IR 05000321/2023002
| ML23216A113 | |
| Person / Time | |
|---|---|
| Site: | Hatch, 07200036  |
| Issue date: | 08/08/2023 |
| From: | Division Reactor Projects II |
| To: | Brown R Southern Nuclear Operating Co |
| References | |
| IR 2023002, IR 2023001 | |
| Download: ML23216A113 (1) | |
Text
SUBJECT:
EDWIN I. HATCH - INTEGRATED INSPECTION REPORT 07200036/2023001, 05000321/2023002, AND 05000366/2023002
Dear R. Keith Brown:
On June 30, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Edwin I. Hatch. On July 27, 2023, the NRC inspectors discussed the results of this inspection with Matt Busch, Plant Manager, and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. One of these findings involved a violation of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the 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.
August 8, 2023 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Alan J. Blamey, Chief Reactor Projects Branch 2 Division of Reactor Projects Docket Nos. 07200036, 05000321, and 05000366 License Nos. DPR-57 and NPF-5
Enclosure:
As stated
Inspection Report
Docket Numbers:
07200036, 05000321, and 05000366
License Numbers:
Report Numbers:
07200036/2023001, 05000321/2023002, and 05000366/2023002
Enterprise Identifier:
I-2023-001-0105 and I-2023-002-0018
Licensee:
Southern Nuclear Operating Company, Inc.
Facility:
Edwin I. Hatch
Location:
Baxley, GA
Inspection Dates:
April 01, 2023 to June 30, 2023
Inspectors:
B. Bowker, Reactor Inspector
P. Cooper, Senior Reactor Inspector
L. Day, Resident Inspector
S. Downey, Sr. Reactor Operations Engineer
R. Easter, Resident Inspector
T. Fanelli, Senior Reactor Inspector
J. Hickey, Senior Project Engineer
J. Hickman, Senior Resident Inspector
M. Magyar, Reactor Inspector
D. Neal, Health Physicist
D. Ray, Reactor Inspector
R. Smith, Senior Resident Inspector
Approved By:
Alan J. Blamey, Chief
Reactor Projects Branch 2
Division of Reactor Projects
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Edwin I. Hatch, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Inadequate Procedure for Condensate Hotwell Level Control Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000366,05000321/2023002-01 Open/Closed
[H.12] - Avoid Complacency 71152A A self-revealed Green finding and associated non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a, Procedures was identified for the licensees failure to maintain appropriate condensate system procedures as a result of inadequate implementation of the design change process. Specifically, the procedure instructions for setting the condenser hotwell level controller, for both units, was revised with setpoints that did not ensure proper suction head to the condensate booster pumps. As a result, unit 2 experienced an automatic power runback due to a condensate booster pump low suction pressure condition.
Incorrect Adjustable Speed Drive High-Speed Setpoints Result in Manual Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000366/2023002-02 Open/Closed
[H.8] -
Procedure Adherence 71153 A self-revealed Green finding was identified when the licensee failed to adequately implement the design change process for a modification to the unit 2 adjustable speed drives (ASD).
Specifically, the licensee failed to adequately control, verify, and translate the high-speed stop setpoints for the reactor recirculation pumps ASDs when implementing a design change per IP-ENG-001, Standard Design Process, during the 2R27 refueling outage. The inadequate modification resulted in a trip of both reactor recirculation pumps and manual rapid shutdown (scram).
Violation of Primary Containment Technical Specification 3.6.1.1 Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000366/2023002-03 Open/Closed Not Applicable 71153 A self-revealed Severity Level IV NCV of TS Limiting Condition of Operation (LCO) 3.6.1.1,
Primary Containment, was identified during refueling outage local leak rate testing of the inboard and outboard drywell ventilation vent valves when both valves exceeded the maximum allowable primary containment leakage limits resulting in a failure of the associated penetration and a loss of containment integrity.
Additional Tracking Items
Type Issue Number Title Report Section Status LER 05000366/2023001-00 Unit 2, Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)71153 Closed LER 05000366/2023-002-00 Unit 2, Manual Scram due to Trip of both Reactor Recirculation Pumps 71153 Closed
PLANT STATUS
Unit 1 began the inspection period at 100 percent rated thermal power (RTP). On May 12, 2023, the unit automatically down powered to 79 percent RTP due to a reactor recirculation pump runback caused by a number three speed limiter from a low condensate booster pump suction pressure. The unit was restored to 100 percent RTP on May 14, 2023. The unit was shut down on May 19, 2023, to perform repairs to the A unit auxiliary transformer (UAT). On May 22, 2023, the unit was restarted and returned 100 percent RTP on May 25, 2023. On June 13, 2023, the unit was down powered to approximately 91 percent RTP due to high thrust bearing temperatures on the B circulating water pump. The licensee determined the cause of the high temperatures on the pump and returned the unit to 100 percent RTP on June 14, 2023.
The unit operated at 100 percent RTP for the remainder of the inspection period.
Unit 2 began the inspection period shut down from a manual scram on March 31, 2023. The scram was required due to a trip of both reactor recirculation pumps. The licensee determined and addressed the cause of the pumps trip and on April 1, 2023, the unit was restarted. On April 5, 2023, the unit reached approximately 100 percent RTP. On May 12, 2023, the unit was down powered to 70 percent RTP to perform a rod control sequence exchange and was returned to 100 percent RTP the same day. The unit operated at 100 percent RTP 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
Seasonal Extreme Weather Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated readiness for seasonal extreme weather conditions prior to the onset of summer weather for the following systems:
Plant service water (PSW) and residual heat removal service water on April 24, 2023
Offsite and alternate alternating-current (AC) power systems on April 25, 2023
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 2 B-division of residual heat removal (RHR) while the A-division was in a system outage, on April 11 and 12, 2023
- (2) Unit 2 containment atmospheric control system following the replacement of the nitrogen storage tank rupture disc, on April 21, 2023
- (3) Unit 1 division 2 of the RHR system following a system outage, on April 26, 2023
- (4) Unit 1 'C' emergency diesel generator (EDG) prior to removing the 'A' EDG from service for a planned outage, on May 31, 2023
- (5) Unit 2 reactor core isolation cooling (RCIC) after a quarterly surveillance test, on June 1, 2023
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (6 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 2 turbine building 130-foot elevation on April 5, 2023
- (2) Unit 1 refueling floor 228-foot elevation on April 5, 2023
- (3) Unit 2 refueling floor 228-foot elevation on April 5 and 6
- (4) Unit 1 radwaste building 108-foot elevation on April 17, 2023
- (5) Unit 2 radwaste building 164/178-foot elevation on April 17, 2023
- (6) Diesel generator building on May 17, 2023
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) The inspectors observed and evaluated licensed operator performance in the main control room during a unit 2 plant cooldown, shutdown cooling operations, power ascension activities including main turbine warming, main turbine roll, main generator synchronizing to the grid, and transfer of steam loads from the turbine bypass valves to the main turbine on April 1 and 2, 2023.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated a crew in the simulator performing scenario, H-LR-AF-00107, "Reactor Feed Pump Turbine Cooling, Safety Relief Valve Open, Loss of Feedwater Heating, Loss of Coolant Accident in the Drywell (DW), DW Spray Failure, RCIC Controller Failure," on April 5, 2023.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (3 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Main control room annunciators overall system health (function-oriented approach)
- (2) Unit 2 wide range drywell pressure/radiation monitoring instruments, Z-99-01, being evaluated for maintenance rule (10 CFR 50.65) a(1) status due to greater than three condition monitoring events in the past 36 months
- (3) Unit 2 condenser hotwell, N21-01, level event resulting in a plant level event (PLE)
(i.e., automatic down power greater than 20 percent RTP) requiring evaluation for transition to maintenance rule a(1) status
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01) (4 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 to the A-division RHR system outage, on April 11, 2023
- (2) Unit 1 elevated risk due to placing the B-division RHR system in shutdown cooling during a maintenance outage to repair the 'A' UAT, on May 21, 2023
- (3) Unit 1 elevated risk due to scheduled 'A' EDG outage, on June 5, 2023
- (4) Unit 2 elevated risk due to scheduled B-division core spray (CS) system outage, on June 13, 2023
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (6 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Condition Reports (CRs) 10960049, 10957452. Unit 2 'A' RHR heat exchanger leak through the 'startup' vent valve on the shell side upper head.
- (2) CR 10963936. Unit 1 safety-related bus, 1R23-S021, transfer from normal to alternate supply and resulting engineering review to adjust under-voltage relay setpoints.
- (3) CR 10964839. Unit 2 'A' diesel fuel oil transfer pump in the required action range during the performance of surveillance 34SV-R43-010-0.
- (4) CR 10968270. Unit 1 RCIC room cooler, 1T41B004A, isolated due to PSW leak.
- (5) CR 10971170. Unit 1 recirculation runback to 60.3 percent speed demand due to a degraded relay.
- (6) Work order (WO) SNC1485487. Steam leak repair of a plug for injection valve 2E41-F006 of the high pressure coolant injection (HPCI) system.
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated unit 1 maintenance outage activities due to an oil leak repair on the 'A' UAT bushings, from May 19 through May 23, 2023.
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (7 Samples)
- (1) Unit 2 A-division of RHR following a system outage, on April 13, 2023
- (2) Unit 2 'B' reactor protection system (RPS) following motor generator set system outage, on April 19, 2023
- (3) Unit 1 'A' RPS following motor generator set system outage, on May 4, 2023
- (4) Diesel fire pump, 1X43-C002A, after engine replacement, on May 12, 2023
- (5) Unit 1 'A' EDG following diesel outage, on June 9, 2023
- (6) Unit 2 B-division of CS following a system outage, on June 14, 2023
- (7) Unit 2 corrective actions and retest of the A-division standby gas treatment system following a failure to develop flow, on June 26, 2023
Surveillance Testing (IP Section 03.01) (5 Samples)
- (1) Unit 1 RCIC pump operability, on April 12, 2023
- (2) Unit 2 'A' and 'B' fuel oil transfer pumps surveillance test, on April 20, 2023
- (3) Unit 1 'C' EDG operability, on April 24, 2023
- (4) Unit 1 and unit 2 reactor manual scram functional test, on May 8, 2023
- (5) Unit 2 HPCI pump operability, on May 18, 2023
Inservice Testing (IST) (IP Section 03.01) (1 Sample)
- (1) Unit 2 'A' CS pump operability, on April 5,
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
BI01: Reactor Coolant System (RCS) Specific Activity Sample (IP Section 02.10)===
- (1) Unit 1 (April 1, 2022, to March 31, 2023)
- (2) Unit 2 (April 1, 2022, to March 31, 2023)
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
- (1) Unit 1 (April 1, 2022, to March 31, 2023)
- (2) Unit 2 (April 1, 2022, to March 31, 2023)
71152A - Annual Follow-up Problem Identification and Resolution Annual Follow-up of Selected Issues (Section 03.03)
The inspectors reviewed the licensees implementation of its corrective action program related to the following issues:
- (1) Corrective action reports (CARs) 393867, 398875, and 391000 associated with the following unit 2 events:
1) main turbine trip on initial synchronization to the grid, 2) erroneous generation indications, 3) recirculation 'B' trip during the plant hydro, 4) failure to start the 'A' and 'B' recirculation pumps initially, and 5) condenser hotwell level setpoints adjusted per design change resulting in a recirculation pump runback.
The inspectors verified corrective actions generated from the above listed causal products have been completed as scheduled.
- (2) CAR 277846, associated with safety system functional failures reporting errors and extent of condition review (i.e., last three years: 2020 to 2023). No additional errors were identified by the inspectors.
71153 - Follow-Up of Events and Notices of Enforcement Discretion Event Follow-up (IP Section 03.01)
- (1) Manual reactor scram of unit 2 following a loss of both recirculation pumps. The inspectors reviewed the licensees initial response, cause determination, and plant recovery activities on April 2, 2023. This completed an inspection sample started in the first quarter of 2023.
- (2) The inspectors evaluated the licensee's response to a unit 1 recirculation pump runback on May 12, 2023.
Event Report (IP Section 03.02) (2 Samples)
The inspectors evaluated the following licensee event reports (LERs):
- (1) LER 05000366/2023-001-00, "Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)" (ADAMS Accession No.
ML23095A088) on February 7, 2023. The inspection conclusions associated with this LER are documented in this report under Inspection Results, Section 71153. This LER is Closed.
- (2) LER 05000366/2023-002-00, "Unit 2 Manual Scram due to Trip of both Reactor Recirculation Pumps" (ADAMS Accession No. ML23146A191) on March 31, 2023. The inspection conclusions associated with this LER are documented in this report under Inspection Results, Section 71153. This LER is Closed.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===60855 - Operation of Independent Spent Fuel Storage Installations Inspections were conducted using the appropriate portions of the 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) 2690, Inspection Program for Storage of Spent Reactor Fuel and Reactor-Related Greater-than-Class C Waste at Independent Spent Fuel Storage Installations (ISFSI) and for 10 CFR Part 71 Transportation Packagings."
Operation Of An ISFSI===
- (1) From May 15 - 19, 2023 the inspectors performed a review of the licensees ISFSI activities to verify compliance with regulatory requirements. During the on-site inspection, the inspectors observed and reviewed licensee activities in each of the five safety focus areas including occupational exposure, public exposure, fuel damage, confinement, and impact to plant operations.
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. Additionally, the inspectors performed independent walkdowns of the heavy load lifting equipment and the ISFSI haul path. The inspector also performed an independent radiation survey of the ISFSI pad.
===71003 - Post-Approval Site Inspection for License Renewal The NRC continued monitoring the licensees performance at Edwin I. Hatch Nuclear Plant by conducting a Post-Approval Site Inspection for License Renewal - Phase 4 in accordance with the license renewal inspection program (LRIP). Per IMC 2516, the LRIP is the process used by NRC staff to verify the adequacy of aging management programs (AMPs) and other activities associated with an applicants request to renew an operating license of a commercial nuclear power plant beyond the initial licensing period under 10 CFR Part 54, Requirements for the Renewal of Operating Licenses for Nuclear Power Plants.
The inspectors reviewed the licensees implementation of the AMPs shown below by selecting a sample of SSCs within the scope of the respective AMPs. The inspectors performed the following activities, as applicable to each SSC, to determine that there is a reasonable assurance that the effects of aging are being adequately managed: walked down all accessible SSCs to observe their general condition and identify any signs of aging related degradation; interviewed plant personnel; reviewed completed work orders; reviewed applicable monitoring and trending data; and reviewed the acceptability of inspection and test results. For each AMP shown below, the inspectors also reviewed a sample of aging related issues entered into the licensees corrective action program to verify that age-related degradation is being identified at an appropriate threshold and corrected.
Post-Approval Site Inspection for License Renewal===
- (1) Condensate storage tank inspections
Unit 1 condensate storage tank (1P11A100)
- (2) Fire protection activities
Fire damper (1X41C029A)
Fire door (1L48D134)
Fire protection strainers (1X43D034, 2U43D003)
Fire sprinkler head (1X43130W13)
- (3) Gas systems inspection activities
10" RCIC pump turbine exhaust line to suppression pool water level
20" HPCI pump turbine exhaust line from turbine case to suppression pool water level
Unit 2 HPCI pump (2E41C001)
Unit 2 RCIC pump (2E51C001)
Unit 2 HPCI pump turbine exhaust drain pot to torus check valve (2E41F040)
Unit 2 RCIC turbine exhaust outboard isolation valve (2E51F040)
Unit 2 RCIC turbine exhaust line test connection valve(2E51F041)
Unit 2 RCIC turbine exhaust inboard isolation valve (2E51F001)
- (4) Insulated cables and connections AMP
Accessible cables and connections in the diesel generator building, unit 1 turbine building, and unit 2 reactor building.
The following inaccessible low voltage cables: PUEB41M02, RIE409M01, and RIE412M01
- (5) Passive component inspection activities
Passive components associated with unit 1 A EDG (1R43S001A)
Passive components associated with unit 2 C EDG (1R43S001C)
Diesel 1B fuel oil day tank (1R43-A001B)
Unit 2 RHR service water pump mini flow lines from valves (F207A, B, C, and D) to common headers termination point in intake structure
Unit 2 turbine building exhaust fans (2U41C002A, 2U41C002B) ductwork flex connectors and gaskets
Vacuum pump discharge check valve (2E51-F028)
- (6) RHR heat exchanger augmented inspection and testing program
Unit 1 RHR heat exchangers (1E11B001A, 1E11B001B)
Unit 2 RHR heat exchangers (2E11B001A, 2E11B001B)
- (7) Structural monitoring program
EDG building
Intake structure
Off-gas stack
- (8) Torus submerged components inspection T-quencher assemblies (1B21-FO13G, 1B21-F013H, 2B21-FO13G, and 2B21-F013K)
RCIC system turbine exhaust piping and diffuser assemblies (1T23-X212 and 2T23-X212)
Torus to drywell vacuum breaker drain lines (1T23-F323E and 1T23-F323F)
INSPECTION RESULTS
Inadequate Procedure for Condensate Hotwell Level Control Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green NCV 05000366,05000321/2023002-01 Open/Closed
[H.12] - Avoid Complacency 71152A A self-revealed Green finding and associated non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a, Procedures was identified for the licensees failure to maintain appropriate condensate system procedures as a result of inadequate implementation of the design change process. Specifically, the procedure instructions for setting the condenser hotwell level controller, for both units, was revised with setpoints that did not ensure proper suction head to the condensate booster pumps. As a result, unit 2 experienced an automatic power runback due to a condensate booster pump low suction pressure condition.
Description:
On March 11, 2023, with unit 2 operating at 100 percent RTP, the main control room operators received a high-water level alarm for the main condenser hotwell. During review of the alarm response, 34AR-650-222-2, Main Condenser Hotwell Level Abnormal, version 1.5 and system operating procedure, 34SO-N21-007-2, Condensate & Feedwater System, version 57.2, the reactor operators noted that the hotwell level controller setpoint (for controller B) was higher than the setpoint specified in the operating procedure.
Operators then adjusted (i.e., lowered) the controller setpoint to match the procedure. Shortly thereafter, unit 2 received a condensate booster pump suction low alarm that was immediately followed by a recirculation pump runback to 79 percent RTP. Operators stabilized the plant at this power level and an investigation was initiated to determine the cause of the event.
The licensee determined that the hotwell level setpoint in the operating procedure was incorrect and resulted in vortexing of the condensate pumps which in turn caused the condensate booster pump low suction pressure runback. The hotwell level controller setpoints in the operating procedure had been revised during the previous refueling outage (2R27, January to February 2023), under design change package (DCP) SNC497427, to match the unit 1 procedure setpoints that had been previously revised under another DCP in 2013 (DCP SNC466230). A review of the 2013 DCP identified that the modification added, in part, hotwell level controller setpoints that were incorrect. Specifically, A and B level controller setpoints were added without any calculation or bases documents that supported the adequacy of the level setpoints. These latent errors were not recognized by the licensee when revising the unit 2 procedure because the change was assessed as low risk and minimal scope given the approved status of the unit 1 DCP. As a result, the licensee fast-tracked the unit 2 DCP and curtailed design calculations/analyses, impact reviews, and stakeholder engagement normally required by the design change process to ensure adequate development and implementation of the change. As required by the standard design process procedure (IP-ENG-001, Standard Design Process, revision 3), fast track modifications require additional risk mitigation actions (Attachment 9, section 3) to compensate for the reduced level of design planning, calculations, reviews, and approvals.
The licensees causal analysis determined, in part, that not using some form of additional risk mitigating actions to drive stakeholder engagement and interfaces to determine design change impacts was the primary contributor to incorrectly changing the unit 2 procedure.
These barriers could have ensured proper transition between design and implementation of the change.
Corrective Actions: The licensee revised both unit 1 and unit 2 condensate and feedwater system operating procedures with correct setpoint variables and revised the instructions for adjusting the setpoints. Additionally, a procedure step was added to require operations management approval prior to adjusting hotwell level setpoints. The licensee also revised the conduct of operations procedure (NMP-OS-007 Conduct of Operations) to update the attachment titled SRO Decision Making Assist Tool, to require more stakeholder engagement actions when making a decision that could affect the plant operations.
Corrective Action References: Condition reports (CR) 10955449, 10955883; technical evaluations (TE) 1125949, 1125042, and 1124304; and corrective action report (CAR)398875.
Performance Assessment:
Performance Deficiency: The failure to develop and implement a procedure modification for condenser hotwell level controller setpoints in accordance with IP-ENG-001, Standard Design Process, that ensured proper suction head to the condensate booster pumps 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 Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in a condensate booster pump low suction pressure condition that led to a reactor recirculation pump runback to the number three speed limiter and plant down power.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Utilizing the initiating events screening questions from Exhibit 1 of IMC 0609, Appendix A, the inspectors determined that the finding was of very low safety significance (i.e., Green).
Specifically, the finding did not result in a reactor trip, nor the loss of mitigation equipment relied upon to place the reactor a stable condition.
Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. The licensee failed to properly maintain a cautious attitude when implementing the design change and failed to ensure the correct values for hotwell level control were incorporated in the updated procedure revision prior to the unit startup.
Enforcement:
Violation: Technical Specification 5.4.1.a, Procedures, requires, in part, that written procedures shall be maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, of February 1978. Appendix A, Item 4.n, requires procedures for operation of the condensate system. Procedures 34SO-N21-007-1/2 (unit 1/2), Condensate & Feedwater System, Versions 52.10/57.2, provides instructions for operating the hotwell level controllers to, in part, provide adequate suction head for the condensate pumps.
Contrary to the above, between approximately 2013 and March 2023, the licensee failed to adequately maintain the units 1 and 2 condensate system operating procedures. Specifically, the unit 1 procedure was inadequately revised in 2013 and more recently, inadequate design control measures failed to recognize the unit 1 latent procedure errors when duplicating the 2013 changes into the unit 2 procedure. As a result, unit 2 experienced an automatic runback to approximately 79 percent RTP when a hotwell low level condition caused a low suction pressure on the condensate booster pump.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Incorrect Adjustable Speed Drive High-Speed Setpoints Result in Manual Scram Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000366/2023002-02 Open/Closed
[H.8] -
Procedure Adherence 71153 A self-revealed Green finding was identified when the licensee failed to adequately implement the design change process for a modification to the unit 2 adjustable speed drives (ASD).
Specifically, the licensee failed to adequately control, verify, and translate the high-speed stop (HSS) setpoints for the reactor recirculation pumps ASDs when implementing a design change per IP-ENG-001, Standard Design Process, during the 2R27 refueling outage. The inadequate modification resulted in a trip of both reactor recirculation pumps and manual rapid shutdown (scram).
Description:
On March 31, 2023, with unit 2 operating at 97 percent RTP, the A pump ASD tripped followed by the B pump ASD tripping 14 seconds later. As required by operating procedures, reactor operators initiated a manual scram after losing reactor recirculating pumps.
The ASD units control the speed of the reactor recirculating pumps. Maximum recirculation pump speed is limited by the ASD controller via the HSS setpoint so as not to exceed the maximum allowable core flow. Over-frequency relays monitor ASD output frequency (i.e.,
pump speed) and will trip the ASD as a backup to the ASD controller maximum speed limiter.
The ASD HSS setpoint should always be set lower than the over-frequency relays setpoint such that the ASD speed is capped and will not exceed the over-frequency relay trip setpoints unless there is a malfunction in the ASD control. The licensee determined that both the 'A' and 'B' ASDs tripped due to actuation of the Basler (over-frequency) relays. A review of the ASD parameter files shows that the 'A' HSS setpoint was 96.4 percent flow and the 'A' Basler relay setpoint was 94.75 percent flow, while the 'B' HSS setpoint was 96.4 percent flow and the 'B Basler relay setpoint was 94.92 percent flow. When tripped, the 'A' ASD speed was 94.8 percent flow, and the 'B' ASD speed was 94.9 percent. This confirmed that the HSS setpoints were incorrectly set above the Basler relay setpoints and that the trip was due to the Basler relays being actuated.
The cause of HSS not being properly set was due to the inadequate implementation of the design change process for a modification to upgrade the code and programmable logic controller of the ASD, DCP SNC1219886, ASD Modification During 2R27, during the 2R27 refueling outage. The design change process was described in IP-ENG-001, Standard Design Process (SDP), revision 3.0, and defined a critical characteristic as those properties or attributes that are essential for the items form, fit, and functional performance.
Critical characteristics for design are the identifiable and/or measurable attributes of a replacement item or SSCs that provide assurance that the replacement item or SSC will perform its design function. Attachment 5, Design Equivalent Change Package, of the procedure required implementation of design steps to
- (1) document design inputs, requirements, and critical characteristics that are applicable to the design or have changed (step 3.3.4.B.4.a),
- (2) ensure that all the critical characteristics/requirements of the change package are properly evaluated (step 3.3.4.a), and
- (3) using the bounding technical requirements and design inputs, the responsible engineer and design team shall specify the design requirements/critical characteristics that need to be tested, the acceptance criteria, and any specific test methods in the change package (step 3.3.7). The licensee did not follow 5, step 3 of the procedure, and therefore failed to adequately control/verify/
translate critical characteristics in the design phase of the modification. The ASD HSS setpoints were a critical characteristic and the lack of design control resulted in HSS values being incorrectly specified from 2015 unit 1 parameters. Additionally, there was inadequate vendor oversight in the development of the factory acceptance test as well as inadequate technical rigor in the site functional test, which required control of critical characteristics.
Because the HSS setpoints were not adequately controlled the licensee failed to translate them into testing requirements in the design phase.
Corrective Actions: The licensee
- (1) corrected the HSS setpoints to the previous cycle values, which are lower than the Basler relay setpoints,
- (2) issued a standing order to provide operating limits on recirculating pump speed,
- (3) audited all modifications prior to implementation to ensure that critical characteristics are appropriately tested in the functional test procedure, and
- (4) conducted an extent of condition to verify unit 1 did not have the same condition.
Corrective Action References: CR 10961416 and CAR 415509
Performance Assessment:
Performance Deficiency: The inspectors determined that the licensees failure to implement the design change process in accordance with procedure IP-ENG-001, Standard Design Process when establishing control of critical characteristics for a modification to the unit 2 ASD was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human 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.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Utilizing the initiating events screening questions from Exhibit 1 of IMC 0609, Appendix A, the inspectors determined that the finding was of very low safety significance (i.e., Green).
Specifically, the finding did not result in the loss of mitigation equipment relied upon to place the reactor a stable condition.
Cross-Cutting Aspect: H.8 - Procedure Adherence: Individuals follow processes, procedures, and work instructions. The licensee failed to properly implement a design change due to the failure to follow procedural required actions that ensured critical design characteristics were properly incorporated into the recirculation pump ASDs design change that resulted in both recirculation pumps tripping and subsequent required a manual scram.
Enforcement:
Inspectors did not identify a violation of regulatory requirements associated with this finding.
Violation of Primary Containment Technical Specification 3.6.1.1 Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000366/2023002-03 Open/Closed Not Applicable 71153 A self-revealed Severity Level (SL) IV NCV of TS Limiting Condition of Operation (LCO)3.6.1.1, Primary Containment, was identified during refueling outage local leak rate testing (LLRT) of the inboard and outboard drywell ventilation vent valves when both valves exceeded the maximum allowable primary containment leakage limits resulting in a failure of the associated penetration and a loss of containment integrity.
Description:
On February 7, 2023, during refueling outage 2R27, Hatch unit 2 primary containment isolation valves 2T48-F319 and 2T48-F320 (referred to as valves F319 and F320 from here on out) exceeded both the administrative limit and the maximum allowable leakage (La), resulting in a failure to maintain primary containment integrity through the associated penetration, 2T23-X026. These valves are the inboard (F319) and outboard (F320) 18-inch drywell ventilation vent valves.The licensee submitted LER 05000366/2023-001-00, "Primary Containment Penetration Exceeded Maximum Allowable Primary Containment Leakage Rate (La)" (ADAMS Accession No. ML23095A088) on February 7, 2023, for this event.
The failures were caused by defective T-rings in the replacement valves provided by the original equipment manufacturer that were installed in F319 and F320 locations in 2020 and 2021, respectively. The T-ring design had been changed in that the seating area was rotated 90 degrees from the proper orientation during the manufacturing process. This caused the T-rings to be more susceptible to relaxation (i.e., creep, cold flow, compression set) due to the additional mechanical stress and to pull away from the disc to cause seal failure. Despite the deficient T-ring, both valves passed the as-left LLRT after replacement in 2020 and 2021.
Following unit 2 startup from a maintenance outage on October 17, 2022, drywell pressure exhibited a declining trend. Troubleshooting identified and repaired the source of the leak.
The drywell ventilation vent valves (F319 and F320) were inspected and ruled out as the source of the leakage during troubleshooting. Based on the troubleshooting and effectiveness of repairs (i.e., resulting drywell pressure rise) the licensee determined that the defective T-rings did not degrade until the vent valves were opened to de-inert the primary containment, which began on January 28, 2023, to support the unit 2 refueling outage. Specifically, the T-rings relaxed back to the manufactured state without the resistance of the valve disc when the valves were opened during the prolonged de-inerting period.
Corrective Actions: The licensee replaced the defective T-rings and satisfactorily tested (LLRT) both valves. Additionally, the licensee modified the valves by re-orienting them in the flow path to allow primary containment pressure to be felt on the butterfly disk and assist in seating the valves against the T-rings for a better seal.
Corrective Action References: CRs 10946508, 10945741, 10946502, 10956992, and 10958712. CAR 380222
Performance Assessment:
The NRC determined this violation was not reasonably foreseeable and preventable by the licensee and therefore is not a performance deficiency.
Specifically, the installation of containment isolation valves with defective T-rings was outside the licensees ability to foresee and correct.
Enforcement:
Traditional Enforcement is being used to disposition this violation with no associated Reactor Oversight Process performance deficiency per section 3.10 of the Enforcement Manual.
Severity: The inspectors assessed the severity of the violation using Section 6.1 of the Enforcement Policy and determined the significance is characterized as SL IV, due to the low potential safety consequences. Leakage through the valves was determined to be much less than the > than 100 percent containment volume per day criteria from Inspection Manual Chapter (IMC) 0609 Appendix H, Containment Integrity Significance Determination Process, Table 7.2, Phase 2 Risk Significance -Type B Findings at-Power. Specifically, as documented in NRC inspection report 05000321/2020010 and 05000366/2020010 (ADAMS Accession No. ML20357B073), Hatch unit 2 experienced (during cycle 26) primary containment leakage through these valves from October 22, 2019 until January 4, 2020, when the licensee realized the valves were the source of leakage. Due to this leakage, nitrogen was continuously added to the unit 2 drywell to maintain pressure during cycle 26. A third-party vendor calculated the leakage rate during this period to be approximately 89% of containment volume per day, resulting in a very low safety significant finding (i.e., Green)based on the Table 7.2 criteria in IMC 0609 Appendix H. During cycle 27, unit 2 did not experience the continuous drywell pressure loss seen in cycle 26, requiring continuous nitrogen makeup. Instead, unit 2 primary containment was regularly vented throughout the cycle, consistent with the normal practice. Based on review of the drywell pressure trends (including cycle 27 venting) versus cycle 26 nitrogen additions, it can be reasonably concluded that leakage through these valves during cycle 27 was considerably less than cycle 26 and therefore below the 100 percent containment volume per day criteria in IMC 0609 App H. Additionally, the exposure time was determined to be approximately 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br />.
Violation: Hatch Nuclear Plant unit 2 TS LCO 3.6.1.1, Primary Containment, required, in part, primary containment shall be operable in modes 1, 2 and 3. Also, it requires primary containment to be restored within one hour (Condition A.1), otherwise be in mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and mode 4 in 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (Condition B.1). Contrary to the above, while unit 2 was in modes 1, 2, and 3 between January 28 and January 29, 2023, the primary containment was not operable in that leakage rates exceeded the maximum allowable primary containment leakage limits (La) for greater than one hour and the unit was not placed in mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Enforcement Action: This violation is being treated as an NCV, 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 July 27, 2023, the inspectors re-exited the integrated inspection results to Mr. Matt Busch, Plant Manager (PM), and other members of the licensee staff.
On July 14, 2023, the inspectors presented the integrated inspection results to Mr. Johnny Weissinger, Site Vice President (SVP), and other members of the licensee staff.
On May 25, 2023, the inspectors presented the license renewal inspection results to Mr. Matt Busch, PM, and other members of the licensee staff.
On May 18, 2023, the inspectors presented the ISFSI operation inspection results to Mr. Johnny Weissinger, SVP, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Operation with Degraded System Voltage
6.1
Naturally Occurring Phenomena
20.2
NMP-GM-025
Seasonal Readiness Process (Hatch Summer 2023)
05/11/2023
Procedures
NMP-OS-020
Station Response to Southern Company Alert Conditions
2.0
H-16000
Unit 1 Nitrogen Inerting System P&ID
53.0
Drawings
H-26083
Unit 2 Nitrogen Inerting System P&ID
45.0
Residual Heat Removal System
46.4
Residual Heat Removal System
45.2
Reactor Core Isolation Cooling System
28.4
Diesel Generator Standby AC System
29.5
Containment Atmospheric Control and Dilution Systems
29.7
Procedures
RHR Service Water Pump Operability
2.4
Work Orders
SNC970789
Check Valve Internal Inspection of E11F005B
04/25/2023
Corrective Action
Documents
Resulting from
Inspection
Condition Reports
(CR)
10962383, 10962386
NMP-ES-035-
019-GL02-F06
Diesel Generator Building Elevation 130
1.0
NMP-ES-035-
019-GL02-F17
SHT 1 of 2
U1 Reactor Building
1.0
NMP-ES-035-
019-GL02-F26
SHT 1 of 2
U2 Turbine Building
1.0
NMP-ES-035-
019-GL02-F34
SHT 1 of 2
U2 Reactor Building
1.0
NMP-ES-035-
019-GL02-F38
U2 Radwaste Building El. 164/178
1.0
Drawings
NMP-ES-035-
GL02-F18
U1 Radwaste Building El. 108
1.0
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
H-LR-AF-00107
Reactor Feed Pump Turbine Cooling, Safety Relief Valve
Open, Loss of Feedwater Heating, Loss of Coolant Accident
in the Drywell (DW), DW Spray Failure, RCIC Controller
Failure
2.0
NMO-GM-005-
2
Human Performance Tools Instructions
2.2
NMP-OS-007
Conduct of Operations
18.2
NMP-OS-007-001
Conduct of Operations Standards and Expectations
20.0
Procedures
NMP-TR-405
Simulator Exercise Guide and Evaluation
1.2
Corrective Action
Documents
Condition Reports
(CRs)
10782681, 10808797, 10809029, 10826537, 10830164,
10834710, 10836065, 10836075, 10838151, 10838444,
10840717, 10841444, 10844832, 10845271, 10845283,
10845625, 10845859, 10846731, 10848170, 10850994,
10893079, 10902023, 10921781, 10906357, 10908441,
10917467, 10925913, 10937215, 10943102, 10946848,
10951649, 10954274, 10954487, 10955326, 10956529,
10957337, 10961529, 10962705, 10962319,
Engineering
Evaluations
Technical
Evaluation
1111085, 1121250
Maintenance
Rule Function
Scoping
N21 Condensate/Feedwater Maintenance Rule SSC
Function Consolidation and Performance Criteria Selection
form
06/24/2019
Maintenance
Rule Function
Scoping
Z99 PAMS Pseudo Maintenance Rule SSC Function
Consolidation and Performance Criteria Selection form
Miscellaneous
Maintenance
Rule Function
Scoping
H12-01 Annunciators
03/23/2023
Procedures
NMP-ES-027
10.4
Hatch Unit 2 RICT Configuration Risk Monitor Current Risk
Summary Report
04/10/2023
Hatch U1 RICT On-Line Configuration Risk Monitor
06/09/2023
Miscellaneous
Hatch U2 RICT Configuration Monitor Current Risk
Summary Report
06/13/2023
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-ET-23-2E11-
00036
04/10/2023
NMP-GM-031-
2-F01
RICT Projection Details (Hatch Unit 1 for 1A EDG Outage)
3.1
Procedures
NMP-OS-010-002
Hatch Protective Equipment Logs
2.3
Branch Work Order #1413-
2808-01
Design Package Cover Sheet for Applying Sealant Material
for 2E41F006
06/09/2023
Calculations
SMNH-12-008
Transient Heat-Up Analysis for Hatch Unit 1 & 2 RCIC Pump
Rooms.
04/30/2012
Corrective Action
Documents
Condition Reports
(CRs)
10957452, 10960049, 10963936, 10965633, 10968270,
10971170
H-17842
Unit 1 Feedwater Control System C32 Elementary Diagram
Sheet 1
30.0
H-44833
Unit 1 Feedwater Control System C32 Elementary Diagram
Sheet 5
8.0
H-44834
Unit 1 Feedwater Control System C32 Elementary Diagram
Sheet 6
10.0
Drawings
H13367
Single Diagram 120/208V Station Service Switchgear 1A
Sheet 1 of 2 MPL 1R23-S021
24.0
NMP-ES-050-F01
RER Response Form
11/28/2022
SNC1481086
RER Response for Unit 1 RECIRC Runback to 60.3%
05/21/2023
Engineering
Evaluations
SNC1490936
Leak rate Analysis-2E41F006
06/08/2023
Miscellaneous
Hatch Project Support - Engineering RCIC Operation
Without Room Cooling (Letter)
2/09/2001
Operability
Evaluations
NMP-AD-012-F01
Prompt Determination of Operability of the 2A Fuel Oil
Transfer Pump
09/30/2020
AGASTAT Timing Relay Calibration
30.6
Westinghouse - ABB Type SV Relay
1.4
NMP-AD-012
15.1
NMP-AD-012-F01
Operability Determination Support Basis
03/31/2023
Procedures
NMP-AD-031-
GL01
Past Operability / Functionality review
1.0
Work Orders
Work Orders
1465023, 1465119, 1465120, 1485487
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
(SNC)
Corrective Action
Documents
Condition Report
(CR)
10968707, 10968402, 10982950
Corrective Action
Documents
Resulting from
Inspection
Condition Report
10966778
Drawings
H-26078
Standby Gas Treatment System P&ID
35.0
2-23-056
Annunciator and Plant Component Sheet
06/26/2023
2-DT-23-2T46-
00005
10.CFR.50.59 Screening
06/26/2023
Operability
Evaluations
2-DT-23-T46-
0005
Applicability Determination
06/26/2023
20VAC RPS Power Supply System
15.9
15.3
RPS Channel Test Switch Functional Test
5.18
Reactor Manual Scram Functional Test
5.2
Reactor Manual Scram Functional Test
3.6
Residual Heat Removal Pump Operability
2.2
RHR Valve Operability
24.3
Core Spray Pump Operability
06/15/2023
Core Spray Pump Operability
24.6
Core Spray Valve Operability
06/14/2023
HPCI Pump Operability
39.2
RCIC Pump Operability
29.2
Diesel Generator 1C Monthly Test
20.4
Diesel Generator 1A Fast Start Test
16.5
DG Fuel Oil Transfer Pump Surveillance Test
5.7
RS-1-1
Diesel Fire Pump 1X43-C002A Engine Replacement
Functional Test
1.0
Procedures
Diesel, Alternator, And Accessories Inspection
31.29
Work Orders
Work Orders
(SNC)
118253, 621102, 621111, 831521, 849041, 854225, 954467,
955851, 956420, 955850, 984289, 1078685, 1084252,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
1131905, 1115263, 1462365, 1087833
Reactor Coolant Weekly (Sample Sheets for Unit 1 - April
22 through March of 2023)
71151
Procedures
Reactor Coolant Weekly (Sample Sheets for Unit 2 - April
22 through March of 2023)
Corrective Action
Documents
Corrective Action
Reports (CARs)
277846, 393867, 398875, 391000, 393867, 390477, 431972,
10704520, 10968633
Engineering
Evaluations
Technical
Evaluations (TEs)
27537, 1127539, 1127540, 1127541, 1127616, 1127617,
27621, 1127622, 1127624, 1127625, 1127627, 125160,
25159, 1125034, 1125026, 1125161, 1125949, 1125042,
24304, 1130223
ALS UIN
09F66D0
1R43N101A - Woodard UG8 SN#256969, Governor Oil
Analysis
05/09/2023
Crew Learning
Safety System Functional Failure Report Not Submitted
Crew Learning
Crew Learning Communication and Turnover - Hatch
CDO/Projects
05/23/2023
Crew Learning
CT Circuit Checks
03/08/2023
Miscellaneous
H-ES-PP-23106
Engineering Case Study on Design Implementation
1.0
Condensate & Feedwater System
2.11
71152A
Procedures
Condensate & Feedwater System
57.3
Corrective Action
Documents
Corrective Action
Report (CAR)
380222, 415509
H22023A
Outage Critical Path Schedule
03/31/2023
19:45
LER 2023-001-00
Primary Containment Penetration Exceeded Maximum
Allowable Primary Containment Leakage Rate (La)
04/05/2023
LER 2023-002-00
Unit 2, Manual Scram due to trip of both reactor recirculation
pumps
05/26/2023
Miscellaneous
NMP-OS_009-
F01
Restart Plant Review Board (PRB) Review Checklist
04/01/2023
Reactor Recirc ASD High-Speed Stops Setpoint
Determination
8.0
Procedures
IP-ENG-001
Standard Design Process (SDP)
3.0
23