IR 05000321/2024011
| ML24354A088 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 12/19/2024 |
| From: | Masters A NRC/RGN-II/DORS |
| To: | Coleman J Southern Nuclear Operating Co |
| References | |
| IR 2024011 | |
| Download: ML24354A088 (1) | |
Text
SUBJECT:
EDWIN I. HATCH NUCLEAR PLANT UNITS 1 & 2 - COMPREHENSIVE ENGINEERING TEAM INSPECTION REPORT 05000321/2024011 AND 05000366/2024011
Dear Jamie Coleman:
On December 11, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Edwin I. Hatch Nuclear Plant Units 1 & 2 and discussed the results of this inspection with Matt Busch 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. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
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 Nuclear Plant Units 1 & 2.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Edwin I. Hatch Nuclear Plant Units 1 & 2.
December 19, 2024 This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Anthony D. Masters, Chief Engineering Branch 1 Division of Operating Reactor Safety Docket Nos. 05000321 and 05000366 License Nos. DPR-57 and NPF-5
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000321 and 05000366
License Numbers:
Report Numbers:
05000321/2024011 and 05000366/2024011
Enterprise Identifier:
I-2024-011-0024
Licensee:
Southern Nuclear Operating Co., Inc.
Facility:
Edwin I. Hatch Nuclear Plant Units 1 & 2
Location:
Baxley, GA
Inspection Dates:
October 21, 2024 to December 11, 2024
Inspectors:
J. Alamudun, Reactor Inspector
B. Bowker, Senior Reactor Inspector
L. Day, Reactor Inspector
T. Fanelli, Senior Reactor Inspector
C. Franklin, Reactor Inspector
M. Hagen, Reactor Inspector
J. Lizardi-Barreto, Reactor Inspector
E. Morris, Resident Inspector
J. Rolland, Mechanical Engineer
T. Su, Reactor Inspector
Approved By:
Anthony D. Masters, Chief
Engineering Br 1
Division of Operating Reactor Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a Comprehensive Engineering Team Inspection at Edwin I. Hatch Nuclear Plant Units 1 & 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Follow Corrective Action Program to Maintain/Replace Flexible Hoses Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321,05000366/2024011-01 Open/Closed None (NPP)71111.21M The NRC identified a Green non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Action," for the failure to assure that a condition adverse to quality was promptly identified and corrected. Specifically, the licensee failed to follow the corrective action program to ensure the flexible hoses on the Emergency Diesel Generator (EDG) were replaced in a timely manner.
Reliability of the Reactor Protection System Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321,05000366/2024011-02 Open/Closed
[H.3] - Change Management 71111.21M The NRC identified two examples of a finding of very low safety significance (Green) and an associated Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to ensure the reliable operation of the Reactor Protection System (RPS) from degraded safety-related components in use beyond their expected life span in the Analog Transmitter Trip System (ATTS).
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000366/2024011-03 Seismic Qualification of Batteries with Cracked Cell Lids and Potential Part 21 for Nonconforming Epoxy Construction 71111.21M Open
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 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
===71111.21M - Comprehensive Engineering Team Inspection The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:
Structures, Systems, and Components (SSCs) (IP section 03.01)===
For each component sample, the inspectors reviewed the licensing and design bases including:
- (1) the Updated Final Safety Analysis Report (UFSAR);
- (2) the Technical Specifications (TS); and
- (3) the Technical Requirements Manual (TRM). The inspectors reviewed a sample of operating procedures (including normal, abnormal, and emergency procedures) and overall system/component health (including condition reports and operability evaluations, if any). The inspectors performed visual inspections of the accessible components to identify potential hazards and/or signs of degradation. Additional component specific design attributes reviewed by the inspectors are listed below.
- (1) Unit 2: Safety Relief Valves, (2B21-F013A-M)
- Design bases and design assumptions
- Energy sources
- Acceptance criteria for tested parameters
- Equipment qualification
- (2) Unit 2: Recirc Pump Discharge Isolation Motor Operated Valve (MOV) (2B31-F031A)
- Environmental qualification
- Protection against external events:
o Seismic o
- Mechanical design o
Weak link analysis o
Required thrust (torque)o Pressure locking and/or thermal binding o
Closure/Opening time o
Maximum allowed leakage o
Maximum differential pressure
- Test/inspection procedures, acceptance criteria, and recent results:
o Leakage o
IST o
TS required surveillance
- Motor power requirements:
o Voltage drop o
Control logic o
Thermal overload o
Required minimum voltage o
Degraded voltage effects o
Brake horsepower o
Motor thermal overload protection o
Cable ampacity
- (3) Unit 1: Feedwater Inboard Isolation Check Valve (1B21-F010A)
Unit 1: Feedwater Outboard Isolation Check Valve (1B21-F032A)
Unit 2: RHR SW Pump Discharge Check Valve (2E11-F005A)
- Environmental qualification
- Protection against external events:
o Flooding o
Seismic
- Mechanical design o
Maximum allowed leakage o
Maximum differential pressure
- Test/inspection procedures, acceptance criteria, and recent results:
o Leakage o
IST o
Leak Rate Testing (LRT)
- Potential degradation is monitored or prevented
- Component replacement is consistent with in service/equipment qualification life
- Equipment qualification is suitable for the environment
- Design requirements
- (5) Analog Transmitter Trip System Design & Reliability
- Visual Inspection during walkdown
- Environmental conditions
- Design requirements
- Surveillance testing & maintenance records
- Periodic testing, inspection, and post-test analyses
- Conformance with manufacturer instructions for installation, maintenance, testing and operation
- (6) Electrical Penn Design and Repairs (EQ & Cont. Isolation)(1T52-X105C & 2T52-X105A)
- Environmental conditions
- Design requirements
- Surveillance testing & maintenance records
- Periodic testing, inspection, and post-test analyses
- Conformance with manufacturer instructions for installation, maintenance, testing and operation
- (7) Unit 1: RCIC Control Circuits (Include Relays and Valves)
- Design bases and design assumptions
- Energy sources
- Acceptance criteria for tested parameters
- Equipment qualification
- Condition Reports review
- (8) Station Service 1A & 2B Batteries
- Potential degradation is monitored or prevented
- Component replacement is consistent with in service/equipment qualification life
- Equipment qualification is suitable for the environment and design basis seismic event
- Design requirements
- (9) Unit 2: Emergency Diesel Generator 2C
- Conformance with manufacturer instructions for installation, maintenance, testing and operation
- Surveillance testing and maintenance records
- Component replacement, including Rubber Couplings and Hoses, is consistent with in service/equipment qualification life.
- Periodic testing, inspection, and post-test analyses
Modifications (IP section 03.02) (3 Samples)
- (1) SNC1348498 Unit 1 Reactor Feed Pump Minimum Flow Valve Upgrade (2)
===1039002601 Unit 1 Plant Service Water (PSW) Modifications
- (3) SNC759053 Feedwater Heater & MSR Drain Tank Level Control Upgrade 10 CFR 50.59 Evaluations/Screening (IP section 03.03)===
- (1) SNC1185724 LTAM H-17-0068 APRM Upscale/Rod Block Setpoint Change
- (2) SNC1295040 UNIT 2 2H11P603-115 Annunciator (Primary Containment Pressure High) Setpoint Change
- (3) SNC1183781 U1 Off gas Power Source
- (4) SNC1336904 Unit 2 Main Steam Line Isolation Logic Modification to Comply with New TS 3.7.10
- (5) SNC1179371 Unit 1 Plant Service Water Pump Seismic Restraint and Pump Column Spacing
- (6) SNC1440523 Change Orientation of Valve Disc (2T48-F319 and F320)
- (7) SNC1390882 Unit 2 RHRSW Division 1 Weld Overlay
- (8) SNC1453978 Replacement Cooling Coils for 1T41B004A and 1T41B004B
- (9) SNC1463171 2P41-C001C Motor Interference
- (10) SNC1219268 Unit 1 Steam Jet Air Ejectors 3rd Stage Controller Power
- (11) SNC1332462 CST Core Spray Pipe
- (12) SNC435914 Unit 2 Refuel Bridge Replacement
Operating Experience Samples (IP section 03.04) (2 Samples)
- (2) IN 2007-21, Supplement 1(2020): Pipe Wear Due to Interaction of Flow-Induced Vibration and Reflective Metal Insulation
INSPECTION RESULTS
Failure to Follow Corrective Action Program to Maintain/Replace Flexible Hoses Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321,05000366/2024011-01 Open/Closed None (NPP)71111.21M The NRC identified a Green non-cited violation (NCV) of Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action," for the failure to assure that a condition adverse to quality was promptly identified and corrected. Specifically, the licensee failed to follow the corrective action program to ensure the flexible hoses on the Emergency Diesel Generator (EDG) were replaced in a timely manner.
Description:
PM 2R43S001C13 was created in 2009 to establish an EDG flexible hose replacement program (all EDG flexible hoses replaced on a 12-year frequency). Work order SNC676493 was generated in 2017 to implement this program by replacing all flexible hoses and couplings on the 2C EDG. During the replacement work in 2017, CR 10422542 identified 5 flexible hoses that were not replaced and deemed them acceptable for reuse with the recommendation to replace them at the next available opportunity. The CR was closed on October 26, 2017, with no follow-on work order to replace the 5 re-used flexible hoses on the 2C EDG as recommended by engineering.
The diesel vendor recommends the hoses be inspected and replaced every 10 years, and industry operating experience indicates hose failures have occurred after 15 years of service.
These five hoses have now been in service for greater than 15 years, which is beyond the periodicity set by their PM and into the range of OE failures.
Corrective Actions: The other EDGs were reviewed for hose replacements. The licensee initiated a CR to document the five 2C ED hoses exceeding the PM replacement frequency.
The CR states that a work order needs to be created to replace the five flexible hoses on the 2C EDG during the upcoming system outage on 7/14/2025. Engineering also performed an evaluation justifying continued operation of the 2C EDG.
Corrective Action References: CR 11125432
Performance Assessment:
Performance Deficiency: The failure to follow the corrective action program to promptly correct a condition adverse to quality by closing a condition report needed to track replacement of EDG flexible hoses was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inspectors determined the assurance of the reliability of the hoses was impacted by extension of the service life beyond the vendor recommendation and industry OE.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Exhibit 2
- Mitigating Systems Screening Questions, Item A. Mitigating SSCs and PRA Functionality (except Reactivity Control Systems), which specifies that if the finding affects the design or qualification of a mitigating SSC, but it maintained its operability or PRA functionality, then it screens to 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:
Violation: 10 CFR Part 50 Appendix B, Criterion XVI, Corrective Action, states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, since October 2017, the licensee failed to ensure corrective measures were taken to replace EDG flexible hoses prior to exceeding the service life. Specifically, the licensee closed the CR documenting the need to replace five EDG flexible hoses without verifying another document existed to track timely completion of this work.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Reliability of the Reactor Protection System Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000321,05000366/2024011-02 Open/Closed
[H.3] - Change Management 71111.21M The NRC identified two examples of a finding of very low safety significance (Green) and an associated Non-Cited Violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to ensure the reliable operation of the Reactor Protection System (RPS) from degraded safety-related components in use beyond their expected life span in the Analog Transmitter Trip System (ATTS).
Description:
The inspectors reviewed the RPS preventative maintenance to determine if it could mitigate age related degradation of the ATTS components. Title 10 CFR 50.55a(h)(2),
Codes and standards, requires compliance with IEEE 279-1971, Criteria for Protection Systems for Nuclear Power Generating Stations, and the licensing basis for the protection systems. The Updated Final Analysis Report (UFSAR) specified that the licensing basis for ATTS qualification was based on the criteria of 279-1971 Section 4.4, Equipment Qualification, via its implementation described in 323-1971/74, IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Generating Stations. For the ATTS, the original General Electric (GE) generic qualification dated 4/1977 was found acceptable by the NRC to meet these standards on 6/27/1978 if performed as written per IEEE 323-1974. This generic qualification was NEDO-21617, ATTS Design Submitted to NRC/w SER (ML19267A276).
The qualification test report, SX19762-A, Analog Trip System Qualification Report Volume 1, which was performed per the generic qualification, specified that the limited age conditioning performed produced age-related anomalies in the master trip units (MTU), slave trip units (STU), and voltage converters that affected their performance and limited their life span. GE extended the expected life span of the ATTS cards in 1984 to 40 years.
Example 1, trip units; The current Hatch position is that the ATTS trip units are used in service until they fail and then are replaced. Based on limited age conditioning and analysis, the GE qualification objective of the ATTS trip units was to achieve a 40-year life span for the MTU and STU with calibration trending. The inspectors noted that during testing, the MTU and STU exhibited some non-calibration-related, age-related anomalies that indicated a shorter life than 40 years. Although calibration trending was an indicator of certain aging effects, it did not track age-related degradation that does not immediately affect calibration indications. The GE performance specification (22A716), stated, a forty-year life objective is stipulated for the MTU and STU. Any components which can be expected to have greater than 2.5 failures per million hours of operation under Trip Unit operating conditions of this specification will be identified for change out at specific times. Age related failures were identified at Hatch before the ATTS was 25 years old, which aligns with the age conditioning failures during qualification.
In 2010, a root cause analysis investigated several significant failures of the trip units. It identified failed capacitors as direct causes of many failures and aging degradation as contributing causes along with inadequate emphasis on replacements and preventative maintenance (PM). By 2010, these trip unit failures have exceeded the stipulation of 2.5 failures per million hours of operation. As a result of the root cause findings, a PM replacement strategy was adopted. However, this was cancelled in favor of running the ATTS components to failure due to cost. Now in 2024, the ATTS has been in operation for more than 39 years and there have been many more age-related failures that could impair the ATTS in performing its safety function. The licensee staff has resubmitted a request to implement a Preventative Maintenance Change Request (PMCR) 106841. The inspectors determined that the ATTS was not meeting the requirements in IEEE 279-1971 sections 4.3, Quality of Components and Modules for not maintaining the quality of the ATTS with replacement schedules and 4.4, Equipment Qualification for not verifying that the protection system equipment would meet, on a continuing basis, the performance requirements. Hatch entered this issue into the corrective action program under CR11134949 and Example 2, voltage converters; The report (SX19762-A) identified the voltage converter's life limiting components were the electrolytic capacitors and trim potentiometers which both had approximately equivalent life spans. After approximately 1999, Operating Experience (OE)widely identified the acceptable reliable life span for electrolytic capacitors as 10 years, including shelf life. The OE was documented in the Electric Power Research Institutes (EPRI) guidance in TR12175-1999 for capacitors specifically and TR1003096-2001 for power supplies containing capacitors. EPRI specified that under the best conditions, a 5-to-10-year replacement schedule can be expected. Later, the NRC issued information notice 2012-11, Age-Related Capacitor Degradation. As a result of the OE the licensee changed the original 27-year life to a replacement frequency of 10 years.
A PMCR for Unit 1 increased the assumed life span from 10 to 18 years based on the temperature of the control room. To change the Unit 2 replacements, Unit 2 copied the U1 justification. The inspectors determined the justification failed to account for the 32.5°F temperature rise inside the ATTS panels identified in the qualification report, and the higher temperature rise inside the voltage converters at the capacitors, which would nullify this justification.
Corrective Actions: The licensee initiated several Condition Reports to address the concerns noted above. CR 11129834 which was to have the PMs changed back to the original 10-year frequency. Additionally, CR 11129840 identifying locations that are beyond the 10-year frequency established with CR 11129834, 15 voltage converters on Unit 1 and 9 voltage converters on Unit 2.
Corrective Action References: CRs 11129834,11129840, and 11134949
Performance Assessment:
Performance Deficiency: The failure to ensure the reliable operation of the RPS from degraded safety-related components in use beyond their expected life span was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the safe operation of the RPS throughout the life of the plant effects the reliability and capability of the safety systems that respond to initiating events to prevent undesirable consequences.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Exhibit 2
- Mitigating Systems Screening Questions, Item A. The finding was a deficiency affecting the design or qualification of a mitigating SSC, and it maintain its operability or PRA functionality Screen to Green.
Cross-Cutting Aspect: H.3 - Change Management: Leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
Since 2023, leaders did not use a systematic process for evaluating the reliability of the reactor protection system when implementing a change to the assumed life span of its components so that nuclear safety remained the overriding priority.
Enforcement:
Violation: Title 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of SSCs. IEEE 279 Section 4.4, Equipment Qualification, requires, that type test data or reasonable engineering extrapolation based on test data shall be available to verify that protection system equipment shall meet, on a continuing basis, the performance requirements determined to be necessary for achieving the system requirements.
Contrary to the above, since 2009, the licensee failed to review for suitability of application the life span of electrolytic capacitors and trip unit cards with type test data or reasonable engineering extrapolation based on test data to verify that the reactor protection system (Analog Transmitter Trip System) could meet, on a continuing basis, the performance requirements determined to be necessary for achieving the system requirements.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Unresolved Item (Open)
Seismic Qualification of Batteries with Cracked Cell Lids and Potential Part 21 for Nonconforming Epoxy Construction URI 05000366/2024011-03 71111.21 M
Description:
The team identified an Unresolved Item (URI) concerning large cracks on battery cell lids/covers on the Hatch Unit 2 (05000366) Trains A and B Class 1E station battery cells. Some of the B cells had an epoxy applied to cover the cracks. The epoxy was not maintaining its adherence to the lid material and there were still uncovered cracks on these lids. The A cells had an extensive number of large cracks that were not previously identified by licensee surveillances (performed from August to October 2024) prior to teams observations. A Condition Report (CR)111234451 documented the observation and recommended applying the epoxy to the lids.
The team reviewed the material compatibility and qualification documentation for the epoxy.
The compatibility data stated that it was not suitable for application in acidic environments, yet this epoxy would be exposed to the acidic electrolyte and vapors from the lead-acid battery cells during and after curing. For qualification, the team noted that there was no Technical Evaluation (TE) approving the use of the epoxy for the application. Furthermore, there was no commercial grade qualification report, from the supplier or battery cell manufacturer, that documented the acceptable application for the epoxy on Class 1E electrical basic components. In addition, the manufacturer specified that this same epoxy was used during manufacturing to adhere the lid to the cell jar.
The team inquired as to whether the large cracks could affect the seismic qualification of the batteries. In the degraded condition, the cracks could open, and the lid could become loose possibly releasing the electrolyte inside. These conditions could affect the functional performance of the battery during and after a design basis seismic event. The licensee/manufacturer stated that the cell lids do not have a safety function, they do not provide structural support to the jar, and the jar contains the electrolyte during seismic interactions, not the lid. The licensee provided seismic test reports to support this position.
Nevertheless, the tests did not demonstrate the functionality of the battery cells as result of electrolyte leakage or demonstrate that the required electrolyte levels could be maintained without the lids to perform safety functions either short or long term.
Specifically, one test set up included sealing the cracks with tape while another test only verified electrical continuity, and did not prove the cell's ability to provide power for the required accident time. During testing, new hairline cracks appeared and existing cracks propagated further. Based on this testing, the team determined that the new cracking questioned the premise that the lids serve no structural or seismic support. Moreover, none of these tests evaluated seismic behavior of the battery cells repaired with the epoxy adhesive, with wide open large cracks, or with the lids removed.
A significant number of battery cells contain large cracks in their lids that could allow a significant amount of electrolyte to spill out during seismic events, which is a concern that could affect the safety functions (i.e., a 10 CFR Part 21 defect). The team reviewed the battery vendor, C&D Technologies, report and noted the vendor concluded that this condition could not create a substantial safety hazard and therefore was not a defect. The definition of a defect in 10 CFR Part 21 also includes a condition or circumstance involving a basic component that could contribute to the exceeding of a safety limit, as defined in the technical specifications of a license for operation issued under part 50 or part 52 of this chapter. Therefore, this condition may still meet the definition of a defect requiring reporting per 10 CFR Part 21.
Planned Closure Actions: The NRC needs to review the additional new information provided by the licensee to determine if a performance deficiency exists. If this condition and repairs bound the seismic and environmental qualifications, determine whether reasonable assurance can support the functionality of the basic component, determine if the degraded condition meets the definition of a defect per 10 CFR Part 21 reporting requirements, and whether the repairs are adequate and used a qualified material.
Licensee Actions: The licensee entered these issues into the corrective action program. The NRC will contact the site if any further evaluations are needed.
Corrective Action References: CR11123451
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On December 11, 2024, the inspectors presented the Comprehensive Engineering Team Inspection results to Matt Busch and other members of the licensee staff.
- On November 8, 2024, the inspectors presented the preliminary inspection results and on 12/11/2024 the inspectors presented an update to the inspection results to Matt Busch and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
BH0-C-C113.0
Bolt Capacities for Partial Thread Engagement
Rev. 0
SNC613983
U2 Refuel Floor Rail Modifications
Rev. 2
HNP-2B31-F031A
Thrust and Torque Calculation
Rev. 1
S-63279
Seismic Test Report for C&D Batteries with Cracked Cells
and Cracked Lids
Rev. 1
S-70804B
Seismic Calculation for Battery Racks for Battery Room 1B
Rev. 1
S-76562
Refueling Platform Assembly Structural Evaluation
Rev. 2
S43532-C
Seismic Calc. for Gould Battery Racks
Rev. 3
SCNH-93-090
Seismic Qualification of Battery Racks for the Replacement
of Batteries 2R42-S001B
08/09/1993
SCNH-93-091
Seismic Qualification of Battery Racks for the Replacement
of Batteries 2R42-S001A
08/09/1993
SENH-11-002
Evaluate the Station Auxiliary System for Hatch Unit 2 for
Various Plant Configurations to Support Acceptability of
Electrical Equipment Operation
Rev. 7
SMNH-14-009
Emergency Diesel Generator Lube Oil Consumption
Rev. 1
SMNH-93-003
DP for Reactor Recirculation System MOV's
Rev. 5
SMNH-95-011
Weak Link Analysis of 28" Lunkenheimer Valves
Rev. 0
SMNH-95-016
Load Sensitive Behavior Comparison
Rev. 0
SMNH-95-031
Valve Factor for Lunkenheimer Valves
Rev. 1
Calculations
SNC105747
DCP U1 Refuel Bridge Replacement
Rev. 7
Corrective Action
Documents
10950274, 102619,
11067437,
11068256,
11070876,
11092197,
2010103997,
2009101597,
2010103638,
278645, 10422542,
10919507,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
10959509,
11050625,
11050644,
11088925,
11105111, 106841,
84533
11121037
2R25-S035 has a broken lock
10/22/2024
11121172
Documentation not Found in D2
10/22/2024
11121243
RHR SW Strainer 1D Outlet Isolation Valve
10/22/2024
11121338
2T48F320 Anaconda flex conduit rubber sleeve
10/22/2024
11121346
2T48F319 valve is very close to 2G41 pipe
10/22/2024
11121383
U1 HPCI Thread Engagement (Lube Oil)
10/22/2024
11121553
Peak Containment Pressure Response to a DBA-LOCA
10/23/2024
11121877
Remove Flow Arrows on 2T48-F319 and 2T48-F320
10/24/2024
11121957
Lighting Tie Wire Broken in Battery Room
10/24/2024
11121959
Missing anchor bolts and nuts in crane supports
10/24/2024
11121976
Unsecured valve restraints identified during EDG 2C NRC
walkdown
10/24/2024
11123451
2A Station Service Lid Cracking
10/29/2024
11123814
Refuel Bridge Rail Missing Anchor
10/30/2024
11125236
ASME Section XI Class III Boundary Discrepancy
11/5/2024
11125432
EDG 2C flex hose not replaced
11/05/2024
11125705
Missing Initials on QC Hold Point for WO SNC1155961
11/6/2024
11125709
Procedure Review Needed
11/06/2024
11125769
IM 631271 Shelf-Life
11/06/2024
11126278
Battery epoxy question
11/8/2024
11129834
Inadequate justification for extending ATTS power supply
electrolytic capacitor replacements
11/21/2024
11129840
Schedule ATTS power supplies to be replaced
11/21/2024
Corrective Action
Documents
Resulting from
Inspection
11134949
ATTS trip card service life potential violation
2/12/2024
15-476-1019-3
Limitorque Wiring Diagram
2/23/1967
LVP Electric Penetration Assembly for E. I. Hatch Unit 1
08/23/2001
Drawings
B-44106
Edwin I. Hatch Nuclear Plant Unit No. 1 Control Building @
Rev. 0
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
EL. 112'-0" Battery Room 1B - Cell Rack Modifications -
Sections and Details
B19631 Sheet 1
Edwin I. Hatch Nuclear Plant Unit No. 1 Penetration Seals
Type, Number and As-Built Location
Rev. 7
B19631 Sheet 18
Edwin I. Hatch Nuclear Plant Unit No. 1 Penetration Seals
Type, Number and As-Built Location
Rev. 5
B27046
MCC Equipment List: Rx Bldg 600 Volt MCC 2E-A
2R24S018A
Rev. 18
H-13410
Hatch Nuclear Plant Wiring Diagram Miscellaneous
Instruments
Rev. 27
H-13427
Hatch Nuclear Plant Wiring Diagram Control Console -
Panel A - H11-P650
Ver. 49
H-13428
Hatch Nuclear Plant Wiring Diagram Control Console
Panel A-H11-P650
Ver. 33
H-14237
Hatch RCIC Sys. E51, Elem. Diagram Sht. 8 of 9
Rev. 13
H-16334
Hatch RCIC Sys. P & ID Sheet No. 1
Ver. 51
H-17147
Hatch RCIC Sys. E51 Elementary Diagram Sheet 1 of 9
Ver. 58
H-21039
Rev. 48
H-26003
Reactor Recirculation System P&ID
Rev. 46
H-27021
Single Line Diagram - RX Bldg. 600V AC Essential MCC
2E-A & MCC 2E-B MPL 2R24-S018A & 2R24-S018B
Rev. 21
H16335
Hatch RCIC Sys. P & ID Sheet No. 2
Ver. 36
S-64356
Instruction Manual Fisher Fieldvue DVC6200 Digital Valve
Controller
Rev. 2
S-70805/400432-D
Rack Layout, Twin Tier for 40 Cells of NCX-33 Heavy
Seismic
2/01/1996
S70814/400514-D
Rack Layout, Twin Tier for 20 Cells of NCX-27 Heavy
Seismic
2/01/1996
SNC105747C025
Edwin I. Hatch Nuclear Plant Unit No. 1 Reactor Building
Platform Crane Rail @ EL. 228'-0" Plan Sections & Details
Rev. 2
SNC105747C026
Edwin I. Hatch Nuclear Plant Unit No. 1 Reactor Building
Superstructure Concrete Plan at EL. 228'-0" Pour 2 Neat
Line
Rev. 1
SNC613983C002
Edwin I. Hatch Nuclear Plant Unit No. 2 Reactor Building
Rev. 2
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Refueling Platform Rail Plan Sections & Details at EL. 228'-
0"
SNC613983C002
Edwin I. Hatch Nuclear Plant Unit No. 2 Reactor Building
Refueling Platform Rail Plan Sections & Details at EL. 228'-
0"
Rev. 2
SNC613983C005
Edwin I. Hatch Nuclear Plant Unit No. 2 Reactor Building
Reactor Well Shield Concrete Neat Line Plan at EL. 228'-0"
Rev. 1
SNC613983C006
Edwin I. Hatch Nuclear Plant Unit No. 2 Reactor Building
Fuel Pool & Cask Area Liner Plate & Neat Line West Wall
Elevation
Rev. 1
SNC613983C007
Edwin I. Hatch Nuclear Plant Unit No. 2 Reactor Building
Fuel Pool Neat Line Plan at EL. 228'-0"
Rev. 1
SNC759053J027
Hatch Feedwater Heater & MSR Hotwell, 1st & 2nd Stage
Reheater Valve Positioner & Control Valve
Rev. 4
SNC759053J091
Hatch 2nd Stage High Level Control Valves DRAG Valve
x 14 in
Rev. 2
SX28647/062310-D
Layout for 120 Cells NCX 1650/ NCX 1650 Heavy Seismic
Resistant
9/12/1975
SX28733
Catalog No. 1624.7C
Rev. 0
SNC1096017
Replace 2T48F320
Rev. 0
SNC1179371DECP
Unit 1 Plant Service Water Seismic Restraint and Pump
Column Spacing
09/24/2021
SNC1183781DECP
Design Equivalent Change Package: U1 Offgas Power
Source
2/14/2022
SNC1185724DECP
Design Equivalent Change Package: LTAM H-17-0068
APRM Upscale/Rod Block Setpoint Change
05/27/2022
SNC1219268
U1 Steam Jet Air Ejectors 3rd Stage Controller Power
Rev. 1
SNC1219574DECP
Unit 1 ASD Siemens Eagle Code and PLC Upgrade
Rev. 1
SNC1295040DECP
Design Equivalent Change Package: Unit 2 2H11P603-115
Annunciator (Primary Containment Pressure High) Setpoint
Change
04/29/2022
SNC1332462
Unit 1 Condensate Storage Tank Piping Replacement
Rev. 2
Engineering
Changes
SNC1336904DECP
Design Change Package: The Unit 2 Main Steam Line
Isolation logic will require modification to comply with the
08/15/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
new TS 3.7.10
SNC1348498DCP
Unit 1 Reactor Feed Pump Minimum Flow Valve Upgrade
03/01/2024
SNC1390882
U2 RHRSW Division 1 Weld Overlay
Rev. 0
SNC1440523
U2 Change Orientation of Valve Disc of 2T48F319 and
2T48F320
Rev. 1
SNC1440523DECP
Change Orientation of Valve Disc
Rev. 1
SNC1453978
Replacement Cooling Coils for 1T41B004A and
1T41B004B
Rev. 0
SNC1463171
2P41-C001C Motor Interference
Rev. 0
SNC429939DCP
Unit 2 Modified Three Stage Safety Relief Valve
Replacement
2/16/2013
SNC435914
U2 Refuel Bridge Replacement
Rev. 11
SNC759053
Unit 1 Feedwater Heater & MSR Drain Tank Level Control
Upgrade N22
Rev. 7
SNC980080
07/19/2023
SNC980081
NFPA 805-Coordination of 2R25-S035 & 2R24-S048
2/16/2021
SNC980801DECP
U2 RCIC MOV Circuit Modifications
Rev. 2
C&D Technologies Summary Report of Cover Crack
Project
05/15/2013
C&D Technologies VLA Cell Cover Crack Impact to Safety-
Related Function; Hydrogen Out Gassing Test Report
2/29/2016
1R25S029
Distribution Panel Commercial Grade Dedication Plan Test
Results for Panel 1R25-S029
Rev. 1
1R25S069
Distribution Panel Commercial Grade Dedication Plan Test
Results for Panel 1R25-S069
Rev. 1
DCP SNC759053
Unit1 Feedwater Heater & MSR Drain Tank Level Control
Upgrade N22 50.59 Evaluation
Rev. 3
Analog Transmitter / Trip Unit System for Subject:
Engineered Safeguard Sensor Trip Inputs, submitted to
NRC/w SER (ML19267A276)
04/27/1978
Licensing Topical Report-Analog Trip System for
Engineered Safeguard Sensor Trip Inputs
07/1982
Engineering
Evaluations
QR-105463-
01/S61204
Environmental and Seismic Qualification Report of Type
LCR-29 and LCY-35 125/250 Volt DC Storage Battery
2/23/1994
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Station Battery 2A and 2B
QR-66171-01
Environmental and Seismic Qualification Report of Type
LCR-29 and LCY-35 125/250 Volt DC Station Service
Battery 1R42-S001A and 1R42-S001B
01/22/1993
S-77858
Hatch Unit 1&2 Failure Modes and Effects Analysis of
Ovation Workstations
Ver. 2
S-77859
DCIS Platform & Standards Failure Mode and Effect
Analysis of the Ovation Control and I/O System
Ver. 2
S-77861
Hatch Nuclear Automation Instrumentation and Control
Systems Software Failure Analysis of the Ovation Platform
Ver, 2
S60633
Seismic Report - Unit 2 Station Service Batteries 2A & 2B
MPL# 2R42-S001A & B
09/22/1992
S78054
Unit No. 1 & 2 Seismic Summary Document for Fisher 18
20 PCIV
Rev. 1
SCNH-93-012
Evaluate Rack Mods Station Service Battery 1A & 1B
Seismic Evaluation of Modified Battery Racks
2/15/1992
SNC1185724AD
Applicability Determination: APRM Rodblock Setpoint
Change
01/26/2022
SNC1336904AD
Applicability Determination: The Unit 2 Main Steam Line
Isolation logic will require modification to comply with the
new
08/04/2022
SNC1718229
Plant Hatch Station Service Battery 1B - Warped/ Buckling
Negative Battery Plate
Rev. 1
SX-19762-
A/NEDO-30039-1
Analog Trip System Qualification Report Volume 1
Rev. 1
TE-1162930
Technical Evaluation for NRC issued Information Notice 2007-21, Supplement 1: Pipe Wear Due to Interaction of
Flow-Induced Vibration.
10/01/2024
C&D Technologies Lid Cracking Memorandum
05/16/2014
HPCI Turbine And Auxiliaries Major Inspection
10/31/2022
Miscellaneous
H-20-0075
The 1A & 1B SJAE 3rd stage controllers 1N11R501 &
1N11R502 share a common power source. This represents
an SPV
08/11/2020
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
IPS-2103
Factory Test Procedure for Electrical Penetration
Assemblies for Hatch Nuclear Plant
Rev. A
MIS-16-004
Hatch 5th 10-Year Interval Inservice Testing Program
2/09/2024
S 60969
Replacement Drywell Electrical Penetration Assy-Seismic
Qualification Report
04/29/1992
S-42606 - 2B31-
F031A,B
Class 1 Nuclear Design Report for 800# Stainless Steel
Discharge Gate Valves
06/22/1983
S-43279
Limitorque Valve Actuator Qualification for Nuclear Power
Station Service
Rev. 1
S70616
Electrical Penetrations Installation
Instructions Unit 1
01/02/1986
S70731
Qualification Type Test Report - Limitorque Valve
Actuators with Type LR Motors.
04/13/1985
SE-DS-00-08
Specification for Electrical Penetration Assemblies for
Hatch Nuclear Plant
Rev. 1
SS-2102-303
Technical Specification for Electrical Penetration
Assemblies for Hatch Nuclear Plant
Rev. 2
SX25652
Instruction Manual for Large Nuclear Gate, Globe, and
Rev. 18
SX25718B
Lunkenheimer Maintenance Manual GE VPF-3174-67
Rev. 2
TR-00021
IEEE 535-2013 Environmental & Seismic Qualification Test
Report of C&D Technologies LCUN-33 (LCR-33) Cells & 2-
Step Battery Rack Assembly
Rev. 0
2R43S001C13
PM Replace All Flexible Hoses and Coupling on T
Rev. 0
25 VDC and 125/250 VDC System
Rev. 26.6
Diesel Generator Standby AC System
Rev. 30.11
Recirculation System Valve Operability
Rev. 9
RHR Service Water Pump Operability
Rev. 19.8
Installation of Nelson Electric Fire Stops & Seals
Rev. 3.6
Primary Containment Type B and Type C Leak Rate
Testing
Rev. 25
Primary Containment Periodic Type B & Type C Leakage
Tests
Rev. 37.4
Procedures
General Maintenance Procedure
Rev. 5.10
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Preventative Maintenance Procedure
Rev. 17.1
Preventative Maintenance Procedure
Rev. 18.1
Battery/Individual Cell Surveillance
Rev. 19.6
Battery Inspection
Rev. 9.12
Surveillance Procedure
Rev. 31.33
Diesel, Alternator, And Accessories Inspection
Rev. 31.33
HE-S-07-001
Procurement Specification for 125 Volt Diesel Generator
1A and 1C Batteries for E. I. Hatch Nuclear Plant - Unit 1
Rev. 1
HE-S-22-001
Safety Related Distribution Panels for Unit 1 and Unit 2
Specification
Rev. 4
NMP-ES-021
Structural Monitoring Program for the Maintenance Rule
Rev. 12
NMP-ES-202
Functional Classifications of Components, Parts, and
Materials
Rev. 1
NMP-ES-214
Shelf Life Determination
Rev. 1.1
NMP-GM-002
Corrective Action Program
Ver. 18
TR 1003096
Power Supply Maintenance and Application Guide
2/2001
TR12175
Capacitor Application and Maintenance Guide
8/1999
SNC10304721
PO EDG Hose from Fairbanks Morse LLC.
2/03/2023
SNC32021-0598
PO EDG Hardware from Fairbanks Morse LLC.
03/12/2020
SNC32021-0872
PO Turbo Charger Hose from Fairbanks Morse LLC.
11/12/2023
SNG10292667
PO Change Order SNG10292667
Rev. 1
Shipping Records
SNG10292667
Material Qualification Epoxy Adhesive PO Change Order
SNC10292667
08/16/2022
Work Orders
SNC866540,
SNC1146145,
SNC670102,
SNC871484,
SNC1023682,
SNC1413918,
SNC346525,
SNC316128,
SNC349471,
SNC551582,
SNC1074524,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
SNC1094642,
SNC1094633,
SNC1718232,
SNC1728349,
SNC1002505,
SNC1442287,
SNC1144256,
SNC1440905,
SNC1439444,
SNC1437970,
SNC10292667,
SNC1033474,
SNC1097738,
SNC1101445,
SNC1161566,
SNC1457940,
SNC1516109,
SNC1713941,
SNC1723211,
SNC653144,
SNC653148,
SNC676493