IP Section 03.01) (9 Samples)
The inspectors evaluated the following post-maintenance testing activities to verify system operability and/or functionality:
(1) 34SV-P41-003-2, Standby Diesel Service Water System Operability, Version 8.2, following replacement of the discharge check valve and weld repair of the1/4-inch line tap, on February 2, 2022.
(2) 34SV-E41-001-2, High Pressure Coolant Injection Valve Operability, Version 13.7, following valve 2E41-F001 packing replacement, on February 3, 2022.
(3) 57SV-C51-014-1, SRM FT&C, Version 6.4, following the calibration of the source range monitor (SRM) 1B, on February 8, 2022.
(4) 34SV-R43-023-1, Diesel Generator 1A LSFT, Version 2.3, following relay 1E DVR 27-10 replacement and wiring correction, on February 17, 2022.
(5) 57CP-C51-012-0, LPRM I/V Curve Using GE Provided Laptop Program Data Sheet, Version 20.3, 57GM-MIC-016-0, LPRM Cable and Connectors, Version 2.2, and 57IT-C51-001-0, Neutron Monitoring System Detector Pre-installation Tests, Version 5.5,
10 following the replacement of four local power range monitor (LPRMs) detector strings, on February 19, 2022.
(6)
SRM Detector 'D' new cable per work order SNC1154781.
(7) 34SV-B31-001-1, Recirculation System Valve Operability, Version 7.11, following the repair of recirculation discharge valve dual indication and the removal of temporary change configuration SNC1196615, on February 25. 2022.
(8) 42SV-C11-003-0, Control Rod Scram Timing, Version 12.1, following the replacement of 15 Control Rod Drive Mechanisms and six control rod blades, on February 26, 2022.
(9) 34SV-R43-006-2, Diesel Generator 2C Fast Start Test, Version 18.3, following the repair of the 2C emergency diesel generator automatic voltage regulator assembly on March 31, 2022.
71111.20 - Refueling and Other Outage Activities Refueling/Other Outage Sample (IP Section 03.01) (2 Samples)
(1)
The inspectors evaluated forced outage on unit 1 due to manual scram inserted due to pressure oscillations that were not understood at the time of the shutdown and recovery activities, this inspection started in the fourth quarter of 2021 and the inspection activities were completed on January 4, 2022.
(2)
The inspectors evaluated unit 1 refueling outage 30 activities from February 5 to March 3, 2022.
71111.22 - Surveillance Testing The inspectors evaluated the following surveillance testing activities to verify system operability and/or functionality:
Surveillance Tests (other) (IP Section 03.01) (6 Samples)
(1) 34SV-R43-001-2, Diesel Generator 2A Monthly Test, Version 30.1, on January 11, 2022.
(2) 57SV-SUV-004-1, Excessive flow Check Valve (EFCV) Operability. Version 16.10, on February 2, 2022.
(3) 34SV-R43-020-1, Diesel Generator 1A LOCA/LOSP Logic System Functional Test (LSFT), Version 2.5, on February 9, 2022.
(4) 34SV-B21-012-1, Low Low Set LSFT, Version 1.2, on February 18, 2022.
(5) 34SV-R43-024-1, Diesel Generator 1B LSFT (From U1), Version 3.1, on February 24, 2022.
(6) 42SV-TET-003-1, Primary Containment Integrated Leak Rate Test, Version 6.0. on February 26. 2022.
Inservice Testing (IP Section 03.01) (1 Sample)
(1) 34SV-E51-002-2, Reactor Core Isolation Cooling Pump Operability IST, Version 27.1 on January 6, 2022.
Containment Isolation Valve Testing (IP Section 03.01) (1 Sample)
11 (1) 42SV-TET-001-0, Primary Containment Type B and Type C Leak Rate Testing, Version 30.2, for 1B21-F022A/B/C/D and 1B21-F028A/B/C/D inboard and outboard main steam isolation valves (MSIVs) on February 9, 2022.
FLEX Testing (IP Section 03.02) (1 Sample)
(1)
Work Order SNC1138590, FL60KW3Y-H, FLEX 60 KW Generator Performance Test Work Instructions, on March 30, 2022.
71114.06 - Drill Evaluation Drill/Training Evolution Observation (IP Section 03.02) (1 Sample)
The inspectors evaluated:
(1)
Licensed reactor operators respond to a HPCI steam line break in the simulator and the inspectors observed the crew make an emergency declaration and notification. This represented a drill and exercise performance opportunity on March 30, 2022.
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 during unit one (U1) refueling outage number 30 (H1R30).
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 during H1R30 including as low as reasonably achievable (ALARA) plans and radiation work permits (RWPs).
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 exiting the radiologically controlled area (RCA) at the radiation protection control point during H1R30.
(2)
Licensee surveys of potentially contaminated material leaving the contaminated area at the U1 drywell access.
(3)
Licensee surveys in the reactor building railroad bay for contamination in the bay itself and potentially contaminated material in this area for release from the RCA during H1R30.
12 Radiological Hazards Control and Work Coverage (IP Section 03.04) (5 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
(1)
RWP 22-1614, Sub-Pile Room Work - Carousel PMs / Repairs, Lighting, Cable Pulls, and LPRMs / RPIS / Shootout Steel / CRD Mob / Demob during H1R30 (2)
RWP 22-1611 DW - Radiological High Risk / Work Involving Removal or Exposure to Irradiated Materials and High Rad Decon during H1R30.
(3)
RWP 22-1601 RP Job Coverage, Surveys, and Chemistry Supporting Activities in U1 Drywell during H1R30 (4)
RWP 22-1205 Rx Vessel Disassembly/Reassembly, Cavity/Dryer Separator Work &
Support during H1R30.
(5)
RWP 22-1208 Unload/Reload Core; Move/Shuffle Fuel on the refuel floor during H1R30.
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 and Very High Radiation Areas:
(1)
Unit 1 reactor water cleanup pump room A.
(2)
Unit 1 reactor water cleanup pump room B (3)
Unit 2 turbine building northeast hallway ladder on 164-foot elevation.
(4)
Unit 2 TIP room unit 2 reactor building, 130-foot elevation.
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 during H1R30.
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &
Transportation Radioactive Material Storage (IP Section 03.01) (2 Samples)
The inspectors evaluated the licensees performance in controlling, labeling, and securing the following radioactive materials:
(1)
Radioactive materials in the Radwaste Systems and Components areas while performing a walkdown of the systems.
(2)
Radioactive materials stored in the Interim Radwaste Storage Facility while observing performance of a licensee surveillance procedure.
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:
13 (1)
Waked down unit 2 (U2) liquid radwaste systems including U2 waste surge tank, U2 waste collector tank, U2 floor drain collector tank.
(2)
Walked down U2 radwaste control room boards.
(3)
Walked down accessible portions of U2 solid radwaste systems.
Waste Characterization and Classification (IP Section 03.03) (3 Samples)
(1)
Reviewed sample ID # 20Dec20-009 SF, scaling factors for DAW.
(2)
Reviewed sample ID # 20Dec20-007 SF, 2021 scaling factors for unit 1 composite spent resins.
(3)
Reviewed sample ID# 24Feb21-018 SF, 2021 scaling factors for unit 2 torus filters.
Shipment Preparation (IP Section 03.04) (1 Sample)
(1)
The inspectors observed the preparation of radioactive shipment #22-RM-006.
Shipping Records (IP Section 03.05) (5 Samples)
The inspectors evaluated the following non-excepted radioactive material shipments through a record review:
(1)
Reviewed shipment 22-RM-006, LSA-1 shipment of DAW.
(2)
Reviewed shipment 21-RW-035, Type B shipment of resin.
(3)
Reviewed shipment 20-RM-048, Type A shipment.
(4)
Reviewed shipment 20-RW-038, LSA-1 shipment of DAW.
(5)
Reviewed shipment 21-RW-017, LSA-II shipment of dewatered resin.
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) (2 Samples)
(1)
Unit 1 (January 1, 2021, to December 31, 2021)
(2)
Unit 2 (January 1, 2021, to December 31, 2021)
IE03: Unplanned Power Changes per 7000 Critical Hours Sample (IP Section 02.02) (2 Samples)
(1)
Unit 1 (January 1, 2021, to December 31, 2021)
(2)
Unit 2 (January 1, 2021, to December 31, 2021)
IE04: Unplanned Scrams with Complications (USwC) Sample (IP Section 02.03) (2 Samples)
(1)
Unit 1 (January 1, 2021, to December 31, 2021)
(2)
Unit 2 (January 1, 2021, to December 31, 2021)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
14 (1)
U1 February 1, 2021, through January 31, 2022 U2 February 1, 2021, through January 31, 2022 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
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U1 February 1, 2021, through January 31, 2022 U2 February 1, 2021, through January 31, 2022 71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03) (3 Samples)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
(1)
CR 10855849 and corrective action report (CAR) 279683, review of corrective actions for tritiated spill that was contained on February 2, 2022, and corrective actions from tritium issues that occurred in 2021.
(2)
The inspectors conducted focus groups and interviews of over 30 personnel in the radiation protection department, reviewed the corrective actions associated with Employee Concerns Program (ECP) and Request for Information, safety conscious work environment (SCWE) evaluations, and reviewed notes from recent safety culture monitoring panels. The inspectors determined that personnel are willing to raise nuclear safety concerns and feel they are empowered to stop work when they identify issues. All employees stated that they would raise nuclear safety concerns through multiple avenues without fear of retaliation. All employees stated that they had confidence in the stations ECP and would not hesitate to use it should they need to. The inspection team verified that the corrective actions developed by the station were able to repair the trust within the RP organization and allow employees to feel comfortable raising concerns without fear of retaliation. Specifically, changes in management and the filling available supervisor positions and the addition of one supervisor position seems to have significantly supported the safety conscious work environment.
(3)
CR10840663 and CAR 287331, review of corrective actions for EAL threshold values incorrect for the spent fuel pool level for RG2 and RS2 on March 16, 2022.
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02) (1 Sample)
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Semi-annual trend review of pump/motor failures of the PSW and RHRSW pumps during cycle year 2021 year.
71153 - Follow Up of Events and Notices of Enforcement Discretion Event Followup (IP Section 03.01) (1 Sample)
15 (1)
The inspectors evaluated the overflow spill of the outfall tank that contained tritiated water that was contained and licensees response on February 3, 2022.
Event Report (IP Section 03.02) (3 Samples)
The inspectors evaluated the following licensee event reports (LERs):
(1)
LER 05000321/2021-001-00, Unit 1 Automatic Reactor Scram on Low Reactor Water Level due to Loss of Reactor Feed Pumps (ADAMS Accession No. ML21274A418) on August 3, 2021. The circumstances surrounding this LER are documented in this inspection report, in the Inspection Results section (FIN 05000321/2022004).
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LER 05000321/2021-002-00, Unit 1 Low Pressure Coolant Injection Inoperable Longer than the Allowed Technical Specification Completion Time (ADAMS Accession No. ML21274A738) on August 3, 2021. The circumstances surrounding this LER are documented in this inspection report, in the Inspection Results section (NCV 05000321/2022005).
(3)
LER 05000321/2022-001-00, Unit 1 Manual Reactor Scram due to Reactor Pressure Perturbations (ADAMS Accession No. ML22049B535) on December 29, 2021. The inspectors determined that it was not reasonable to foresee or correct the cause discussed in the LER therefore no performance deficiency was identified.
INSPECTION RESULTS Incorrect Emergency Action Level Threshold Values for Spent Fuel Pool Level Instrumentation Cornerstone Significance Cross-Cutting Aspect Report Section Emergency Preparedness Green NCV 05000321,05000366/2022001-01 Open/Closed
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Resources 71152A The inspectors identified a Green NCV of Title 10 Code of Federal Regulations (CFR) Part 50.54(q)(2), Title 10 CFR Part 50.47(b)(4), and Title 10 CFR Part 50, Appendix E, Section IV.B for failure to maintain the effectiveness of the Plant Hatch emergency plan and a standard emergency classification scheme based on facility system parameters. Specifically, from October 10, 2017, until November 30, 2021, the licensees emergency classification scheme during an irradiated/spent fuel event RS2 and RG2, contained spent fuel pool (SFP) level threshold values which were outside the indicating range of the level instrumentation. Given these incorrect values, the inspectors determined that the licensee would still have declared a Site Area Emergency (SAE) or General Emergency (GE) in a timely and accurate manner given other emergency action levels (EALs) would also be in effect under these certain conditions.
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Description:==
On October 10, 2017, NMP-EP-141-002, Ver. 2.0, Hatch Emergency Action Levels (EALs) and Basis (NEI 99-01 Rev. 6) was approved for implementation. One of the items being implemented was replacing the SFP level EAL placeholder value with an actual instrument reading. Version 1.0 of the subject NMP procedure reflected the Rev. 4 EAL scheme. Version 2.0 of the NMP procedure was written to reflect the Rev. 6 EAL scheme.
A change in the writer's guide for NMP procedures resulted in significant formatting changes, along with other non-technical changes, from what was approved in the Safety Evaluation
16 Report (SER) (ML17023A237). The SER included changes made as a result of a Request for Additional Information response. For example, placeholder values were added for RA2, RS2, and RG2. However, plant modifications were not completed at the time of the NEI 99-01 Rev. 6 submittal to the NRC. The NRC-provided guidance described as acceptable, the SFP level EALs in a grayed-out format and once plant modifications were completed, the licensee could replace the placeholder values (Level 3 and Level 2) with actual instrument readings. Specifically, the revision changed RS2 from Level 2 to: spent fuel pool level at 1.4 feet; Operating Modes: All; and RG2 from Level 3 to: spent fuel pool level cannot be restored to at least 1.4 feet for 60 minutes or longer; Operating Modes: All. However, the SFP value of 1.4 feet was incorrect because the value did not consider a 4-inch dead zone from the bottom of the detector (at 1.33 feet) which would give the lowest detectable level of 1.67 feet. Therefore, 1.4 feet would not be indicated on the instrument to classify RS2 and RG2. On November 11, 2021, the Emergency Preparedness (EP) Department recognized that there was a difference between the calculation design output and the EAL threshold value and wrote a condition report (CR) to evaluate the threshold value of the spent fuel pool level for EALs RS2 and RG2.
The inspectors reviewed the CR and discussed with the licensee the significance of the error, which was not evident in the initial CR wording. Further investigation by the licensee found that an uncontrolled and inaccurate figure/diagram was used as an input to determine the SFP EAL threshold value for RS2 and RG2. Additionally, the licensees EP procedure for maintaining the emergency plan, did not contain adequate guidance to ensure source document verification and validation was used to support EAL changes. The issue of inadequate guidance will be pursued in a separate violation write-up included in the Inspection Results section of this report.
The inspectors reviewed the compiled information describing timeframes, types of errors, and the impact of those errors, to help understand the significance of the errors. The licensees ability to declare emergencies based on facility system parameter values was determined to be ineffective, but because event classification using other EALs would still result in a SAE and GE being declared in a timely and accurate manner for the initiating conditions RS2 and RG2 in NMP-EP-141-002, Hatch Emergency Action Levels and Basis (NEl-99-01 Revision 6, NMP-EP-141-002-F01, Hatch - Hot Initiating Condition Matrix and NMP-EP-141-002-F02, Hatch-Cold Initiating Condition Matrix), the significance of the issue becomes less (see Significance Section below).
Corrective Actions: The licensee entered the issue into the corrective action program on November 11, 2021. Other actions taken by the licensee were to issue a Standing Order (November 15, 2021), correct the EALs (November 30, 2021), and initiate a root cause evaluation.
Corrective Action
References:
Condition Report (CR) 10840695, Technical Evaluation (TE) 1098288, and Corrective Action Report (CAR) 287331.
Performance Assessment:
Performance Deficiency: The failure to maintain the effectiveness of an emergency plan to meet the requirements of Title 10 CFR Part 50.47(b)(4) and Part 50 Appendix E to have a standardized EAL scheme with adequate methods, systems, and equipment in use based on facility system and effluent parameters for assessing and monitoring actual or potential offsite consequences of a radiological emergency, was a performance deficiency.
17 Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, for a decreasing SFP level under certain plant conditions, EALs RS2 and RG2 were rendered ineffective, but other EALs would result in equitable classification levels being declared in a timely and accurate manner, as described in Inspection Manual Chapter 0609, Appendix B, Table 5.4-1, Significance Determination for Ineffective EALs and Over classification.
Significance: The inspectors assessed the significance of the finding using Appendix B, Emergency Preparedness SDP. And using the corresponding Table 5.4-1, Significance Examples §50.47(b)(4) (issue date September 22, 2015), the performance deficiency was determined to have very low safety significance (Green) because a SAE and GE EAL had been rendered ineffective, but because of other EALs, an equitable classification would be made in an accurate and timely manner.
Cross-Cutting Aspect: H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. The cause of the finding was determined to be associated with a cross-cutting aspect in the Resources component of the Human Performance area because the licensee failed to provide adequate procedural guidance to ensure nuclear safety is maintained when making EAL changes. Specifically, EP procedure NMP-EP-310, Maintaining the Emergency Plan, did not ensure source documents were verified and validated prior to making EAL changes.
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Enforcement:
Violation: Title 10 CFR Part 50.54(q)(2) requires that a holder of a nuclear power reactor operating license under this part, shall follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(4) requires a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Title 10 CFR Part 50, Appendix E, Section IV.B., Assessment Actions, requires in part, that the means to be used for determining the magnitude of, and for continuously assessing the impact of, the release of radioactive materials shall be described, including EALs that are to be used as criteria for determining the need for notification and participation of local and State agencies, the Commission, and other federal agencies, and the EALs that are to be used for determining when and what type of protective measures should be considered within and outside the site boundary to protect health and safety. The EALs shall be based on in-plant conditions and instrumentation, in addition to onsite and offsite monitoring.
Contrary to the above, since October 10, 2017, until November 30, 2021, the licensee failed to maintain the effectiveness of their emergency plan and a standard emergency classification scheme based on facility system parameters. Specifically, the licensees emergency classification scheme during an irradiated/spent fuel event RG2 & RS2, contained SFP level threshold values which were outside the indicating range of the instrument, but would result in declaring a SAE or GE in a timely and accurate manner given other EALs would also be in effect. The failure to maintain the effectiveness of an emergency plan to
18 meet the requirements of 10 CFR Part 50.47(b)(4) and Part 50 Appendix E is identified as NCV 05000321, 366/2022001-01, Incorrect Emergency Action Level Threshold Values for Spent Fuel Pool Level Instrumentation Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Implement the NRC-Approved Emergency Action Levels Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000321,05000366/2022001-02 Open/Closed Not Applicable 71152A The inspectors identified a SL-IV non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(4), for changes made to the Plant Hatch Radiological Emergency Plan (E-Plan), and associated implementing procedure, that required prior NRC approval because the changes were determined to be a reduction in effectiveness of the E-Plan. Specifically, the licensee did not implement the NRC-approved changes prior to making changes to the emergency action levels (EAL) and associated implementing procedure.
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Description:==
On March 3, 2016, Southern Nuclear Operating Company (SNC) requested amendments to the licenses for its three nuclear operating sites to adopt the NEI 99-01 Revision 6 EAL schemes. For Plant Hatch, this included spent fuel pool (SFP) EALs RS2 and RG2 with threshold values listed as Level 3. On August 29, 2016, the NRC issued a request for additional information that asked the licensee to confirm that all EAL setpoints and indications used in the proposed EAL scheme were within the calibrated range(s) of the instrumentation. The SNC response confirmed that the proposed EAL scheme setpoints and indications were within the calibrated ranges. On March 16, 2017, the NRC issued a Safety Evaluation Report (SER) that approved the proposed EAL scheme change. The SER did not specify values for EALs RS2 and RG2 since SNC had not yet completed the modifications for SFP level instrumentation. However, the Level 3 value was an acceptable threshold for the purposes of issuing the SER. On October 14, 2017, SNC documented in a 50.54(q) change evaluation, the move from NEI 99-01 Revision 4 EALs to NEI 99-01 Revision 6 EALs. Part of this change included replacement of the Level 3 threshold value for EALs RS2 and RG2 with actual values derived from an Operations procedure. These values were later (November 11, 2021) determined to be incorrect and are being dispositioned as an NRC violation in the Inspection Results section of this report. The inspectors determined that not implementing the NRC-approved EALs as written, but instead making changes to these EALs prior to implementation, was in violation of the regulatory process. In addition, when the change to the SFP EAL threshold values were made, incorrect values were used and caused a reduction in the effectiveness of the emergency plan and EALs.
Corrective Actions: The licensee entered the issue into the corrective action program.
Corrective Action
References:
Condition Report 10873304.
Performance Assessment: The licensees failure to implement the NRC-approved changes to the EALs and associated implementing procedure was determined to impede the NRCs ability to perform its regulatory function and is dispositioned using the Traditional Enforcement process.
19 Enforcement: This finding is a violation of NRC requirements because it had the potential for impacting the NRCs ability to perform its regulatory function. Therefore, traditional enforcement is applicable, in accordance with Inspection Manual Chapter 0611 and 0612, Appendix B, Figure 2. This finding is determined to be a SL-IV violation in accordance with Section 6.6.d.1 of the Enforcement Policy because it involves the licensees ability to meet or implement a regulatory requirement not related to assessment or notification such that the effectiveness of the emergency plan is reduced.
Violation: Title 10 of the Code of Federal Regulations, Part 50.54(q)(4) states, in part, that changes to a licensees E-Plan that reduce the effectiveness of the plan, may not be implemented without prior approval of the NRC. Contrary to the above, the licensee failed to implement the NRC-approved changes to their E-Plan, EALs, and the associated implementing procedure. Specifically, the licensee made changes to the EALs prior to implementing the NRC-approved EALs and because the changes were determined to be incorrect EAL threshold values, a reduction in effectiveness of the emergency plan and EALs resulted.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Licensee-Identified Non-Cited Violation 71152A This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Violation: The licensee identified a non-cited violation (NCV) of 10 CFR 50.47(b)(16) for the failure to maintain the effectiveness of its emergency plan by ensuring procedures for use by the emergency preparedness (EP) organization are adequately maintained. The licensee identified some procedural inadequacies following their root cause evaluation for a spent fuel pool (SFP) emergency action level (EAL) discrepancy. The licensee determined that an uncontrolled and inaccurate figure/diagram was used as an input to determine the SFP EAL threshold value for RS2 (site area emergency) and RG2 (general emergency). Contrary to the above, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, the licensees emergency preparedness implementing procedure for maintaining the emergency plan did not contain adequate guidance to ensure source document verification and validation was used to support EAL changes.
Significance/Severity: Green.
Corrective Action
References:
Condition Report 10840695, Technical Evaluation 1098288, and Corrective Action Report 287331.
Inoperability of Plant Service Water (PSW) and Residual Heat Removal Service Water (RHRSW) Pumps Due to Inadequate Maintenance Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321/2022001-03 Open/Closed
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Resources 71152S
20 The inspectors identified a Green non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a for the licensees failure to incorporate adequate maintenance procedural instructions for pump & motor maintenance and overhaul activities as recommended by Regulatory Guide (RG) 1.33, section 9.a. Specifically, preventative maintenance (PM) procedures 52PM-E11-005-1,2 Plant Service Water Pump & Motor Major Inspection/Overhaul and 52PM-P41-036-1,2 RHRSW Pump And Motor Maintenance did not contain instructions to ensure service water pump subcomponents were properly aligned and maintained. This resulted in both PSW and RHRSW pump and motor failures.
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Description:==
Since 2019, several service water pumps had been replaced due to degradation to flow and differential pressure. Per corrective action report (CAR) 277075, these failures were attributed to silting issues that have challenged the service life of these pumps.
However, three service water pumps failed shortly after being replaced. Failure analyses were performed, and the common failure was due to alignment issues that resulted in a hard rub on the pump shaft to its stuffing box. This led to either the pump shaft seizing to the bushing or in one case shearing. After a pump overhaul, the 1B RHRSW pump ran for five minutes before the shaft seized and eventual sheared. The failure analysis concluded that the pump failure was attributed to significant misalignment of the pump shaft through the discharge head. Additionally, after a pump overhaul, the 1D RHRSW pump ran for seven months prior to the pumps shaft seizing to the discharge head. Failure analysis identified that the pump failure was caused by the out of dimensional and geometric tolerance of parts (discharge head) and the use of the wrong bearing retainers during column assembly. Finally, after a pump overhaul, the 1C PSW pump ran for 15 seconds before the pump seized to the discharge head. The pumps motor lower bearings were wiped, and the discharge head bore was found out of concentricity and flanges were out of parallelism. Additionally, the pump motor thrust bearing was found wiped. This resulted in the pump shaft shearing and its rotating assembly detaching from the column.
The inspectors reviewed the service water pump trend data which included the associated root cause analysis CAR 279720, multiple corrective action documents and associated PM and in-service test procedures. It was identified that the sites PM procedure for the PSW and RHRSW pumps lacked sufficient detail to ensure subcomponents were properly aligned and maintained. Specifically, pump subcomponents and their associated surfaces were allowed to corrode and erode causing significant enough wear to challenge the sites ability to ensure proper pump alignment. Furthermore, pump subcomponents such as the pump discharge heads for both the PSW and RHRSW pumps were routinely reused without proper inspections being performed to ensure adequate clearances between the pump discharge head and sub-base were maintained.
Corrective Actions: The licensee entered this issue into their corrective action Program as CR 10855728. Additionally, the licensee replaced/refurbished the three service water pumps subcomponents so appropriate alignment could be achieved and revised applicable job plans/procedures to incorporate industry best practices.
Corrective Action
References:
CR 10855728, CAR 279720, and corrective action report (CAR) 277075 Performance Assessment:
Performance Deficiency: The failure to provide appropriate procedures, instructions, or drawings for maintenance that could affect the performance of safety-related equipment per TS 5.4.1.a and Regulatory Guide 1.33 was a performance deficiency. Specifically, there were
21 no instructions or drawings to ensure that the service water pumps were properly maintained and aligned resulting in pump and motor failures.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the lack of procedural guidance ensuring proper alignment at maintenance impacted the reliability of the PSW and RHRSW pumps to perform their functions.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The affected cornerstone was Mitigating Systems, as determined by Inspection Manual Chapter (IMC) 0609,, Initial Characterization of Findings. Utilizing IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the screening questions from Exhibit 2 of the Mitigating Systems section were utilized and the mitigating SSCs and PRA functionality questions were addressed. The performance deficiency was screened to Green because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC) and the service water system maintained its operability and PRA functionality.
Cross-Cutting Aspect: H.1 - Resources: Leaders ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety.
Specifically, from 2019 to 2021 PM procedures 52PM-E11-005-1, 52PM-E11-005-2, 52PM-P41-036-1, 52PM-P41-036-2 were all revised several times giving sufficient opportunities to ensure each procedure contained adequate instructions for proper maintenance and alignment.
Enforcement:
Violation: Technical Specification 5.4.1.a Procedures, states, in part, that written procedures shall be implemented covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, February 1978. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Appendix A, Paragraph 9.a, Procedures for Performing Maintenance, requires that maintenance that can affect the performance of safety related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
Contrary to the above, since 2019, maintenance procedures 52PM-E11-005-1, 52PM-E11-005-2, 52PM-P41-036-1, 52PM-P41-036-2 did not provide procedures, instructions, or drawings regarding the PSW and RHRSW pump and motor replacement that were appropriate to the circumstances. Specifically, there were no instructions or drawings to ensure that subcomponents (i.e., discharge head/sub-base) surfaces were not degraded, bolting torque requirements were specified and sufficient clearances were maintained to prevent occurrences of misalignment resulting in PSW and RHRSW pump and motor failures.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
22 Automatic Reactor Shutdown on Low Reactor Water Level Due the Implementation of a Design Change with Single Point Vulnerability Resulting in a Total Loss of Reactor Feed Pumps Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000321/2022001-04 Open/Closed None (NPP) 71153 A self-revealed Green finding was identified when the licensee failed to perform a single point vulnerability (SPV) analysis per licensee procedure NMP-ES-044, Preparation of Design Change Packages. Specifically, the licensee failed to perform this analysis that resulted in implementation of the General Electric (GE) Mark VIe Digital Feedwater system being powered from a single power supply. This resulted in Unit 1 automatic reactor shutdown on low reactor water level (RWL) due to a loss of both running feed pumps and the inability to recover the feed pumps.
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Description:==
On August 3, 2021, at 10:26 AM Unit 1 automatically scrammed from a RWL low level signal +3 inches caused by a complete loss of both operating reactor feed pumps. The cause of the loss of the operating reactor feed pumps was a trip of the normal power supply breaker to the 1R23-S021 bus. This bus powers both redundant power supplies to the feed water control system. Loss of this bus causes up to a 0.5 second time delay to transfer to alternate power and a momentary loss of power to the reactor feed pump control system that results in a trip of both reactor feed pumps.
Additionally, this automatic reactor shutdown became complicated when the undervoltage relay for auto bus transfer logic failed to actuate, preventing switchgear 1R23-S021 to swap to the alternate power supply. This failure to transfer power caused a complete loss of power to Bus 1R23-S021 after the 0.5 second maximum time to transfer had occurred. This resulted in sustained loss of power to distribution panel 1R25-S021 that supplies power to both the 1A and 1B reactor feed pumps control system. This resulted in the inability of the operating crew to recover the reactor feed pump turbines (RFPTs) after the automatic shutdown.
After the automatic shutdown, RWL continued to decrease, because there were no operating feed pumps. Water level decreased to the automatic start for high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC). The plant operators used these systems to restore RWL to the normal band of +32 inches to +42 inches. Furthermore, the reactor recirculation pumps tripped on low RWL. Operators stabilized the plant and proceeded to cold shutdown.
The licensee determined the direct cause of the loss of normal power supply to switchgear 1R23-S021 was spurious malfunction of the trip unit for the supply breaker. The reason for the failure of the undervoltage relay to prevent the auto swap to alternate power was determined to be incorrect configuration of the relay from the time it was installed during plant construction and the failure to perform a post maintenance test after installation in 1978. The licensee also determined that having redundant power supplies coming from the same power supply was a vulnerability missed during the implementation of the GEs Mark VIe reactor feed pump controls design change. The licensee had erroneously given credit to a GE failure modes analysis, that stated the Mark VIe had sufficient redundance built into the system that there was no SPVs identified in the Mark VIe. Although this failure mode analysis was correct, it did not look at power supplies to the system. The licensee elected not to perform
23 the SPV analysis form that was optional per the design implementation procedure at the time based on this GE information. This modification was implemented in 2014. Since then, the licensee has changed its design change procedure by adding additional steps for a more rigorous check for SPVs.
Corrective Actions: The licensee replaced the tripping device for the breaker that initiated event. They also replaced the transfer and undervoltage relays that contributed to cause of the failure for the bus to auto transfer. They satisfactorily performed testing of these relays and the trip device prior to restarting the unit. The licensee has developed a design change to add an uninterruptible power supply (UPS) to the Mark VIe digital feed water control system to maintain the power to the feed water pump controls during the bus transfer if such event should occur again; this would prevent a loss of reactor feed pumps or a low RWL automatic shutdown. This UPS has been implemented in the 2022 spring Unit 1 refueling outage.
Corrective Action
References:
Condition report (CR) 10817773; technical evaluations (TE) 10993164, 1093165, and 1093166; and corrective action report (CAR) 279607.
Performance Assessment:
Performance Deficiency: The inspectors determined the licensees failure to properly perform a single point vulnerability analysis per the licensees procedure NMP-ES-044, Preparation of Design Change Packages was a performance deficiency. This resulted in the implementation of the GE Mark Vle digital feedwater system being powered from a single power supply and resulted in an automatic reactor shutdown and a complicated automatic reactor shutdown due to a total loss of reactor feed water.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance 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 failure to properly analysis for SPV during a design change of digital feed water controls resulted in complicated scram from a total loss of feed water.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The affected cornerstone was Initiating Events, as determined by Inspection Manual Chapter (IMC) 0609, Attachment 4, Initial Characterization of Findings. Utilizing IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the screening questions from Exhibit 1 of the Initiating Event section were utilized and the transient initiator questions were addressed.
Although this finding did cause a reactor trip, it did not result in a complete loss of the condensate system due the fact the licensee still had all condensate and condensate booster pumps available and a viable flow path, thus resulting in this finding being of very low safety significance (Green).
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement: Inspectors did not identify a violation of regulatory requirements associated with this finding.
24 Inoperability of 1B Low Pressure Coolant Injection (LPCI) Loop Beyond the Technical Specification Allowed Completion Time Due to Inadequate Maintenance Procedure Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321/2022001-05 Open/Closed
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Conservative Bias 71153 A self-revealing Green non-cited violation of Unit 1 Technical Specification (TS) 3.5.1 Emergency Core Cooling Systems (ECCS), RPV water Inventory Control, and Reactor Core Isolation cooling (RCIC) System Conditions A and B, and TS 5.4.1.a Procedures, was identified when the licensee failed to incorporate adequate maintenance procedural instructions for electrical components. Specifically, maintenance procedure did not contain adequate information for fuse installation and component retesting after fuse replacement.
This resulted in the inoperability of B loop of LPCI for approximately 30 days, greater than the TS allowed outage time.
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Description:==
On July 3, 2021, the licensee observed a loss of light indication on recirculation pump discharge isolation valve 1B31F031B. The licensees investigation found the 30A control power fuses had failed (i.e., blown). The fuses were replaced, and then blew again when valve operation was attempted. Troubleshooting was performed which included removal and reinstallation of the control power fuses. It was later determined that a valve motor failure caused the blown fuses. The licensee replaced the valve motor, successfully stroked the valve to both the full open and full close positions and declared the valve operable on July 5, 2021.
On July 5, 2021, the recirculation discharge valve lost light indication again. The licensee replaced the bulb multiple times, and the light would illuminate for approximately 15 seconds each time before going out. Further, troubleshooting was performed by electrical maintenance and the light indication returned and remained illuminated. The electrical maintenance personnel felt that pressing on terminations within the motor control center (MCC) during troubleshooting may have been the reason the valves light indication returned. The electrical maintenance personnel checked the tightness and voltage for the termination points and checked from screw to screw on the fuse holder that bolted the lugs to the fuse holder in place. Based on these satisfactory readings, troubleshooting effort was stopped, and no retest of the discharge valve was performed. Although, no firm conclusion was determined from the troubleshooting, the licensee did not write a condition report (CR) documenting these facts.
On July 6, 2021, operations requested engineering to provide an input on potential impacts to the valves operability during a Loss of Coolant Accident (LOCA) based on the discharge valve showing intermittent light indication during troubleshooting. The engineering department identified a potential issue with the control fuses/control fuse holders or the limit switch, and that if it was an issue with the control fuses, this could present a challenge to discharge valve operation. Since light indication had returned, operations did not allow additional troubleshooting to be performed based on the reactor being at power and risk of losing recirculation flow if the discharge valve closed. On July 19, 2021, operations determined that the valve was operable based on continuous light indication since the terminations were checked and historical light indication trends for the valve did not impact the valve function or operation.
25 Following an automatic reactor scram on August 3, 2021, when an operator attempted to close the discharge valve 1B31F031B for the B recirculation pump per procedure, the light indication was lost. Operations dispatched maintenance personnel to the MCC and neither the breaker nor the overloads were tripped. The light indication returned with no action performed by maintenance. Additionally, when troubleshooting was performed on August 5, 2021, maintenance personnel found all connections were tight, but a control fuse holder was loose. The fuse holder bayonet connectors were spread which allowed the control fuses to move freely which resulted in a momentary loss of connection that caused the light indication issue for the discharge valve. This also meant the valve would not perform its safety function to close if a LOCA occurred. This condition was then resolved through replacement of the fuse holder. Operations performed a stroke test for the valve and declared it to be operable.
During the investigation into the issue, it was determined the initial loss of light indication investigation from July 3-5, 2021, resulted in replacing the discharge valve motor.
Maintenance troubleshooting activities of removing and inserting the control power fuses during the motor replacement exacerbated the degraded fuse holders bayonet connectors. Furthermore, a gap in the post maintenance testing guidance for fuse replacement prevented initial identification of the problem, in that the procedure did not contain adequate steps to ensure fuse tightness and component retesting after fuse replacement.
Corrective Actions: The licensee entered this issue into their corrective actions program. The licensee immediately corrected the issue by replacing the fuse holders following the August 5, 2021, investigation. The licensee performed a gap analysis of fuse replacement process and revised NMP-MA-014-001, Post Maintenance Testing Guide, Version 5.8, to ensure the PMT process is performed in area of fuse installation and requiring component retesting after fuse replacement. Additionally, the licensee coached individuals involved in the CR review process for the incorrect operability determination.
Corrective Action
References:
CRs 10810391, 10810631, 10817701, and 10827732; technical evaluations (TEs) 1094466 and 1093497, and corrective action report (CAR) 279617.
Performance Assessment:
Performance Deficiency: The inspectors determined that an inadequate maintenance procedure for performing post maintenance test instructions for fuse installation and component retesting after fuse replacement, resulted in the inoperability of B loop of LPCI for greater than its allowed TS outage time and was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the B loop of LPCI was inoperable beyond its allowed completion time due to inadequate post maintenance testing.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The affected cornerstone was Mitigating Systems, as determined by Inspection Manual Chapter (IMC) 0609,, Initial Characterization of Findings. Utilizing IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the performance deficiency
26 required a detailed risk evaluation because the degraded condition represented a loss of the PRA function of one train of a multi-train TS system for greater than its TS allowed outage time.
A regional Senior Reactor Analyst (SRA) conducted a detailed risk assessment using the guidance in IMC 0609 Appendix A and the Risk Assessment Standardization Project (RASP)
Handbook. The SRA modelled the condition using the Hatch Unit 1&2 SPAR model version 8.58 dated 2/28/2017 and SAPHIRE 8 Version 8.2.3. The exposure period set to 1 month. Because the recirculation pump discharge isolation valve is not modeled in the Hatch SPAR model, the SRA modelled the condition by setting LCS-CKV-CC-LOOPB: LOOP B INJECTION VALVE F015B FAILS TO OPEN to true for Medium and Large Break Loss of Coolant Accident sequences only. This simulates the LPCI flow being diverted from the core by a pipe break in the B reactor coolant loop, which is the Updated Final Safety Analysis Report (UFSAR) describe safety function of the valve. The dominant accident sequence was a Large Break Loss of Coolant Accident with a failure of Low-Pressure Core Spray and the A loop of Low-Pressure Injection. The risk was less than 1 x E-7, which make this a finding of very low safety significance (Green).
Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. The finding was assigned a cross cutting aspect of Conservative Bias [H.14] in the human performance area, because the licensee did not fully understand the cause of the lights going out on the recirculation pump discharge valve and chose to accept that the valve was operable based on continuous light indication.
Enforcement:
Violation: Hatch Unit 1 Technical Specification 5.4.1.(a) Procedures, states, in part, that written procedures shall be implemented covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, February 1978. Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Appendix A, Paragraph 9.a, Procedures for Performing Maintenance, requires that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances.
Hatch Unit 1 Technical Specification (TS) states in section 3.5.1, Emergency Core Cooling Systems (ECCS), RPV water Inventory Control, and Reactor Core Isolation cooling (RCIC)
System, requires each ECCS injection/spray subsystem shall be operable in Modes 1 through 3; if a low-pressure emergency core cooling system (ECCS) injection/spray subsystem is inoperable, it shall be returned to operable status within seven days or in accordance with TS 3.5.1.B, the unit shall be shut down and within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> be in Mode 3 (hot shutdown).
Contrary to the above, on July 12, 2021, the licensee failed to accomplish activities affecting quality due to inadequate maintenance procedure, for performing post maintenance test instructions for fuse installation, in that the procedure did not contain adequate information to ensure fuse tightness and an adequate PMT to ensure that the discharge valve operated properly. This resulted in the inoperability of B loop of LPCI for approximately 30 days, which is greater than the TS allowed outage time, and the unit had not been placed in Mode 3 within the required time.
27 Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS The inspectors verified no proprietary information was retained or documented in this report.
On April 11, 2022, the inspectors presented the integrated inspection results to Sonny Dean and other members of the licensee staff.
On February 11, 2022, the inspectors presented the Radiation Protection Baseline inspection results to Sonny Dean and other members of the licensee staff.
On February 18, 2022, the inspectors presented the ISI inspection results to Sonny Dean and other members of the licensee staff.
28 DOCUMENTS REVIEWED Inspection Procedure Type Designation Description or Title Revision or Date SCNH-13-020 Hatch Probable Maximum Flood Hydraulics - Severe Accident Management (SAM) for Fukushima Near-Term Task Force (NTTF) Recommendation 2.1 Flooding Reevaluation 1.0 Calculations SCNH-13-021 Evaluation of Plant Hatch Local Intense Precipitation - Severe Accident Management (SAM) for Fukushima Near-Term Task Force (NTTF) Recommendation 2.1 Flooding Evaluation 1.0 Miscellaneous HNP-2-FSAR-2 Hatch Nuclear Plant, Final Safety Analysis Reports Updates, Chapter 2 35 34AB-Y22-002-0 Naturally Occurring Phenomena 20.1 34AR-650-206-2 Plant Service Water Valve Pit Sumps Level High 1.3 71111.01 Procedures NMP-OS-017 Severe Weather (Attachments 1, 2, 3, and 8) 01/02/2022 Condition Reports 10786800, 10798410, 10801311, 10805421, 10805697, 10805947, 10830512, 10833387, 10851603, 10857742, 10858345, 10858478, and 10858480 Corrective Action Documents Condition Reports 10815484, 10817589, 10818648, 10828888, 10837447, 10837589, 10837733, 10838299, 10841947, 10846969, 10847946, 10847988, 10847147, 10848752, 10850247, 10851011, 10851149, 10851168, 10853742, 10856243, 10859834 31EO-EOP-011-1 RCA RPV Control (ATWS) 12.1 31EO-EOP-109-1 Alternate Boron Injection 4.8 31EO-EOP-109-1 Alternate Boron Injection 4.8 34SO-C41-003-1 Standby Liquid Control System 12.8 34SO-C41-003-1 Standby Liquid Control System 12.8 34SO-E11-010-1 Residual Heat Removal System 46.1 34SO-E21-001-1 Core Spray System 24.9 34SO-E51-001-2 Reactor Core Isolation Cooling (RCIC) System 28.2 Procedures 34SO-G71-001-1 Decay Heat Removal System 19.1 71111.04 Shipping Purchase Order SQUIBB Valves 04/11/2019
29 Inspection Procedure Type Designation Description or Title Revision or Date 10202067 Purchase Order 10246495 SQIBB Valves 06/10/2021 Records Purchase Order SNG10202087 SQUIBB Valves 04/17/2020 SNC1072756, 1171614, 930993 Work Orders SNC1226497 Corrective Action Documents Condition Report 10850153 01/10/2022 NMP-ES-035-019-GL02-F12 U1 Reactor Building (Torus Area 87 foot and 114-foot Elevations) 1.0 NMP-ES-035-019-Gl02-F13 U1 Reactor Building ELVE 130 1.0 NMP-ES-035-019-GL02-F21 Unit 1 Radwaste Building 156' elevation 1.0 NMP-ES-035-019-GL02-F29 Unit 2 Reactor Building below 130 Foot Level 1.0 71111.05 Fire Plans NMP-ES-035-019-GL02-F62 CAS Security Building 1.0 Calculations BH2-M-387 Mechanical Calculations 07/14/1983 Corrective Action Documents Condition Report CR10783738 Hatch Final Safety Analysis Report, Sections 2.4 and 9.3 35.0 Miscellaneous Edwin I. Hatch Individual Plant Examinations 12/01/1992 34AB-T22-003-1 Secondary Containment Control 5.18 34AB-T22-003-2 Secondary Containment Control 4.4 34SO-G11-013-2 Drywell and Reactor Building Sumps Systems 2.1 34SO-G11-018-2 Radwaste Collector System Operating Procedure 9.3 42SV-EEL-002-0 Surveillance of Seals Used For The Water Protection of Electrical Equipment 2.5 71111.06 Procedures 57IT-T45-002-2 Sump Actuation Valve Actuation Test 0.3 71111.07A Corrective Action Documents Condition Report 10858927
30 Inspection Procedure Type Designation Description or Title Revision or Date Procedures NMP-ES-012-001 Inspection of Heat Exchangers 2.1 Work Orders SNC1078268 1E11B001B - RHR Heat Exchanger 02/22/2022 H-LT-SG-50322 Shutdown Cooling (SDC) Operations / Loss of SDC / SOER 09-1 30.0 Miscellaneous H-LT-SG-50472 Security Threat with LOSP or Loss of Ultimate Heat Sink 15.0 71111.11Q Procedures 34GO-OPS-013-1 Normal Plant Shutdown 32.6 Cause Determination Reports CAR 279694, 279530, 279192, 279720, 277075 Condition Reports CR10837941,10846119, 10835589, 10837855, 10846729, 1042371101 Corrective Action Documents Condition Reports CR10859981, CR10859984, and CR10863023 Corrective Action Documents Resulting from Inspection Condition Reports CR 10855728 Drawings H-16071 Unit No. 1 Turbine Building Main Steam Leak Detection System Instrument Locations 2.0 Engineering Changes SNC1100336, Design Change Package Unit 1 Main Steam Leak Detection Temperature Switch Replacement Design Change 02/03/2022 Engineering Evaluations Technical Evaluation TE1093338, 1093768 3002000922 Generation Maintenance Applications Center Conventional Vertical Pump Maintenance Guide H-ME-JIT-PSWALIGN-PSW Pump Align JITT 3.0 Miscellaneous SNC1112006 Qualification Report 71111.12 Procedures 34SV-P41-001-2, 16.3 Plant Service Water Pump Operability 16.3
31 Inspection Procedure Type Designation Description or Title Revision or Date 52PM-E11-005-1 Plant Service Water Pump & Motor Major Inspection/Overhaul 9.9 52PM-E11-005-2 Plant Service Water Pump & Motor Major Inspection/Overhaul 7.10 52PM-P41-036-1/2 RHRSW Pump and Motor Maintenance 10.0/9.9 NMP-ES-007 Code of Engineering 16.0 NMP-ES-086, Vendor Technical Information Program 4.0 NMP-MA-037-001 Shaft Alignment - Rim and Face Method 1.1 SCM-004 Procurement of Materials and Services 15.2 SCM-005 Warehouse Operations 41.2 Shipping Records Purchase Order SNG10274777 Thermocouple 01/06/2022 Work Orders Work Orders SNC1112006,105936,1063991,1160313,1112006,1178903, 556025, 312584,109274, 428986,1126953,1120503 Corrective Action Documents Resulting from Inspection Condition Reports 10855685 Miscellaneous Hatch Unit 2 R8 On-Line Configuration Risk Monitor Current Risk Summary Report 02/02/2022 34GO-OPS-007-0 Shutdown/Refueling Mode Preparation Strategy for Flex 1.2 34GO-OPS-024-0 Outage Safety Assessment Forms (for the Week of February 13, 2022) 5.5 NMP-GM-031 On-Line Configuration Risk Management Program 9.0 NMP-GM-038-002 Hatch Diverse and Flexible Coping Strategies (FLEX)
Program Document 4.1 71111.13 Procedures NMP-OS-010-002 Hatch Protected Equipment Logs 11.1 BH1-PD-3675 Evaluation of TAP Pipe Supports for Reduced Allowable (Sheets 182 and 386) 16.0 SCNH-22-001 RHRSW Pumps with Degraded Seismic Restraint 1.0 71111.15 Calculations SMH-97-002 Pipe Stress Analysis of HPCI System Piping form Penetration Number X11 to HPCI Turbine Inlet (Problem 2M) 3.0
32 Inspection Procedure Type Designation Description or Title Revision or Date Condition Report 10858662 Corrective Action Documents Condition Report 10850131 01/10/2022 H-13412 Elementary Diagram Diesel Generator 1A 53.0 H-16865 HPCI, & RHR System Steam to HPCI and RHR HT. EXCH. 'B' 12.0 H-17763 Residual Heat Removal System E11 Elementary Diagram Sheet 4 36.0 Drawings H-53343 4160V Switchgear Bus 1E Degraded Voltage Relay Panel 1H21-P080 1.0 Engineering Evaluations Technical Evaluation 767733 02/03/2014 Operability Evaluations NMP-ES-050-F01 RER Response Form for 1D RHRSW Pump Seismic Evaluation without Seismic Restraint To Support Operability 01/12/2022 Procedures NMP-AD-012 Operability Determinations 15.0 Drawings H-17866 Unit 1 Reactor Recirc Pump and ASD "A" System B31 System Elementary Diagram 33.0 SNC1175128 DECP to Add a UPS for Backup Power for Unit 1 RFPTs TMR Panels 0.0 Engineering Changes SNC1196615 Temporary Change Configuration - Bypass Unit 1 Recirculation Pump 1A ASD Runback #1 on 01/03/2022 Miscellaneous Changes from EPG/SAG Rev. 3 to EPG/SAG Rev. 4 Volume II (EPG-Hot) and Volume V (SAG-Modes 1-4) 30AC-OPS-006-0 Verification Program For Emergency Operating Procedures (Individual Verification Forms for all EOPs Revision 4) 6.4 31EO-EOP-010-1 RC RPV Control for Hot Conditions 13.0 31EO-EOP-012-1 PC Primary Containment Control for Hot Conditions 8.0 31EO-EOP-014-1 SC - Secondary Containment Control for Hot Conditions RR-Radioactivity Release Control Modes 1-3 Hot Conditions 14.0 31EO-EOP-015-1 CP-1 11.0 31EO-EOP-016-1 CP-2 RPV Flooding 11.0 71111.18 Procedures 31EO-EOP-017-CP-3 ATWS RPV Control 14.0
33 Inspection Procedure Type Designation Description or Title Revision or Date 1
31EO-EOP-020-1 RC RPV Control for Cold Conditions (Mode 4) 1.0 31EO-EOP-021-1 Containment and Radioactivity Release Control - Cold Shutdown 1.0 31EO-EOP-030-1 Decay Heat Removal - Refuel (Mode 5) 1.0 31EO-EOP-031-1 Containment and Radioactivity Release Control - Refuel 1.0 31EO-SAG-001-1 SAG-1, Reactor Vessel and Primary Containment Water Inventory Control for Modes 1-4 4.0 31EO-SAG-002-1 SAG-2, Reactor Vessel and Primary Containment Water Inventory Control for Modes 1-4 8.1 31EO-SAG-003-1 SAG - 3 Reactor Vessel And Primary Containment Water Inventory Control for Refuel 1.1 34AB-B21-002-1 RPV Water Level Corrections 5.16 34ABT22-001-1 Loss of Secondary Containment Integrity 0.7 PSTG Hatch Specific Technical Guideline 21.1 SNC1213404 Testing of the UPS 1N21S900 and 1N21S901 02/25/2022 SNC1218283 Work Orders SNC1218287 MRP-3523-J Qualification Report for AVR Assembly 07/19/2011 Miscellaneous NMP-ES-024-301 Liquid Penetrant Examination Record for work order SNC1218290 02/02/2022 34SV-E41-001-2 HPCI Valve Operability 13.7 34SV-P41-003-2 Standby Diesel Service Water System Operability 8.2 34SV-R43-006-2 Diesel Generator 2C Fast Start Test 18.3 52CM-MEL-005-0 Fuse Replacement 10.2 57CP-C51-012-1 LPRM Detector I/V Curve 20.3 57GM-MIC-016-0 LPRM Cable And Connectors 2.2 57IT-C51-001-1 Neutron Monitoring System Detector Pre-installation Tests 5.5 71111.19 Procedures 57SV-C51-014-1 SRM FT&C 6.4
34 Inspection Procedure Type Designation Description or Title Revision or Date SNC 1222334 SNC 1268302 Time Out Called During 2C EDG Post Maintenance Run 03/26/2022 SNC 1269119 EDG 2C Automatic Voltage Regulator Board Replacement 03/27/2022 SNC 1269119-36 Diesel Generator 2C AVR Bump Testing / Tuning 03/28/2022 SNC 1269119-60 Diesel Generator 2C AVR Testing / Tuning 03/30/2022 SNC 1273268 Fuses Needed to be Replaced in 2C EDG Control Panel 03/30/2022 SNC1110822 Steam Leak Repair of Unit 2 HPCI Steam Supply Valve 2E41-F001 02/03/2022 SNC1153248 SNC1154781 1C51N001D - SRM Detector 20-17 RE (SRM D)
SNC1154781 New Cable Pull For SRM E 02/13/2022 SNC1196619 1B31-F031A Temporary Change Configuration Change Removal 02/23/2022 SNC1196870 Investigate dual indication on valve 1B31F031A 02/23/2022 SNC1218290 Replace/weld 1/4-inch line tap fitting 02/02/2022 SNC1225507 4160V Bus 1E de-energized during 1A EDG Start Logic LSFT, replace relay 1S32K817-3 (27-10) 02/11/2022 SNC1227205 Perform point to point verification.
2/12/2022 SNC1228017 Relocate wiring in panel 1H21P080 02/13/2022 Work Orders SNC1237448 1B31AF031A Pump Discharge Valve 02/23/2022 GE: Hatch Throttle Pressure Setpoint Issue After Pressure Transmitter Failed 12/30/2021 Reactor Pressure / Reactor Power / CV Position Graph (12/29/21 Unit 1 Scram) 12/30/2021 H1R30 Shutdown Safety Assessment 0.0 Fault Tree for Unit 1 12/29/21 Scram Reactor Pressure Fluctuations o 12/29/21 Resulting in Unit 1 Manual SCRAM 12/30/2021 HAT1ILRT.22-L220227A Hatch Unit 1 2022 Containment ILRT Preliminary Test Results 02/27/2022 Miscellaneous NMP-RE-008-F01 Detailed Reactivity Management Plan 02/03/2022 71111.20 Procedures 31GO-OPS-010-Scram/Transient Analysis 7.6
35 Inspection Procedure Type Designation Description or Title Revision or Date 0
34AB-E11-001-1 Loss of Shutdown Cooling 3.19 34GO-OPS-003-1 Startup System Status Checklist (from Unit1 1R30 Outage) 02/26/2022 34GO-OPS-013-1 Normal Plant Shutdown (Attachment 1 -
Cooldown/Depressurization Check [OPS-0175])
32.6 34GO-OPS-013-1 Normal Plant Shutdown 32.4 34GO-OPS-015-1 Maintaining Cold Shutdown or Refuel Conditions (Attachment 1 - Monitoring Cold Shutdown and Refuel Parameters
[OPS_1085])
15.2 34SO-N30-001-1 Main Turbine Operation (Section 7.5 - Actions for a failed Throttle Pressure Transmitter) 34.0 NMP-AD-002-F04 Troubling Shooting Log (ASD A Speed Limiter #1) 01/03/2022 NMP-OM-002 Shutdown Risk Management 6.0 NMP-OS-028 Adverse Condition Monitoring Program (Unit 1 Startup from outage - January 2022) 01/01/2022 NMP-RE-007 H1C31 Core Loading Verification 02/23/2022 NMP-RE-008-F01 Detailed Reactivity Management Plan (H1C30-123121-051) 2.1 Condition Report 10850376 01/11/2022 Condition Report 10857409 Condition Report 10855367 Condition Report 10870584 Corrective Action Documents Condition Report 10847900 Condition Report 10850404 01/11/2022 Condition Report 10855181 Corrective Action Documents Resulting from Inspection Condition Report 10873269 Miscellaneous 2A EDG Risk and Trending Assessment and Deficiency Map 12/15/2022 34SV-B21-012-1 Low Low set LSFT 1.2 34SV-E51-002-2 Reactor Core Isolation Cooling Pump Operability IST 27.1 71111.22 Procedures 34SV-R43-001-2 Diesel Generator 2A Monthly Test 30.1
36 Inspection Procedure Type Designation Description or Title Revision or Date 34SV-R43-020-1 Diesel Generator 1A LOCA/LOSP Logic System Functional Test 2.5 34SV-R43-023-1 Diesel Generator 1A LSFT 2.3 34SV-R43-024-1 Diesel Generator 1B LSFT (From U1) 3.1 42SV-TET-001-1 Summarized LLRT Test Results 1R30 03/02/2022 57SV-SUV-004-1 EFCV Operability 16.10 Work Orders SNC 1138590 FLEX PM: 60KW Generator Full Load Test 03/30/2022 Emergency Preparedness Observation Form - LOCT DEP Observation 03/30/2022 71114.06 Miscellaneous H-LT-AF-00114 CPE Design - Loss of CRD, RFPT Controller Malfunction, Loss of Iso-Phase Bus Cooling, Loss of the 2A Drywell Chiller, Torus Leak (Isolation), Loss of Feed Water Heating, HPCI Steam Leak, Emergency Depressurization.
1.0 Corrective Action Documents CRs 10688560,
- 10742297, 10774411,
- 10802840, 10709626, 10757638, and 10792319.
CRs 10688560, 10742297, 10774411, 10802840, 10709626, 10757638, and 10792319.
Various 34SO-G11-023-2 Radwaste Phase Separator Operating Procedure Rev. 15.4 34SO-G11-024-2 Radwaste Spent Resin Tank Operating Procedure Rev. 8.1 34SO-G11-037-1 Processing Sludges & Spent Resins Rev. 8.4 62-RP-RAD-050-0 Waste Separation and Temporary Storage Facility and Sea land Storage Facility Rev. 4.0 62RP-RAD-049-0 Cask Handling for NRC Approved Type A 14-190 and/or 14-210 Casks Rev. 1.1 8-120B Cask Book CASK BOOK FOR MODEL 8-120B USA/9168/B(U)-96 Rev. 48 NMP-HP-400 Control and Accountability of Radioactive Sources Rev. 3.15 NMP-HP-405 Shipment of Radioactive Waste and Radioactive Material Rev. 4.0 NMP-HP-406 Performing Surveys for Shipments of Radioactive Containers Rev. 2.2 71124.08 Procedures NMP-HP-407 Radioactive Materials - Additional Transportation Controls for Category 1 and 2 Quantities of Radioactive Materials Rev. 3.2
37 Inspection Procedure Type Designation Description or Title Revision or Date NMP-HP-408 Solid Radioactive Waste Scaling Factor Determination and Implementation and Waste Classification Rev. 2.0 NMP-HP-415 Storage of Radwaste in Outdoor Process Shields Rev. 2.1 TR-TP-008 Pre-Shipment Leak Test for Model 8-120B Cask NRC Certificate of Compliance 9168 Rev. 24
- 10688560, 10742297,
- 10774411, 10802840,
- 10709626, 10757638, and 10792319 Corrective Action Reports Various Corrective Action Documents 10855849 02/02/2022 64CH-SAM-028-0 Releases Via Planned and Unplanned Routes: Sampling and Analysis Version 11.2 Procedures HNP ODCM Offsite Dose Calculation Manual Revision 26 Self-Assessments CAR #279683 Area for Improvement 08/05/2021 71151 Work Orders Work Order
- SNC1168290 1R25S020-TCCI-Instrallation of temporary power to the 1R25-S020 panel 07/16/2021 Condition Report 10855849 Condition Report 10840695 Corrective Action Report 287331 Corrective Action Report 279686 Corrective Action Documents Corrective Action Review 279683 H-46546 Tritium Monitoring Facility Relocation Plan 1.0 H-46547 Tritium Monitoring Facility Relocation Underground Storage Tank and MISC. Supports Plan 1.0 Drawings H-46556 Tritium Monitoring Facility Piping from Underground Tank Discharge Structures 1.0 71152A Miscellaneous Employee Case 20210825-11
38 Inspection Procedure Type Designation Description or Title Revision or Date Concerns Program Case File NEI 07-07 Industry Groundwater Protection Initiative - Final Guidance Document 1.0 NMP-GM-024-F04 Nuclear Safety Culture Leadership Team Report Quarterly Meeting Minutes 02/02/2021 NMP-GM-024-F04 Nuclear Safety Culture Leadership Team Report Quarterly Meeting Minutes 02/08/2022 NMP-GM-024-F04 Nuclear Safety Culture Leadership Team Report Quarterly Meeting Minutes 10/27/2021 NMP-GM-024-F04 Nuclear Safety Culture Leadership Team Report Quarterly Meeting Minutes 06/07/2021 64CH-SAM-028-0 Releases Via Planned and Unplanned Routes: Sampling and Analysis 11.2 Procedures NMP-ES-086 Vendor Technical Information Program 4.0 Work Orders SNC116890 Installation of Temporary Power to the 1R25-S020 Panel 1.0 Calculations Cause Determination Reports CAR 279694, 279530, 279192, 279720, 277075 Corrective Action Documents Condition Reports CR10837941, 10846119, 10835589, 10837855, 10846729, 1042371101 Condition Report CR10855728 Corrective Action Documents Resulting from Inspection Condition Reports 10855728, 10855765 Engineering Evaluations Technical Evaluation TE1093338, 1093768 3002000922 Generation Maintenance Applications Center Conventional Vertical Pump Maintenance Guide H-ME-JIT-PSWALIGN-PSW Pump Align JITT 3.0 71152S Miscellaneous SNC1112006 Qualification Report
39 Inspection Procedure Type Designation Description or Title Revision or Date 34SV-P41-001-2 Plant Service Water Pump Operability 16.3 52PM-E11-005-1/2 Plant Service Water Pump & Motor Major Inspection/Overhaul 9.9/7.10 52PM-P41-036-1/2 RHRSW Pump and Motor Maintenance 10.0/9.9 NMP-ES-007 Code of Engineering 16.0 Procedures NMP-MA-037-001 Shaft Alignment - Rim and Face Method 1.1 Work Orders Work Orders SNC1112006, 105936, 1063991, 1160313, 1112006, 1178903, 556025, 312584, 109274, 428986, 1126953, 1120503 Cause Determination Report (CAR) 279617 09/17/2021 Cause Determination Report (CAR) 279607 09/21/2021 Cause Determination Report (CAR) 290990 02/02/2022 Condition Report (CR) 10855849 Condition Report (CR) 10860678 Condition Report (CR) 10817773 08/03/2021 Corrective Action Documents Condition Reports (CR) 10810391, 10810631, 10817701 and 10827732 Technical Evaluation (TE) 1094466, 1093497 09/29/2021 Engineering Evaluations Technical Evaluations (TE) 10993164, 1093165, and 1093166 71153 Miscellaneous License Event Report (LER)
Automatic Reactor Scram on Low Reactor Water Level due to Loss of Reactor Feed Pumps 10/01/2021
40 Inspection Procedure Type Designation Description or Title Revision or Date 2021-001-00 License Event Report (LER) 2021-002-00 Low Pressure Coolant Injection Inoperable Longer than the Allowed Technical Specification Completion Time 10/01/2021 License Event Report (LER) 2022-001-00 Manual Reactor Scram Due to reactor Pressure Perturbations 02/28/2022 IP-ENG-001 Standard Design Process (EB-17-06) 2.0 NMP-AD-002 Conduct of Problem Solving and Troubleshooting 13.6 NMP-ES-005-F01 Single Point Vulnerability Documentation Form (Feeder Breaker for 1R23S021 Bus) 08/26/2021 NMP-ES-044 Preparation of Design Change Packages 13.1 Procedures NMP-ES-095 Interface Procedure for IP-ENG-001, "Standard Design Process" 9.1