IR 05000275/2023004
| ML24031A118 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 02/09/2024 |
| From: | Patricia Vossmar NRC/RGN-IV/DORS/PBA |
| To: | Gerfen P Pacific Gas & Electric Co |
| References | |
| IR 2023004 | |
| Download: ML24031A118 (1) | |
Text
February 09, 2024
SUBJECT:
DIABLO CANYON POWER PLANT, UNITS 1 AND 2 - INTEGRATED INSPECTION REPORT 05000275/2023004 AND 05000323/2023004
Dear Paula A. Gerfen:
On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Diablo Canyon Power Plant, Units 1 and 2. On January 11, 2024, the NRC inspectors discussed the results of this inspection with Adam Peck, Site Vice President, and other members of your staff. The results of this inspection are documented in the enclosed report.
Two findings of very low safety significance (Green) are documented in this report. Both of these findings involved violations of NRC requirements. One Severity Level IV violation without an associated finding is documented in this report. We are treating these violations as non-cited violations (NCVs) consistent with section 2.3.2 of the Enforcement Policy.
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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Diablo Canyon Power Plant, Units 1 and 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 IV; and the NRC Resident Inspector at Diablo Canyon Power Plant, Units 1 and 2. This letter, its enclosure, and your response (if any) will be made available for public inspection and copying at http://www.nrc.gov/reading-rm/adams.html and at the NRC Public Document Room in accordance with Title 10 of the Code of Federal Regulations 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely, Patricia J. Vossmar, Chief Reactor Projects Branch A Division of Operating Reactor Safety Docket Nos. 05000275, 05000323 License Nos. DPR-80, DPR-82
Enclosure:
As stated
Inspection Report
Docket Nos.
05000275, 05000323
License Nos.
Report Nos.
05000275/2023004 and 05000323/2023004
Enterprise Identifier:
I-2023-004-0006
Licensee:
Pacific Gas and Electric Company
Facility:
Diablo Canyon Power Plant, Units 1 and 2
Location:
Avila Beach, CA
Inspection Dates:
October 1, 2023, to December 31, 2023
Inspectors:
A. Athar, Resident Inspector
B. Baca, Senior Health Physicist
D. Dodson, Senior Reactor Inspector
J. Drake, Senior Reactor Inspector
N. Greene, Senior Health Physicist
M. Hayes, Senior Resident Inspector
N. Hernandez, Operations Engineer
C. Smith, Senior Reactor Inspector
Approved By:
Patricia J. Vossmar, Chief
Reactor Projects Branch A
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Diablo Canyon Power Plant, Units 1 and 2, in accordance with the Reactor Oversight Process. The Reactor Oversight Process is the NRCs program for overseeing the safe operation of commercial nuclear power reactors. Refer to https://www.nrc.gov/reactors/operating/oversight.html for more information.
List of Findings and Violations
Failure to Identify and Control Potential Seismically-Induced System Interaction Hazards Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000275,05000323/2023004-01 Open/Closed
[H.9] - Training 71111.20 The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to adequately implement Procedure AD4.ID3, SISIP Housekeeping Activities. Specifically, on November 5, 2023, the licensee failed to identify and control potential seismically-induced system interaction hazards as required by the procedure.
Inappropriate Recategorization of Residual Heat Removal System Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000275,05000323/2023004-02 Open/Closed
[P.2] -
Evaluation 71152A The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.55a(f), Preservice and Inservice Testing Requirements, for the licensees failure to properly categorize valves RHR-2-8701 and RHR-2-8702, Unit 2 reactor coolant system Loop 4 hot leg to residual heat removal pump suction header valves at the Diablo Canyon Power Plant as Active valves in accordance with ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants. Specifically, on March 23, 2023, the licensee inappropriately removed the valves active function from the inservice testing program at Diablo Canyon Power Plant and recategorized the valves as Passive valves.
Failure to Submit and Receive Prior Authorization for a 10 CFR 50.55a(z) Alternative to ASME OM Code Post Maintenance Testing Requirements Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000275,05000323/2023004-03 Open/Closed Not Applicable 71152A The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.55a(z),
Alternatives to codes and standards requirements, for the licensees failure to submit and obtain authorization prior to implementation of an alternative to post-maintenance testing requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). Specifically, since
November 23, 2022, after valve RHR-2-8702, Unit 2 reactor coolant system Loop 4 hot leg to residual heat removal pump suction header valve, had undergone maintenance to adjust stem packing which could affect the valves performance, the licensee failed to assure that valve performance was reconfirmed by an inservice test before it was returned to service, and an alternative request under 10 CFR 50.55a(z) was not submitted and authorized prior to implementation.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000323/2022004-02 Residual Heat Removal Suction Isolation Valve Leakage 71152A Closed
PLANT STATUS
Unit1 began the inspection period shutdown in planned refueling outage24. On November12,2023, the unit began a controlled power ascension and returned to full power on November18,2023. On December8,2023, the unit began a controlled downpower in preparation for a planned maintenance outage and shut down the reactor on December9, 2023.
On December16,2023, the unit began a controlled power ascension and returned to full power on December17,2023. On December29,2023, the unit reduced power to approximately 50percent in anticipation of high ocean swells. The unit returned to full power on December31,2023, and remained at or near rated thermal power for the remainder of the inspection period.
Unit2 began the inspection period at rated thermal power. On October 20, 2023, Unit 2 began a controlled downpower to approximately 20 percent power to repair a lightning arrestor on main bank transformer phase A. The unit returned to full power on October 22, 2023. On November 10, 2023, the unit began a downpower to approximately 45 percent power for condenser in-leakage repairs and circulating water tunnel cleaning. The unit returned to full power on November 17, 2023. On December29,2023, the unit reduced power to approximately 50percent in anticipation of high ocean swells. The unit returned to full power on December30,2023, and remained at or near rated thermal power 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 onsite portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
REACTOR SAFETY
71111.04 - Equipment Alignment
Partial Walkdown Sample (IP Section 03.01) (2 Samples)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Unit 1 reactor vessel refueling level indication system on October 5, 2023
- (2) Unit 1 component cooling water pump 1-1 while component cooling water pump 1-2 was out of service for planned maintenance on December 27, 2023
71111.05 - Fire Protection
Fire Area Walkdown and Inspection Sample (IP Section 03.01) (5 Samples)
The inspectors evaluated the implementation of the fire protection program by conducting a walkdown and performing a review to verify program compliance, equipment functionality, material condition, and operational readiness of the following fire areas:
- (1) Unit 1 containment, 140-foot elevation, on October 6, 2023
- (2) Unit 1 containment, 91-foot elevation, on October 12, 2023
- (3) Unit 1 containment, 117-foot elevation, on October 18, 2023
- (4) Units 1 and 2 radiological control area, 64-foot elevation, on November 24, 2023
- (5) Units 1 and 2 radiological control area, 115-foot elevation, on November 24, 2023
71111.06 - Flood Protection Measures
Flooding Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated external flooding mitigation protections on the auxiliary building and fuel handling building roofs on December 21, 2023
===71111.08P - Inservice Inspection Activities (PWR) The inspectors verified that the reactor coolant system boundary, reactor vessel internals, risk significant piping system boundaries, and containment boundary are appropriately monitored for degradation and that repairs and replacements were appropriately fabricated, examined, and accepted by reviewing the following activities from October 9 to November 30, 2023.
PWR Inservice Inspection Activities Sample - Nondestructive Examination and Welding Activities (IP Section 03.01)===
The inspectors verified that the following nondestructive examination and welding activities were performed appropriately:
- (1) Ultrasonic Examination
- pressurizer spray line, WIB-376, pipe to elbow
- pressurizer spray line, WIB-377, elbow to pipe
- pressurizer spray line, WIB-378, elbow to pipe Visual 1 Examination
- reactor coolant, RV-8010A, Visual-1, studs and nuts Welding Activities
- gas tungsten arc welding
- main steam, DC-1-04-P-VR-MS-1-RV-20, steam generator flash tank relief to main steam, FW-1, FW-2 PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection Activities (IP Section 03.02)
No bare metal visual inspection of the reactor vessel upper head penetrations was required this outage.
PWR Inservice Inspection Activities Sample - Boric Acid Corrosion Control Inspection Activities (IP Section 03.03) (1 Sample)
The inspectors verified the licensee is managing the boric acid corrosion control program through a review of the following evaluations:
(1)
===51147485, 1-PT-474
- 51147487, SI-1-99
- 51147488, CVCS-2-8362A
- 51147501, 1R23 BA walkdowns
- 51147502, CVCS-1-90
- 51147503, LWS-1-44
- 51147504, CS-1-37
- 51147506, CVCS-1-571
- 51147507, CVCS-1-8354C
- 51147508, CVCS-1-603
- 51147509, CVCS-1-8147 PWR Inservice Inspection Activities Sample - Steam Generator Tube Inspection Activities (Section 03.04)===
The inspectors verified that the licensee is monitoring the steam generator tube integrity appropriately through a review of the following examinations:
(1)
- The primary side inspection included all tubes in steam generators 1, 2, 3, and 4, consisting of 100 percent full length bobbin, 100 percent +Point probe inspection of the hot leg top of tube-sheet, and additional inspections of dents/dings and special interest examinations.
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (3 Samples)
- (1) The inspectors observed and evaluated licensed operator performance in the control room during Unit 1 shutdown for refueling outage 1R24 on October 1, 2023.
- (2) The inspectors observed and evaluated licensed operator performance in the control room during the Unit 1 infrequently performed test of engineered safeguards equipment on October 3, 2023.
- (3) The inspectors observed and evaluated licensed operator performance in the control room during Unit 1 startup after refueling outage 1R24 on November 12, 2023.
Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)
- (1) The inspectors observed and evaluated a crew of licensed operators during simulator training on November 28, 2023.
71111.12 - Maintenance Effectiveness
Maintenance Effectiveness (IP Section 03.01) (2 Samples)
The inspectors evaluated the effectiveness of maintenance to ensure the following structures, systems, and components (SSCs) remain capable of performing their intended function:
- (1) Unit 1 pipe snubbers on October 26, 2023
- (2) Units 1 and 2 emergency lighting on December 28, 2023
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 1 Yellow risk due to reduced inventory and one offsite power source on October 5, 2023.
- (2) Unit 1 Yellow outage risk due to lowered reactor coolant system inventory with startup power cleared on October 30, 2023
- (3) Unit 1 forced outage for replacement of pressurizer safety valve 8010B on December 11, 2023
- (4) Unit 2 auxiliary saltwater 2-2 motor replacement on December 13, 2023
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 03.01) (8 Samples)
The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:
- (1) Unit 1 condensate water storage tank foreign material intrusion on October 2, 2023
- (3) Unit 1 diesel fuel oil storage tanks during cleaning on October 10, 2023
- (4) Unit 1 high unseating thrust for valve 1-8809B on October 19, 2023
- (5) Unit 2 emergency diesel generator 2-3 torque checking of special bolts at fuel header joint not performed on October 25, 2023
- (6) Unit 1 high energy line break door in penetration area with degraded molding (door 348) on October 28, 2023
- (7) Unit 2 emergency diesel generator 2-1 lube oil leak on November 29, 2023
- (8) Unit 1 steam leak upstream of main feedwater valve FCV-441 on December 11, 2023
71111.18 - Plant Modifications
Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (2 Samples)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Unit 1 condensate storage tank 1-1 temporary modification to stabilize floating platform on October 10, 2023
- (2) Unit 1 component cooling water comprehensive pump test procedure change on November 8, 2023
71111.20 - Refueling and Other Outage Activities
Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated refueling outage 1R24 activities from October 1, 2023, to November 12, 2023. This completes the sample that was partially completed and documented in Inspection Report 05000275/2023003 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML23313A088).
71111.24 - Testing and Maintenance of Equipment Important to Risk
The inspectors evaluated the following testing and maintenance activities to verify system operability and/or functionality:
Post-Maintenance Testing (PMT) (IP Section 03.01) (8 Samples)
- (1) Unit 1 valve SI-1-8923B following valve repack on October 17, 2023
- (2) Unit 1 partial integrated safeguards testing following relay replacement on October 18, 2023
- (3) Unit 2 main bank transformer phase A power factor testing after lightning arrester replacement on October 21, 2023
- (4) Unit 1 penetration 71 containment isolation valve leak testing after relief valve maintenance on October 23, 2023
- (5) Unit 1 containment fan cooler 1-3 after motor replacement on November 4, 2023
- (6) Unit 1 auxiliary building ventilation fan S-31 after emergent work for belt replacement on November 20, 2023
- (7) Unit 2 auxiliary saltwater 2-2 after motor replacement on December 20, 2023
- (8) Unit 1 pressurizer safety valve 8010B replacement on December 26, 2023
Surveillance Testing (IP Section 03.01) (5 Samples)
- (1) Unit 1 surveillance test procedure M-13 series 4 KV auto transfer testing on October 2, 2023
- (2) Unit 1 containment hatch closure on October 3, 2023
- (3) Unit 1 train B solid-state protection system testing on October 15, 2023
- (4) Unit 1, Surveillance Test Procedure M-15, "Integrated Test of Engineering Safeguards and Diesel Generator," on November 7, 2023
- (5) Unit 1 component cooling water valve testing on November 9, 2023
Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)
- (1) Unit 1 penetrations 19, 52E, 52F, 78A, 78B, and 51C containment isolation valve leak testing on November 3,
RADIATION SAFETY
71124.01 - Radiological Hazard Assessment and Exposure Controls
Radiological Hazard Assessment (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated how the licensee identifies the magnitude and extent of radiation levels and the concentrations and quantities of radioactive materials and how the licensee assesses radiological hazards.
Instructions to Workers (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee instructs workers on plant-related radiological hazards and the radiation protection requirements intended to protect workers from those hazards.
Contamination and Radioactive Material Control (IP Section 03.03) (3 Samples)
The inspectors observed/evaluated the following licensee processes for monitoring and controlling contamination and radioactive material:
(1)auxiliary building personnel monitoring and release of material from the radiologically controlled area
- (2) Unit 1 containment access point contamination area egress, release of material from a contamination area, and personnel monitoring (3)radiation protection technician control of Unit 1 steam generators 1, 2, 3, and 4's radioactive material, surface contamination, and potential airborne radioactivity areas
Radiological Hazards Control and Work Coverage (IP Section 03.04) (3 Samples)
The inspectors evaluated the licensee's control of radiological hazards for the following radiological work:
- (1) Radiation Work Permit (RWP) 23-0042, "Radioactive Waste Activities," which involved spent resin transfer from two demineralizer beds
- (2) RWP 23-1040, "1R24 Steam Generator Work (Primary Side)" which included manway (strong back) removal, insert removal, FME installation and eddy current testing equipment installation
- (3) RWP 23-1042, "Safety Injection check valve 8948B " involving maintenance of the 8948B check valve High Radiation Area and Very High Radiation Area Controls (IP Section 03.05) (4 Samples)
The inspectors evaluated licensee controls of the following high radiation areas (HRAs) and very high radiation areas (VHRAs):
(1)auxiliary building spent resin storage tank filter gallery 115-foot elevation (locked high radiation area [LHRA]) and spent resin storage tank filters 115-foot elevation (HRA)
- (2) Unit 1 fuel transfer tube area (VHRA)
- (4) Unit 1 steam generators 1, 2, 3, 4 primary side access (HRA and LHRA)
Radiation Worker Performance and Radiation Protection Technician Proficiency (IP
Section 03.06) (1 Sample)
- (1) The inspectors evaluated radiation worker and radiation protection technician performance as it pertains to radiation protection requirements.
71124.04 - Occupational Dose Assessment
Source Term Characterization (IP Section 03.01) (1 Sample)
- (1) The inspectors evaluated licensee performance as it pertains to radioactive source term characterization.
External Dosimetry (IP Section 03.02) (1 Sample)
- (1) The inspectors evaluated how the licensee processes, stores, and uses external dosimetry.
Internal Dosimetry (IP Section 03.03) (3 Samples)
The inspectors evaluated the following internal dose assessments:
- (1) SAPN 51149931, dose assessment for worker 4371, dated April 13, 2022
- (2) SAPN 51149246, dose assessment for worker 4373, dated April 8, 2022 (3)dose assessments for workers 4372 and 4373, dated April 1, 2022
Special Dosimetric Situations (IP Section 03.04) (2 Samples)
The inspectors evaluated the following special dosimetric situations:
- (1) NRC Form 5 and information for declared pregnant workers
- (2) NRC Form 5 and assessments for four workers using EDEX for non-uniform radiation fields
71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, and
Transportation
Shipment Preparation (IP Section 03.04)
- (1) The inspectors observed the preparation of radioactive shipment RMS-23-072 of limited quantity radioactive material (Reactor Coolant Liquid U2 RCS B-10/B-11) on October 12,
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 (October 1, 2022, through September 30, 2023)
- (2) Unit 2 (October 1, 2022, through September 30, 2023)
BI02: RCS Leak Rate Sample (IP Section 02.11) (2 Samples)
- (1) Unit 1 (October 1, 2022, through September 30, 2023)
- (2) Unit 2 (October 1, 2022, through September 30, 2023)
OR01: Occupational Exposure Control Effectiveness Sample (IP Section 02.15) (1 Sample)
- (1) April 1, 2022, through September 30, 2023 PR01: Radiological Effluent Technical Specifications/Offsite Dose Calculation Manual Radiological Effluent Occurrences (RETS/ODCM) Radiological Effluent Occurrences Sample (IP Section 02.16) (1 Sample)
- (1) April 1, 2022, through September 30, 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)auxiliary salt-water pump 2-2 multiple discolored bearing oil samples on November 29, 2023
71152S - Semiannual Trend Problem Identification and Resolution Semiannual Trend Review (Section 03.02)
- (1) The inspectors reviewed the licensees corrective action program for potential adverse trends in housekeeping that might be indicative of a more significant safety issue. The inspectors observation associated with this review is documented in the inspection results section of this report.
INSPECTION RESULTS
Failure to Identify and Control Potential Seismically-Induced System Interaction Hazards Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000275,05000323/2023004-01 Open/Closed
[H.9] - Training 71111.20 The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to adequately implement Procedure AD4.ID3, SISIP Housekeeping Activities. Specifically, on November 5, 2023, the licensee failed to identify and control potential seismically-induced system interaction hazards as required by the procedure.
Description:
The inspectors identified a failure to follow the requirements of Procedure AD4.ID3, SISIP Housekeeping Activities, revision 18. The seismically-induced system interaction program (SISIP) ensures targets recognized as safe shutdown systems, structures, and components, as well as certain sensitive accident-mitigating systems, do not have seismic interferences such that they function properly during and following an earthquake. The program also requires plant workers remain aware of conditions of sources in the plant capable of potential seismic threats to plant equipment caused by transient plant components. Effective implementation of the SISIP program relies on plant workers identifying and documenting seismically-induced system interaction conditions to ensure they are promptly corrected, and these conditions receive the appropriate engineering evaluations.
On November 5, 2023, the inspectors identified an unsecured 55-gallon barrel located in the corridor between the residual heat removal (RHR) pump rooms on Unit 1. The inspectors reviewed the stations Seismically-Induced System Interaction (SISI) Manual and noted that Appendix 1, figure 05, describes the corridor between RHR pump rooms as being an area which includes SISI targets such as manual valves, piping, instrumentation, instrumentation tubing, electrical devices, and electrical raceway. As such, in accordance with the licensee Procedure AD4.ID3, SISIP Housekeeping Activities, it is unacceptable for large barrels to be unrestrained in the corridor between the RHR pump rooms. To restore compliance with the licensees SISIP, the licensee immediately removed the barrel from the area. The station evaluated this issue and determined that it was considered a Seismic Concern.
Corrective Actions: Upon identification of this issue, the licensee promptly removed the identified SISI source from the SISI area to restore compliance with the licensees SISIP.
Corrective Action References: Notification 51213199
Performance Assessment:
Performance Deficiency: The licensees failure to implement a required procedure for identifying and controlling seismically-induced system interaction hazards is a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors 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 unrestrained 55-gallon barrel in the corridor between the Unit 1 RHR pump rooms could leave components such as manual valves, piping, instrumentation, and electrical devices vulnerable to a seismically-induced system interaction if a seismic event were to occur while the unrestrained barrels were present in the area. Furthermore, this issue was similar to IMC 0612, Appendix E, Examples of Minor Issues, section 4, example
- (a) of a more than minor issue because the licensees SISI manual, as written, would never allow storage of transient equipment in the vicinity of SISI targets.
Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix G, Shutdown Safety SDP. The finding screened as having very low safety significance (Green) because, although it was a deficiency affecting the seismic qualification of the RHR system, the RHR system maintained its operability.
Cross-Cutting Aspect: H.9 - Training: The organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee determined that the refueling outage that was in progress may have contributed to workers failing to observe SISIP housekeeping requirements, as many supplemental workers were onsite for the refueling outage.
Enforcement:
Violation: Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part that activities affecting quality shall be described by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings.
The licensee established quality-related station Procedure AD4.ID3, SISIP Housekeeping Activities, as the implementing procedure for controlling potential seismically-induced system interaction hazards. Step 5.1.2 of station procedure AD4.ID3 states that, Transient equipment introduced into the plant shall not create a potential SISI threat or SISI housekeeping concern. The corridor between the RHR pump rooms is an area which includes SISI targets, and it is unacceptable for large barrels to be unrestrained in that location.
Contrary to the above, on November 5, 2023, for SISIP Housekeeping, an activity affecting quality, the licensee failed to prevent transient equipment from being introduced into a SISI area. The licensee failed to accomplish the activity affecting quality in accordance with a documented procedure. Specifically, the licensee failed to implement the SISIP to ensure that transient equipment introduced into the plant did not create a potential SISI concern.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
Inappropriate Recategorization of Residual Heat Removal System Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000275,05000323/2023004-02 Open/Closed
[P.2] -
Evaluation 71152A The inspectors identified a Green finding and associated non-cited violation of 10 CFR 50.55a(f), Preservice and Inservice Testing Requirements, for the licensees failure to properly categorize valves RHR-2-8701 and RHR-2-8702, Unit 2 reactor coolant system Loop 4 hot leg to residual heat removal pump suction header valves at the Diablo Canyon Power Plant as Active valves in accordance with ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code), Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants. Specifically, on March 23, 2023, the licensee inappropriately removed the valves active function from the inservice testing program at Diablo Canyon Power Plant and recategorized the valves as Passive valves.
Description:
As documented in NRC Integrated Inspection Report 05000275/2022004 and 05000323/2022004 (ML23031A137), the inspectors identified an unresolved item (URI 05000323/2022004-02) related to the licensees deferral of a post-maintenance test on valve RHR-2-8702, Unit 2 reactor coolant system (RCS) Loop 4 hot leg to residual heat removal (RHR) pump suction header valve.
The Diablo Canyon Power Plant (DCPP) inservice testing (IST) Program Plan dated March 2, 2016, implementing ASME OM Code (2004 Edition through 2006 Addenda),
Subsection ISTC, paragraphs ISTC-3200(a) and ISTC-3110, as incorporated by referenced in 10 CFR 50.55a, designated valve RHR-2-8702 as an Active Category A valve with applicable ASME OM Code requirements, including exercise stroke time testing, position indication verification testing, and seat leakage testing.
On March 23, 2023, the DCPP licensee updated the IST Program Plan, changing the classification of valves 8701 and 8702 from Active to Passive.
The inspectors reviewed the DCPP Final Safety Analysis Report (FSAR), and noted that FSAR Table 3.9-9, Active Valves, includes valves 8701 and 8702 as Operable in the Safe Shutdown column. Footnote
- (a) to the Safe Shutdown column states:
The valves whose positions are listed in this column are those valves whose operability is relied on to perform an active function such as safe shutdown of the reactor or mitigation of the consequences of a Design-Basis Accident coincidental with loss of offsite power. An entry of "functional" or equivalently "operable" means that the valve must be capable of being opened and/or closed to perform its active function. For DCPP, safe shutdown is defined as Mode 3 following an accident (SSER 7 and SSER 22), Mode 5 following a Hosgri earthquake, and Mode 3, following a fire (10 CFR 50.48(c)).
For valves 8701 and 8702, FSAR Table 3.9-9 specifies in Footnote 23, Valve does not have an active safety function to support accident mitigation or Mode 3 safe shutdown. Valve is active to support achieving post-Hosgri cold shutdown. Valve needs to be seismically qualified for active function for Hosgri only. Post-Hosgri refers to a licensing requirement put in place during initial licensing related to a design-basis seismic event originating at the Hosgri fault line.
ASME OM Code, Subsection ISTA, General Requirements, Paragraph ISTA-1100, Scope, states that these requirements apply to
- (a) pumps and valves that are required to perform a specific function in shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident. The inspectors noted that FSAR Table 3.9-9 included valves 8701 and 8702 as having an active function in shutting down a reactor to the safe shutdown condition of Mode 5 following a Hosgri event, because these valves need to open to start the shutdown cooling mode of the RHR system.
The inspectors also noted that these valves had active functions listed in the FSAR dating back to initial licensing.
Since, section 2.3 in NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants, states that the IST program scope should be developed to be consistent with the FSAR, the inspectors were concerned that valves 8701 and 8702 should be scoped into the DCPP IST Program with respect to their active functions following a Hosgri earthquake.
The inspectors performed a review of the design and licensing basis to gain further context on the function of these valves. The availability of valves 8701 and 8702 for responding to a Hosgri earthquake was relied upon during the initial NRC licensing of DCPP. For example, supplement 8 to NRC safety evaluation report (SER) for DCPP states in Chapter 1.0, Introduction, that the principal outstanding issue in the NRC review of the operating license application has been the earthquake capabilities of the Hosgri fault and its impact on seismic considerations for DCPP. Section 3.2.1, Seismic Classification, in SER, Chapter 3.0, Design Criteria - Structures, Components, Equipment, and Systems, states that in SER, supplement 7, the NRC provided its evaluation of the plants capability for achieving long-term cold shutdown conditions after an earthquake using only redundant equipment qualified for the Hosgri event. The NRC staff stated that the plants capability was acceptable considering resolution of outstanding items in the review. SER, supplement 7, discussed the capability of the RHR system (including valves 8701 and 8702) to bring the plant to cold shutdown following a Hosgri event.
Additionally, NRC SER, supplement 8, Chapter 18.0, Review by the Advisory Committee on Reactor Safeguards [ACRS], summarizes the ACRS review of the licensing of DCPP, including the capability to respond to a Hosgri earthquake. In discussing the Hosgri response, the ACRS letter dated July 14, 1978, references the requirement for the capability to bring DCPP to cold shutdown (Mode 5) using only safety-grade equipment. Additionally, regarding the question of whether the Hosgri earthquake was a design-basis event, NRC letter dated October 12, 2012 (ML120730106), Diablo Canyon Power Plant, Units Nos. 1 and 2 - NRC Review of Shoreline Fault, describes the three design-basis earthquakes used to develop the seismic qualification basis for structures, systems, and components at DCPP. For the Hosgri Earthquake (HE), the NRC letter states that Equipment credited in the HE shutdown path is required to remain functional following a Hosgri design basis earthquake. The NRC letter also states that DCPP has three earthquake scenarios (DE, DDE, and HE) in its design and licensing basis.
As a result, the inspectors concluded that the licensee inappropriately changed the classification of valves 8701 and 8702 in the IST Program from Active to Passive, because the Hosgri earthquake is part of the original licensing and design-basis, the IST program is required to be consistent with the design and licensing basis, and valves 8701 and 8702 are required to open to transition the plant to Mode 5 following a Hosgri earthquake.
Corrective Actions: The licensee has entered this issue into their corrective action program and is evaluating actions to restore compliance.
Corrective Action References: Notification 51176817
Performance Assessment:
Performance Deficiency: The failure to correctly classify safety-related valves RHR-2-8701 and RHR-2-8702 as Active valves in accordance with 10 CFR 50.55a(f) 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, inappropriately recategorizing valves RHR-2-8702 and RHR-2-8701 as Passive valves adversely affected the reliability of the RHR system by removing the applicable ASME OM Code requirements, including exercise stroke time testing, position indication verification testing, and seat leakage testing.
Significance: The inspectors assessed the significance of the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting the design or qualification of the system, did not represent a loss of the PRA function of a single train TS system for greater than its TS-allowed outage time, did not represent a loss of PRA function of two separate TS systems for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, did not represent a loss of a PRA system and/or function as defined in the PRIB or licensee's PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, did not represent a loss of the PRA function of one or more non-TS trains of equipment designated as risk significant in accordance with the licensee's maintenance rule program for more than 3 days, and did not involve the loss or degradation of equipment specifically designed to mitigate a seismic event for greater than 14 days.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the station did not recognize that the FSAR definition of safe shutdown additionally includes achieving Mode 5 after a Hosgri earthquake, which led to inappropriately changing the classification of valves RHR-2-8701 and RHR-2-8702.
Enforcement:
Violation: Title 10 CFR 50.55a(f)(4) requires, in part that throughout the service life of a boiling or pressurized-water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements set forth in the ASME OM Code, as incorporated by reference in 10 CFR 50.55a.
Contrary to the above, from March 23, 2023, to present, the licensee failed to ensure that pumps and valves that were within the scope of the ASME OM Code met the IST requirements set forth in the ASME OM Code. Specifically, the licensee failed to categorize valves RHR-2-8701 and RHR-2-8702 as Active valves in accordance with the ASME OM Code, as incorporated by reference in 10 CFR 50.55a. When the licensee recategorized these valves as Passive valves, the applicable ASME OM Code requirements were removed without proper authorization.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
The disposition of this finding and associated violation closes URI: 05000323/2022004-02.
Failure to Submit and Receive Prior Authorization for a 10 CFR 50.55a(z) Alternative to ASME OM Code Post Maintenance Testing Requirements Cornerstone Severity Cross-Cutting Aspect Report Section Not Applicable Severity Level IV NCV 05000275,05000323/2023004-03 Open/Closed Not Applicable 71152A The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.55a(z),
Alternatives to codes and standards requirements, for the licensees failure to submit and obtain authorization prior to implementation of an alternative to post-maintenance testing requirements of the American Society of Mechanical Engineers (ASME) Code for Operation and Maintenance of Nuclear Power Plants (OM Code). Specifically, since November 23, 2022, after valve RHR-2-8702, Unit 2 reactor coolant system Loop 4 hot leg to residual heat removal pump suction header valve, had undergone maintenance to adjust stem packing which could affect the valves performance, the licensee failed to assure that valve performance was reconfirmed by an inservice test before it was returned to service, and an alternative request under 10 CFR 50.55a(z) was not submitted and authorized prior to implementation.
Description:
As documented in NRC Integrated Inspection Report 05000275/2022004 and 05000323/2022004 (ML23031A137), the inspectors identified an unresolved item (URI 05000323/2022004-02) related to the licensees deferral of a post-maintenance test on valve RHR-2-8702, Unit 2 reactor coolant system (RCS) Loop 4 hot leg to residual heat removal (RHR) pump suction header valve.
In April 2021, a wet boric acid valve packing leak was identified at the follower for valve RHR-2-8702, Unit 2 RCS Loop 4 hot leg to RHR pump suction header valve at DCPP. The boric acid leak was cleaned and the packing retorqued, reducing, but not stopping the leakage. In November 2022, valve RHR-2-8702 was cycled prior to plant restart and the leakage through its lower packing set increased. This resulted in unacceptable leakage to the pressurizer relief tank through the valve packing leak-offline. Detailed inspection of the stem for valve RHR-2-8702 identified damage in the form of linear galling, including areas of raised metal, on the portion of the stem that passes through the valve packing. The licensee believed that this galling damaged the valve packing, thereby causing the observed packing leakage. A repack of the valve while in Mode 5 was performed, with the valve open, for the purpose of eliminating leakage to the pressurizer relief tank. The licensee successfully stroked closed valve RHR-2-8702 following repacking.
Because the maintenance involved a motor-operated valve (MOV), station Procedure AD13.ID4, Post Maintenance Testing, revision 32, Attachment 3, MOV Post-Maintenance Testing Requirements, provides the standard post-maintenance testing activities. The required testing following valve repacking would include: an exercise stroke time test, an inservice leak test, an MOV static diagnostic test, and a seat leakage test. The licensee completed the seat leakage test and the inservice leak check prior to returning Unit 2 to power operations to demonstrate the leakage safety function of valve RHR-2-8702 was maintained. Valve RHR-2-8702 isolates the low-pressure RHR system from the high-pressure RCS to prevent an intersystem loss of coolant accident (LOCA). As such, this valve is maintained in the closed position with power removed from the valves motor operator during Modes 1, 2, and 3 per technical specification 3.5.2. Furthermore, valve RHR-2-8702 is an RCS pressure isolation valve per technical specification 3.4.14, which requires that this valve be leak tested each refueling outage. Thus, the successful performance of the seat leakage test and the inservice leak check demonstrated that the closed safety function of valve RHR-2-8702 was maintained. However, the licensee chose to defer the portion of the post-maintenance test that would demonstrate other functions important to safety. The licensee chose to defer stroke time testing and MOV static diagnostic testing because of the concern that exercising the valve could cause further damage to the packing material by the damaged stem, leading to leakage during the subsequent cycle.
Valve RHR-2-8702 is included on the List of Active Valves in the DCPP FSAR, Table 3.9-9.
This table specifies in NOTE 23, Valve does not have an active safety function to support accident mitigation or Mode 3 safe shutdown. Valve is active to support achieving post-Hosgri cold shutdown. Valve needs to be seismically qualified for active function for Hosgri only.
Post-Hosgri refers to a portion of the licensing basis related to a seismic event originating at the Hosgri fault line. Thus, the opening of valve RHR-2-8702 is a licensing basis function required to achieve cold shutdown, only after a Hosgri earthquake. Calculation DCM S-10, section 4.3.3.3, states that the ability to achieve cold shutdown (Mode 5) with no concurrent accident following a Hosgri earthquake is a licensing commitment. This function would be demonstrated via an exercise stroke time test and an MOV static diagnostic test. The licensee completed an evaluation to support deferment of the exercise stroke time test and the MOV static diagnostic test until refueling outage 2R24, scheduled for April 2024.
The current DCPP IST Program Plan dated March 2, 2016, implementing ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) (2004 Edition through 2006 Addenda), Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, paragraphs ISTC-3200(a) and ISTC-3110, specifies that valve RHR-2-8702 is an Active Category A valve with applicable ASME OM Code requirements, including exercise stroke time testing, position indication verification testing, and seat leakage testing. On November 24, 2022, the licensee provided the inspectors with an evaluation stating that the ability of valve RHR-2-8702 to perform its post-Hosgri opening function is supported by the actuator selected for this application: the Limitorque SMB-1 actuator and 40 ft-lb motor, which the licensee believes has adequate margin to accommodate increases in running loads which may occur with variations in packing friction. The licensees evaluation did not include the specific assumptions used in its calculation of the operating capability of valve RHR-2-8702 to overcome the galling of the valve stem.
In response to inspector questions, the licensee noted that deferral of post-maintenance testing (PMT) is allowed for valve stem packing adjustments based on NUREG-1482, Guidelines for Inservice Testing at Nuclear Power Plants. The licensee stated that a similar approach was used to demonstrate that PMT involving valve cycling may be deferred for replacement of packing material. The licensees evaluation quoted a portion of section 4.4.2 of NUREG-1482 related to packing adjustments and backseating. However, the guidance in section 4.4.2 of NUREG-1482 is not applicable to complete replacement of valve stem packing. For example, section 4.4.2 of NUREG-1482 states that the IST requirements do not prohibit or discourage a licensee from making limited adjustments to valve packing to stop a leak that may be adversely affecting the valve or surrounding components.
Valve RHR-2-8702 is within the scope of the ASME OM Code, as incorporated by reference in Section 50.55a, Codes and standards, in 10 CFR 50.55a. As such, per 10 CFR 50.55a(f)(4), the licensee must meet the inservice test requirements set forth in the ASME Code for valve RHR-2-8702 or submit to the NRC and receive approval for an alternative request under 10 CFR 50.55a(z). After valve RHR-2-8702 underwent maintenance to adjust stem packing which could affect the valves performance, the licensee failed to assure that valve performance was reconfirmed by an inservice test before the time it was returned to service, and an alternative request under 10 CFR 50.55a(z) was not submitted and authorized prior to implementation.
The NRC did not identify any immediate safety concerns that the valve would not be capable of being opened (either remotely or manually) following a post-Hosgri event should the plant be required to be placed in a cold shutdown condition. However, after completing maintenance to repack valve RHR-2-8702, which is within the scope of the ASME OM Code, as incorporated by reference in Section 50.55a, Codes and standards, in 10 CFR 50.55a, the licensee did not complete stroke time testing and MOV static diagnostic testing, as required by the ASME OM Code, and an alternative request under 10 CFR 50.55a(z) was not submitted to the NRC and authorized prior to implementation.
Corrective Actions: The licensee plans to complete the required stroke time testing and MOV static diagnostic testing at the earliest opportunity.
Corrective Action References: Notifications 51200971 and 51115552
Performance Assessment:
The inspectors determined this violation was associated with a minor performance deficiency.
Enforcement:
The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. This violation was more than minor because there was a reasonable likelihood the change would require NRC review and approval prior to implementation.
Severity: This is a Severity Level IV violation in accordance with Sections 6.1 and 6.9 of the Enforcement Policy because it resulted in no potential safety consequences, but it is still of a more than minor concern in the NRCs Traditional Enforcement Process. The licensee successfully stroked close valve RHR-2-8702 following stem repair and repacking. The NRC did not identify any immediate safety concerns that the valve would not be capable of being opened (either remotely or manually) following a post-Hosgri event should the plant be required to be placed in a cold shutdown condition.
Violation: Title 10 CFR 50.55a(z), Alternatives to codes and standards requirements, requires, in part, Alternatives to the requirements of paragraphs
- (b) through
- (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation.
Title 10 CFR 50.55a(f)(4) requires, in part that throughout the service life of a boiling or pressurized-water-cooled nuclear power facility, pumps and valves that are within the scope of the ASME OM Code must meet the inservice test requirements set forth in the ASME OM Code. ASME OM Code (2004 Edition through the 2006 Addenda), Subsection ISTC, paragraph ISTC-3310, Effects of Valve Repair, Replacement, or Maintenance on Reference Values, states, in part, the following:
When a valve or its control system has been replaced, repaired, or has undergone maintenance [Footnote 1] that could affect the valves performance, a new reference value shall be determined, or the previous value reconfirmed by an inservice test run before the time it is returned to service or immediately if not removed from service.
Footnote 1: Adjustment of stem packing, limit switches, or control system valves, and removal of the bonnet, stem assembly, actuator, obturator, or control system components are examples of maintenance that could affect valve performance parameters.
Contrary to the above, since November 23, 2022, after valve RHR-2-8702 had undergone maintenance to adjust stem packing which could affect the valves performance, the licensee failed to assure that valve performance was reconfirmed by an inservice test run before the time it was returned to service, and an alternative request under 10 CFR 50.55a(z) was not submitted and authorized prior to implementation.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with section 2.3.2 of the Enforcement Policy.
The disposition of this violation closes URI: 05000323/2022004-02.
Observation: Turbine Deck Floor Loading During Refueling Outage 71152S On September 27, 2023, the inspectors walked down the 140-foot turbine deck and observed materials stored in a restricted loading area. The licensee had evaluated and approved temporary storage for several items within the restricted loading area, but four of the staged items had not previously been evaluated and approved for temporary storage within the restricted loading area. The licensee performed a subsequent evaluation and determined that the additional items that were placed within the restricted loading area were acceptable and did not overload the turbine deck.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On October 17, 2023, the inspectors presented the occupational radiation safety inspection results to Adam Peck, Site Vice President, and other members of the licensee staff.
- On November 30, 2023, the inspectors presented the Diablo Canyon Unit 1 inservice inspection results to Paula Gerfen, Senior Vice President, Generation and Chief Nuclear Officer, and other members of the licensee staff.
- On January 11, 2024, the inspectors presented the integrated inspection results to Adam Peck, Site Vice President, and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
OP A-2:X
RVRLIS Alignments for Refueling Outages
OP D-1: II
Auxiliary Feedwater System - Alignment Verification
for Plant Startup
Procedures
OP K-10E1
Sealed Valve Checklist for Component Cooling Water
Pump 1-1
Corrective Action
Documents
Notifications
207351, 51207352, 51207353, 51207354,
208392, 51208393, 51208394
Drawings
111805-14
Unit 1 & 2 Radiological Control Area (RCA) & H Block
115'
Drawings
111805-6
Unit 1 & 2 Radiological Control Area (RCA) 54' & 64'
Drawings
111906-15
Unit 1 & 2 Auxiliary Building 54' and 64'
Drawings
111906-21
Unit 1 & 2 Auxiliary Building 115'
Drawings
111906-22
Unit 2 Auxiliary Building 115'
Drawings
11805-35
Unit 1, Containment Building Elev. 91'
Drawings
11805-37
Unit 1, Containment Building Elev. 117'
Drawings
11805-38
Unit 1, Containment Building Elev. 140'
Corrective Action
Documents
Notifications
218543, 51219069, 51219130, 51219188,
219189, 51219260, 51219262, 51219440,
219441, 51219442, 51219511, 51219580,
220120, 51220121
Drawings
59641-1-0
Miscellaneous Exterior Details
MA1.ID17
1: Additional Guidance on Performing
Miscellaneous
Sections 3205-
207, pages 405-
406
Uniform Building Code
1967 Edition
Procedures
AWP E-016
Maintenance Rule Structural Monitoring Programs -
Civil
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
51168982, 51194888, 51182710, 51166676,
51168982, 51148075, 51148692, 51197775,
51197790, 51197926, 51197938, 51197939,
51198078, 51198080, 51198081, 51198083,
51198517, 51198602, 51198605, 51198607,
51198708, 51198995, 51198996, 51199001,
51192482, 51192483, 51192484, 51195639,
51195711, 51190851, 51190853, 51183466,
51183467, 51183468, 51183469, 51183481,
51184015, 51175456, 51176686, 51177382,
51177383, 51177663, 51178440, 51178726,
51175030, 51168816, 51169287, 51169488,
51166041, 51147485, 51154978, 51150425,
51168982, 51169288, 51150425, 51150429,
51150430, 51168451, 51167940
Corrective Action
Documents
Resulting from
Inspection
208240, 51208241, 51208242, 51208243,
208244, 51208382, 51208398, 51208399,
208412, 51208440, 51208565, 51208609,
208638, 51213160
Miscellaneous
Conco Job
- 33941
1R23 Unit 1 Component Cooling Water Heat
Exchanger CCW 1-1 April 2022
04/2022
Procedures
AD4.ID2
Plant Leakage Evaluation
Procedures
ER1.ID2
Boric Acid Corrosion Control Program
Procedures
ER1.ID2
Boric Acid Corrosion Control Program
Procedures
MA1.ID12
Control of Tools for Use on Stainless Steel
Procedures
Visible Dye Liquid Penetrant Examination Procedure
Procedures
Visual Examination of Component Surfaces
Procedures
Visual Examination During Section XI System
Pressure Test
Procedures
Visual Examination of Component and Piping
Supports
Procedures
General Visual Examination of the Containment Liner
Procedures
PDI-UT-2
PDI Generic Procedure for the Ultrasonic
Examination of Austenitic Pipe Welds
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
PDI-UT-3
PDI Generic Procedure for the Ultrasonic Through-
Wall Sizing of Planar Flaws in Similar Metal Piping
G
Procedures
STP R-8A
Reactor Coolant System Leakage Test
Procedures
STP R-8C
Containment Walkdown for Evidence of Boric Acid
Leakage
Work Orders
68066762, 60148506
Procedures
OP L-5
Plant Cooldown from Minimum Load to Cold
Shutdown
111
Procedures
STP M-15
Integrated Test of Engineered Safeguards and Diesel
Generators
Corrective Action
Documents
Notifications
206934, 51210718, 51210719
Miscellaneous
1R24 Outage Safety Plan
Miscellaneous
1X25 Outage Safety Plan
Procedures
AD8.DC55
Outage Safety Scheduling
Procedures
AD8.DC55
2: Unit 1 Outage Safety Checklist - Mode
Loops Filled
Procedures
OP AP-11
Malfunction of Component Cooling Water System
35A
Procedures
OP O-36
Protected Equipment Postings
Corrective Action
Documents
Notifications
204021, 51205821, 51206011, 51206570,
206844, 51209758, 51209818, 51209871,
210111, 51210964, 51211283, 51212456,
216861
Procedures
OM7.ID12
03/21/2023
Procedures
PEP M-21A
Diesel Fuel Oil Storage Tanks Inspection and
Cleaning
Procedures
STP M-10A
Diesel Fuel Oil Storage Tank Inventory
25A
Corrective Action
Documents
Notifications
208190
Procedures
STP P-CCW-11
Routine Surveillance Test of Component Cooling
Water Pump 1-1
Procedures
STP P-CCW-11
Routine Surveillance Test of Component Cooling
Water Pump 1-1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Procedures
STP P-CCW-12
Routine Surveillance Test of Component Cooling
Water Pump 1-2
Procedures
STP P-CCW-12
Routine Surveillance Test of Component Cooling
Water Pump 1-2
Procedures
STP P-CCW-13
Routine Surveillance Test of Component Cooling
Water Pump 1-3
Procedures
STP P-CCW-13
Routine Surveillance Test of Component Cooling
Water Pump 1-3
Corrective Action
Documents
Notifications
218381, 51218382, 51218383, 51218388,
218390, 51218391, 51218392, 51218393,
218395, 51218396, 51218397, 51218398,
218399, 51218410, 51218473, 51219386,
219556
Corrective Action
Documents
Notifications
206021, 51208755, 51208959, 51213817,
213878, 51214821, 51215560, 51218655,
218933
Procedures
MP E-53.10A
Preventive Maintenance of Limitorque Motor
Operators
Procedures
MP E-53.10V1
MOV Diagnostic Testing
Procedures
Containment Hatch Closure Test
Procedures
OP H-1: I
Auxiliary Building Safeguards Ventilation (ABVS) -
Make Available
Procedures
PEP 14-02
CCW Header B Valves Seat Leakage Test
STP I-38-B.4
SSPS Train B SI Reset Timer and Slave Relay K602
Test in MODES 5, 6, or Defueled
Procedures
STP I-83B
Pressurizer Safety Valve Relief Line Acoustic
Channels Calibration
Procedures
STP M-13B1
ENGD SFGDS Auto Timers Setting Verf Loads
Started SSPS Relay K608, Train A
Procedures
STP M-13B2
ENGD SFGDS Auto Timers Setting Verf Loads
Started SSPS Relay K608, Train B
Procedures
STP M-13B3
ENGD SFGDS Auto Timers Setting Verf Loads
Started SSPS Relay K609, Train A
Procedures
STP M-13B4
ENGD SFGDS Auto Timers Setting Verf Loads
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Started SSPS Relay K609, Train B
Procedures
STP M-13F
4KV Bus F Non-SI Auto-Transfer Test
Procedures
STP M-13G
4KV Bus G Non-SI Auto-Transfer Test
STP M-13H
4KV Bus H Non-SI Auto-Transfer Test
STP M-15
Integrated Test of Engineered Safeguards and Diesel
Generators
STP M-93A
Refueling Interval Surveillance - Containment Fan
Cooler System
Procedures
STP-P-ASW-A22
Comprehensive Test of Auxiliary Saltwater Pump 2-2
11C and 12
Procedures
STP V-619
Penetration 19 Containment Isolation Valve Leak
Testing
2A
Procedures
STP V-671
Penetration 71 Containment Isolation Valve Leak
Testing
26A
Procedures
STP V-678
Penetrations 52E, 52F, 78A, and 78B Containment
Isolation Valve Leak Testing
Procedures
STP-V651C
Penetration 51C Containment Isolation Valve Leak
Testing
Work Orders
WO 60061903, 60151715, 60154936, 60157868,
60158169, 60158266, 60159059, 60159321,
60159473, 60159761, 64121822, 64123262,
64133326, 64136165, 64146349, 64146351,
64154483, 64156562, 64218165, 64224898,
286104, 64286108, 64293225, 64293227,
293660
ALARA Plans
RWP 23-0005
Containment Entry at Power
ALARA Plans
RWP 22-1014
1R23 Upper Cavity, Decon, Cavity Entry, Head Set
04/13/2022
ALARA Plans
RWP 22-2009
2R23 Snubber Work and Inspections
ALARA Plans
RWP 22-2014
2R23 Upper Cavity Decon, Cavity Entry, Head Set
11/03/2022
ALARA Plans
RWP 23-1001
1R24 General Access to Containment
09/06/2023
ALARA Plans
RWP 23-1040
1R24 Steam Generator Work (Primary Side)
09/12/2023
Corrective Action
Documents
Notifications
51148721, 51148937, 51148938, 51148989,
51149051, 51151770, 51163436, 51163773,
51164094, 51165501, 51166062, 51166508,
51167508, 51168349, 51168908, 51169372,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
51169804, 51169879, 51173745, 51181235,
51181902, 51186505, 51186557, 51188698,
51189292, 51189965, 51192827, 51198002,
51198790
Corrective Action
Documents
Resulting from
Inspection
Notifications
208152, 51208157, 51208158
Miscellaneous
Hot Spot Trends Log
10/09/2023
Procedures
RCP D-310
RCA Access Control
Procedures
RCP D-330
Personnel Dosimetry Evaluations
Procedures
RCP NISP-RP.02
Radiation and Contamination Surveys
Procedures
RCP NISP-RP.03
Radiological Air Sampling
Procedures
RCP NISP-RP.04
Radiological Posting and Labeling
3A
Procedures
RCP NISP-RP.05
Access Controls for High Radiation Areas
Procedures
RCP NISP-RP.08
Use and Control HEPA Filtration and Vacuum
Equipment
Procedures
RCP NISP-RP.11
Radiological Protection Fundamentals
Radiation
Surveys
84152
Alpha Survey for Unit 2 Fuel Handling Building
10/03/2022
Radiation
Surveys
86795
Unit 2 CTMT Follow-up Survey
04/06/2023
Radiation
Surveys
88311
Unit 1 Turbine Building - 119' TM Instrument
Calibration Facility
09/06/2023
Radiation
Surveys
267
1R24 91' 8149 Valves, TSR# 23-0020, and Job
Coverage
10/11/2023
Self-
Assessments
Quick Hit Assessment: 71124.01 Pre-inspection on
Radiological Hazard Assessment and Exposure
Controls
08/01/2023
Self-
Assessments
22-QP-01
Quality Performance Assessment Report (QAPR)
First Period 2022
06/21/2022
Self-
Assessments
22-QP-02
Quality Performance Assessment Report (QAPR)
Second Period 2022
2/21/2022
Corrective Action
Notifications
50951907, 51149246, 51149931, 51167505,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Documents
51167508, 51168462, 51168724, 51169804,
51186557, 51197777, 51197898, 51198075
Miscellaneous
22 Dry Active Waste Radwaste Correlation Factors
07/17/2023
Miscellaneous
1R23 Alpha Sample Analysis
08/29/2022
Miscellaneous
2R23 Alpha Sample Analysis
04/27/2023
Miscellaneous
Source Term Reduction Efforts Since Cycle 20
09/13/2023
Miscellaneous
Neutron Dosimetry Evaluation at DCNPP, Technical
Support 16-090
08/20/2018
Miscellaneous
Prospective Evaluation for Shallow Dose Equivalent
and Consolidated Guidance for SDE and CEDE
Reporting and Recording 2022 Evaluation
04/14/2023
Miscellaneous
1R23 Alpha Sample Analysis
08/29/2022
Miscellaneous
2R23 Alpha Sample Analysis
04/27/2023
Miscellaneous
Prospective Evaluation for Internal Occupational
Exposure Monitoring - 2017 Evaluation
06/05/2018
Miscellaneous
17025:2017
NVLAP Certificate of Accreditation: Mirion
Technologies (GDS), Inc.
07/01/2023
Miscellaneous
4371
Dose Assessments
04/13/2022
Miscellaneous
4372
Dose Assessments
04/01/2022
Miscellaneous
4373
Dose Assessments
04/08/2022
Miscellaneous
4374
Dose Assessments
04/01/2022
Miscellaneous
4753
Dose Assessments
09/09/2023
Miscellaneous
570605002
Dry Active Waste Part 61 Analyses
01/25/2022
Procedures
RCP D-328
Implementation of Personnel Dosimetry Effective
Dose Equivalent
Procedures
RCP D-330
Personnel Dosimetry Evaluations
Procedures
RCP D-370
Evaluation of Internal Deposition of Radioactive
Material
Procedures
RCP D-958
Mirion Technologies DMC 3000 Dosimeter
Radiation
Surveys
2112
1R23 140' CTMT SLD2 Inspection Post Prep
04/04/2022
Radiation
Surveys
83961
Unit 2 Spent Fuel Pool 2-1 Response to Potential
Personnel Contamination
09/08/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Radiation
Surveys
84600
2R23 Seal Table Thimble Tube Pull
10/20/2022
Radiation
Surveys
84601
2R23 Lower Cavity Gripper and Chemistry
Cleanliness Inspection
10/20/2022
Radiation
Surveys
84842
2R23 140' CTMT SLD2 Inspection
10/26/2022
Radiation Work
Permits (RWPs)
RWP 21-0059
Unit 2 FHB Transfer Canal
Radiation Work
Permits (RWPs)
RWP 21-2026
2R22 Lower Cavity and Transfer Canal Work, Lower
Cavity Decon
Radiation Work
Permits (RWPs)
RWP 22-0006
ISFSI Work
Radiation Work
Permits (RWPs)
RWP 22-1026
1R23 Lower Cavity and Transfer Canal Work, Lower
Cavity Decon
Radiation Work
Permits (RWPs)
RWP 22-2026
2R23 Lower Cavity and Transfer Canal Work, Lower
Cavity Decon
Self-
Assessments
51180323
Quick Hit Self-Assessment for NRC IP 71124.04
07/27/2023
Self-
Assessments
PGE 2021-SI-01
Mirion Technologies (GDS), Inc. Environmental TLD
Service (REMP)
08/03/2021
Miscellaneous
RMS-23-072
Radioactive Materials Shipment package: Unit 2
Reactor Coolant Sample B-10/11 sample shipment to
Mass Spec UN2910
10/12/2023
Radiation
Surveys
373544
Gamma Spectroscopy Report: Reactor Coolant
Liquid Unit 2 RCS B-10/B-11
10/12/2023
Radiation
Surveys
89344
RMS-23-072 U2 RCS B-1011 sample shipment to
Mass Spec UN2910
10/12/2023
Corrective Action
Documents
Notifications
51171693, 51176305, 51191628, 51192320,
51192449, 51204831, 51205021
23 Effluents Data
23
Open EMS report for effluents January 1, 2023,
through September 30, 2023
10/10/2023
22 Annual Radioactive Effluent Release Report
05/01/2023
71151
Miscellaneous
WCAP 16465NP
Pressurized-Water Reactor Owners Group Standard
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
RCS Leakage Action Levels and Response
Guidelines for Pressurized-Water Reactors
AWP O-007
Chemistry NRC Performance Indicators
STP I-1B
Routine Daily Checks Required by Licenses
110 and 134
Procedures
STP R-10C
Reactor Coolant System Water Inventory Balance
and 49
Corrective Action
Documents
Notifications
51144967, 51176086, 51176206, 51177967,
51180644, 51193905, 51201169, 51205794,
205903, 51213423
Corrective Action
Documents
Notifications
51189433, 51189455, 51190213, 51190733,
51191015, 51191584, 51192201, 51192296,
51192326, 51194229, 51194459, 51194469,
51194475, 51194612, 51194613, 51195029,
51195031, 51195050, 51195064, 51195135,
51195703, 51196299, 51196958, 51197850,
51198260, 51198263, 51198769, 51198840,
51198886, 51199501, 51200183, 51200709,
201148, 51201169, 51201227, 51201266,
201331, 51201521, 51201772, 51201829,
201922, 51202217, 51202271, 51202694,
202739, 51203102, 51204215, 51204608,
204833, 51204879, 51205422, 51205495,
205536, 51206004, 51206393, 51206625,
206863, 51208202, 51208242, 51209269,
209292, 51209703, 51210835, 51211856,
211874, 51211944, 51212104, 51212110,
212752, 51212754, 51213012, 51213423,
213502, 51213636, 51214308, 51214434,
214575