IR 05000528/2023004

Integrated Inspection Report Report 05000528/2023004 and 05000529/2023004 and 05000530/2023004

ML24039A131
Person / Time
Site:	Palo Verde
Issue date:	02/12/2024
From:	John Dixon NRC/RGN-IV/DORS/PBD
To:	Heflin A Arizona Public Service Co
References
IR 2023004
Download: ML24039A131 (49)

Text

February 12, 2024

SUBJECT:

PALO VERDE NUCLEAR GENERATING STATION - INTEGRATED INSPECTION REPORT 05000528/2023004 AND 05000529/2023004 AND 05000530/2023004

Dear Adam Heflin:

On December 31, 2023, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Palo Verde Nuclear Generating Station. On February 2, 2024, the NRC inspectors discussed the results of this inspection with you and other members of your staff. The results of this inspection are documented in the enclosed report.

Five findings of very low safety significance (Green) are documented in this report. Three 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 IV; the Director, Office of Enforcement; and the NRC Resident Inspector at Palo Verde Nuclear Generating Station.

If you disagree with a cross-cutting aspect assignment or a finding not associated with a regulatory requirement in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region IV; and the NRC Resident Inspector at Palo Verde Nuclear Generating Station. 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, John L. Dixon, Jr., Chief Reactor Projects Branch D Division of Operating Reactor Safety Docket Nos. 05000528, 05000529 05000530 License Nos. NPF-41, NPF-51 NPF-74

Enclosure:

As stated

Inspection Report

Docket Numbers:

05000528, 05000529 and 05000530

License Numbers:

NPF-41, NPF-51 and NPF-74

Report Numbers:

05000528/2023004, 05000529/2023004 and 05000530/2023004

Enterprise Identifier:

I-2023-004-0008

Licensee:

Arizona Public Service

Facility:

Palo Verde Nuclear Generating Station

Location:

Tonopah, AZ

Inspection Dates:

October 1, 2023 to December 31, 2023

Inspectors:

L. Merker, Senior Resident Inspector

E. Lantz, Resident Inspector

N. Cuevas, Resident Inspector

A. Sanchez, Senior Project Engineer

R. Azua, Senior Reactor Inspector

B. Baca, Senior Health Physicist

W. Cullum, Senior Reactor Inspector

R. Deese, Senior Reactor Analyst

K. Clayton, Senior Operations Engineer

D. Antonangeli, Resident Inspector

Approved By:

John L. Dixon, Jr., Chief

Reactor Projects Branch D

Division of Operating Reactor Safety

SUMMARY

The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated inspection at Palo Verde Nuclear Generating Station, 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 Procedure for an Operability Determination of Recirculation Actuation Signal Sump B Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000529/2023004-01 Open/Closed

[H.14] -

Conservative Bias 71111.15 The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to accomplish an activity affecting quality in accordance with the prescribed procedures by failing to follow procedure 40DP-9OP26, Operations Operability Determination Process, revision 48. Specifically, the licensee utilized a previous prompt operability determination to support operability of recirculation actuation signal sump B, without verifying it adequately covered the current condition.

Inadequate Valve Factors for Motor-operated Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000528,05000529,05000530/2023004-02 Open/Closed

[H.12] - Avoid Complacency 71111.21M The inspectors identified a Green finding and associated NCV of 10 CFR 50.55a(b)(3)(ii)when the licensee failed to establish a program to ensure that motor-operated valves (MOVs)continue to be capable of performing their design basis safety functions.

Failure to Follow Security Guide for Use of Air Rifles for Pigeon Control Results in Inadvertent Closure of a Main Turbine Control Valve at Power Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000530/2023004-03 Open/Closed

[H.12] - Avoid Complacency 71152A The inspectors reviewed a self-revealed Green finding for the failure to follow the security desktop instruction Sec DI-02, Use of Air Rifles for Pigeon Control, revision 5, to verify that there was no hazard to plant equipment prior to using an air rifle for pigeon control.

Specifically, on July 18, 2023, the licensees failure to verify that there was no hazard to plant equipment prior to using an air rifle for pigeon control resulted in damage to a main turbine control valve signal cable while operating at 100 percent power, causing the valve to close and a partial load reject to 87 percent power.

Failure to Maintain FLEX Spent Fuel Pool Makeup Pump Consistent with the Requirements of 10 CFR 50.155(b)(1)

Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000528/2023004-04 Open/Closed

[H.9] - Training 71152A The inspectors identified a Green finding and associated NCV of 10 CFR 50.155(b)(1),

Mitigation of Beyond-Design-Events, in that the licensee failed to maintain and implement mitigation strategies that support the pre-deployment and readiness of FLEX spent fuel makeup pumps to support maintaining or restoring spent fuel pool cooling capabilities.

Specifically, in support of the Unit 1 refueling outage 1R24, the licensee failed to stage the FLEX spent fuel pool makeup pump on the designated seismically designed pad and staged it in a location not supported by the licensees mitigation strategies.

Failure to Follow Systematic Troubleshooting Procedure Results in Turbine Trip and Subsequent Complicated Reactor Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000528/2023004-05 Open/Closed EA-23-092

[H.14] -

Conservative Bias 71153 The inspectors reviewed a self-revealed Green finding when the licensee failed to follow procedure 01DP-9ZZ01, Systematic Troubleshooting, revision 12, to provide high confidence that the direct cause of system/equipment degradation had been corrected and that the system/equipment can be restored to normal operation. Specifically, the licensee did not follow the systematic troubleshooting procedure to provide high confidence that the direct cause of the control oil hydraulic fluid reservoir hi-hi/lo-lo level alarm had been corrected.

Additional Tracking Items

Type Issue Number Title Report Section Status LER 05000528/2023-001-01 LER 2023-001-01 for Palo Verde Nuclear Generating Station Unit 1, Reactor Trip Following a Main Turbine Trip 71153 Closed LER 05000528/2023-001-00 LER 2023-001-00 for Palo Verde, Unit 1, Reactor Trip Following a Main Turbine Trip 71153 Closed

PLANT STATUS

Unit 1 entered the inspection period at approximately 95 percent power during a coastdown for refueling outage 1R24. On October 7, 2023, the unit was shut down for refueling outage 1R24.

On November 11, 2023, the reactor was made critical following completion of the refueling outage and returned to full power on November 15, 2023, where it remained for the remainder of the inspection period.

Units 2 and 3 operated at or near full power for the duration of the inspection period.

INSPECTION SCOPES

Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed activities described in IMC 2515, Appendix D, Plant Status, observed risk significant activities, and completed on-site portions of IPs. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.

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 3, auxiliary feedwater train A with auxiliary feedwater train B inoperable during balance of plant - engineered safety features actuation system train B outage, on October 4, 2023

(2) Unit 1, essential cooling water train A following maintenance outage, on October 25, 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 spray train A pump room, fire zone 30A, on October 17, 2023

(2) Unit 1, containment building 80-foot elevation, fire zones 66A, 66B, 67A, and 67B, on October 26, 2023

(3) Units 1, 2, and 3, class 1E 4kV switchgear B room, fire zone 5B, on November 27, 2023

(4) Unit 1, essential spray pond pump train A and B rooms, fire zones 84A and 84B, on November 30, 2023

(5) Unit 1, lower cable spreading room, fire zone 14, on December 12, 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 in Unit 1 during refueling outage 1R24 from October 9-17, 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

• Feedwater Header 1PSGEL011**PIPEXX to Steam Generator RCEE01B (ISI Zone 60) Circumferential Welds 60-7, 60-14,60-15R, 60-16R, and 60-17 (UT)

• Steam Generator 2 1MRCEE01B**HTEXCH Bowl Weld (UT)

• Pressurizer 1MRCEX02**VESSEL Instrument Nozzle B15.180 (Thermowell). Following the identification of a leak, through weld 5-37-7 (instrument tube to base pad weld), the licensee determined the need to replace the instrument tube and associate base pad. The inspector monitored the licensees effort to remove the instrument tube and the base pad. The inspector evaluated and then monitored the licensees effort to install a new base pad and a new instrument tube. The inspector monitored and verified the licensees post maintenance non-destructive examinations (UT).

Dye Penetrant Examination

• Steam Generator 2 1MRCEE01B**HTEXCH Test Line Isolation Valve 1JSGBUV-0224 in line 1SGAUV0220, monitored old valve removal and installation of replacement valve, including tacking of valve in place prior to welding effort. Monitored and inspected first pass welds, upstream and downstream and subsequent final welds. Inspected post maintenance evaluations and non-destructive examinations (PT).

• Pressurizer 1MRCEX02**VESSEL Instrument Nozzle B15.180 (Thermowell)

(PT)

Visual Examination

• Emergency Diesel Generator A-E05 Jacket Water Cooler Circumferential Inner Flange Weld EW23-H031W (VT-1)

• Emergency Diesel Generator A-E05 Jacket Water Cooler Support 1MDGAE05 (VT-3)

• Essential Chiller Inlet Piping Support 1EW023H031**PIPEXX (VT-3)

• Feedwater Header 1PSGEL011**PIPEXX to Steam Generator RCEE01B Snubbers 1SG084H006**SUPORT, 1SG200H013**SUPORT, 1SG200H014*EAST*SUPORT, and 1SG200H014*WEST*SUPORT (VT-3)

Magnetic Particle Examination

• Steam Generator 2 1MRCEE01B**HTEXCH Support Stand Skirt Weld (MT)

Radiographic Examination

• Feedwater Header 1PSGEL011**PIPEXX to Steam Generator RCEE01B Welds 2, 3 & 5 (RT)

Welding Activities

• Gas Tungsten Arc Welding, o

Steam Generator 2 1MRCEE01B**HTEXCH Test Line Isolation Valve 1JSGBUV-0224 in line 1SGAUV0220, monitored old valve removal and installation of replacement valve, including tacking of valve in place prior to welding effort. Monitored and inspected first pass welds, upstream and downstream and subsequent final welds. Reviewed training records of personnel involved, reviewed applicable procedures and verified use and storage of applied weld material.

PWR Inservice Inspection Activities Sample - Vessel Upper Head Penetration Inspection

Activities (IP Section 03.02) (1 Sample)

The inspectors verified that the license conducted the following vessel upper head penetration inspections and addressed any identified defects appropriately:

(1) The inspector observed the setup of remote visual inspection equipment to be used for the inspection of the reactor vessel upper head and associated penetrations and monitored the collection of data for the examination activities, which consisted of a bare metal visual examination of the reactor vessel head surface and penetrations.

No ultrasonic or eddy current examinations were performed during this refueling 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) The inspector performed a boric acid walkdown on October 12, 2023, and evaluated the licensee's actions relating to the following condition reports: CR 22-03437, CR 22-

===03659, CR 22-04542, CR 22-06709, CR 23-02996, CR 23-02997, CR 23-02994, and CR 23-03531.

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) No steam generator tube inspections were scheduled or performed during this refueling outage.

71111.11A - Licensed Operator Requalification Program and Licensed Operator Performance

Requalification Examination Results (IP Section 03.03) (1 Sample)

(1) The inspectors reviewed and evaluated the licensed operator examination failure rates for the requalification annual operating exam administered the weeks of August 17 to September 28, 2023.

71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance

Licensed Operator Requalification Training/Examinations (IP Section 03.02) (1 Sample)

(1) The inspectors observed and evaluated licensed operator continued training simulator activities, on November 29, 2023.

71111.12 - Maintenance Effectiveness

Maintenance Effectiveness (IP Section 03.01) (5 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 2, emergency diesel generator A, on October 10, 2023

(2) Units 2 and 3, reactor protection system, on October 23, 2023

(3) Units 1, 2, and 3, charging pump discharge pulsation dampeners, on November 14, 2023

(4) Unit 3, balance of plant - engineered safety features actuation system train B, on November 17, 2023

(5) Unit 2, main feedwater pump A turbine system, on December 1, 2023

Quality Control (IP Section 03.02) (1 Sample)

The inspectors evaluated the effectiveness of maintenance and quality control activities to ensure the following SSC remains capable of performing its intended function:

(1) Unit 1, advanced rod control hybrid system installation for the control element drive mechanism control system, on November 14, 2023

71111.13 - Maintenance Risk Assessments and Emergent Work Control

Risk Assessment and Management Sample (IP Section 03.01) (5 Samples)

The inspectors evaluated the accuracy and completeness of risk assessments for the following planned and emergent work activities to ensure configuration changes and appropriate work controls were addressed:

(1) Unit 3, high risk for planned balance of plant - engineered safety features actuation system train B outage for replacement of the fuel building essential ventilation actuation signal train B module, on October 4, 2023

(2) Unit 2, high risk for emergent replacement of the main turbine control oil train B filter, on October 24, 2023

(3) Unit 2, high risk for emergent maintenance to replace a blown fuse on the control element drive mechanism control system, on October 25, 2023

(4) Unit 2, high risk for emergent maintenance to troubleshoot and rebuild instrument air moisture separator air filter due to air leak, on October 25, 2023

(5) Unit 3, high risk for planned adjustment of the pressurizer level controller RCN-LIC-110 lo-limit and hi-limit settings, on December 12, 2023

71111.15 - Operability Determinations and Functionality Assessments

Operability Determination or Functionality Assessment (IP Section 03.01) (12 Samples)

The inspectors evaluated the licensee's justifications and actions associated with the following operability determinations and functionality assessments:

(1) Unit 2, essential cooling water train B operability determination due to low flow during flow balance, on October 3, 2023

(2) Unit 3, essential spray pond train A operability determination due to concrete debris on the essential spray pond floor, on October 17, 2023

(3) Unit 3, balance of plant - engineered safety features actuation system train B load sequencer operability determination due to failed auto-sequencer test, on October 26, 2023

(4) Unit 2, recirculation actuation signal sump B operability determination due to feedwater leakage from steam generator 2 economizer feedwater header check valve, SGE-V-006, on October 27, 2023

(5) Units 1, 2, and 3, emergency diesel generators operability determination due to not performing steps in procedure to clean crankcase exhaust screen during a surveillance test, on October 27, 2023

(6) Unit 3, shutdown cooling heat exchanger train B outlet to reactor coolant loop 2A/2B valve, SIB-HV-658, operability determination for seismic concerns due to valve location near system piping, on October 30, 2023

(7) Unit 1, pressurizer vent valve to reactor drain tank, RCA-HV-103, operability determination due to stroke time outside acceptance criteria, on November 9, 2023

(8) Unit 1, reactor coolant system operability determination due to increased identified leakage post unit heatup and startup, on November 15, 2023

(9) Unit 3, shutdown heat exchanger A outlet to spent fuel pool isolation valve, SIA-V-458, operability determination following the discovery of a scaffold built within one eighth of an inch of the valve yoke, on November 30, 2023

(10) Unit 2, reactor coolant loop 1 to shutdown cooling low pressure safety injection pump A suction valve, SIA-UV-655, operability determination due to blown circuit breaker fuse, on December 1, 2023

(11) Unit 3, safety injection tank 1B operability determination due to abnormal rising trend in level, on December 20, 2023

(12) Unit 3, main control room envelope operability determination due to failed weld on upper connection of control room air lock doors mullion, on December 21, 2023

71111.20 - Refueling and Other Outage Activities

Refueling/Other Outage Sample (IP Section 03.01) (1 Sample)

(1) The inspectors evaluated the Unit 1 refueling outage 1R24 activities from October 7 through November 11, 2023. Operating experience smart sample 2007-03, Crane and Heavy lift Inspection, Supplemental Guidance to 71111.20 and 71111.13, revision 3, was used to inform this sample.

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) (5 Samples)

(1) Unit 3, reactor trip switchgear channel D circuit breaker post-maintenance testing following 18-month replacement, on October 10, 2023

(2) Unit 2, emergency diesel generator train A post-maintenance testing following maintenance outage, on October 11, 2023

(3) Unit 1, containment spray pump A discharge to shutdown cooling heat exchanger valve SIA-HV-678 post-maintenance testing following valve internal work, on October 24, 2023

(4) Unit 1, containment spray pump A post-maintenance testing following mechanical seal replacement, on November 6, 2023

(5) Unit 1, main control room smoke essential exhaust duct isolation damper 1-MH-JA-M57 post-maintenance testing following blade seal and shaft packing replacement, on November 6, 2023

Surveillance Testing (IP Section 03.01) (3 Samples)

(1) Unit 2, essential cooling water pump A inservice testing, on October 23, 2023

(2) Unit 1, emergency diesel generator and integrated safeguards train A testing, on November 7, 2023

(3) Unit 1, safety injection tank 1B discharge check valve SIE-V-245 testing, on November 15, 2023

Containment Isolation Valve (CIV) Testing (IP Section 03.01) (1 Sample)

(1) Unit 1, containment spray train B containment isolation valve SIB-V-165 testing, on November 7, 2023

71114.06 - Drill Evaluation

Drill/Training Evolution Observation (IP Section 03.02) (2 Samples)

The inspectors evaluated:

(1) The inspectors evaluated an emergency preparedness drill on September 19, 2023.

(2) The inspectors evaluated an emergency preparedness drill on September 26,

71124.05 - Radiation Monitoring Instrumentation

Walkdowns and Observations (IP Section 03.01) (8 Samples)

The inspectors evaluated the following radiation detection instrumentation during plant walkdowns:

(1) Area radiation monitors located within the Unit 1 fuel handling building

(2) Area radiation monitors located in the Unit 3 auxiliary building

(3) Personnel contamination monitors located at the Unit 2 radiological controlled area exit

(4) Passive radiation monitors located at the Unit 3 radiological controlled area exit

(5) Portable ion chambers staged for use in the Unit 3 radiological controlled area

(6) Portable friskers located for use in the Unit 1 fuel handling building

(7) Whole body counters located at the site's dosimetry office

(8) Small article monitors located at Unit 2 radiological controlled area exit

Calibration and Testing Program (IP Section 03.02) (13 Samples)

The inspectors evaluated the calibration and testing of the following radiation detection instruments:

(1) Thermo Scientific Model RO-20 ion chamber, serial number 11367

(2) Thermo Scientific Model RO-20 ion chamber, serial number 1426

(3) Small article radiation monitor, SAM-12, serial number 669

(4) Alpha and beta sample counter, serial number 339533

(5) Containment train A high range radiation monitor, RU-148

(6) Main steam line radiation monitor, RU-139

(7) Fuel handling building train B vent monitor low range, RU-145

(8) Thermo Scientific FH 40-G digital survey meter, serial number 25611

(9) Thermo Scientific personnel contamination monitor ipcm-12, serial number 12025

(10) Thermo Scientific passive monitor PM-12, serial number 1216

(11) Ludlum Model 26-1 probe, serial number 18362

(12) Fastscan whole body counter #2

(13) Eberline RM-20, serial number 1612 Effluent Monitoring Calibration and Testing Program Sample (IP Section 03.03) (2 Samples)

The inspectors evaluated the calibration and maintenance of the following radioactive effluent monitoring and measurement instrumentation:

(1) Unit 1, plant vent monitor high range, RU-144

(2) Unit 3, fuel building train B vent monitor high range, RU-146

71124.08 - Radioactive Solid Waste Processing & Radioactive Material Handling, Storage, &

Transportation

Radioactive Material Storage (IP Section 03.01)

The inspectors evaluated the licensees performance in controlling, labeling, and securing the following radioactive materials:

(1) Dry active waste and radioactive material stored and processed in the "Dry Active Waste Processing and Storage" facility.

(2) Radioactive sources stored in the calibration bunker.

(3) Radioactive waste and radioactive material stored in the "Low Level Radioactive Material Storage Facility".

Radioactive Waste System Walkdown (IP Section 03.02) (2 Samples)

The inspectors walked down the following accessible portions of the solid radioactive waste systems and evaluated system configuration and functionality:

(1) Observation of a high activity spent resin transfer using the Unit 1 resin transfer system.

(2) Observation of dry active waste processing and storage in the "Dry Active Waste Processing and Storage" facility.

Waste Characterization and Classification (IP Section 03.03) (2 Samples)

The inspectors evaluated the following characterization and classification of radioactive waste:

(1) Dry active waste: Unit 1 dry active waste sealand dated May 24, 2022 and Unit 1 liquid concentrate dated February 24, 2022

(2) High activity/high level spent resin: Unit 1 resin dated August 2, 2023 and Unit 2 resin dated July 11, 2023

Shipping Records (IP Section 03.05) (5 Samples)

The inspectors evaluated the following non-excepted radioactive material shipments through a record review:

(1) Shipping package 21-RW-033 of an LSA-II shipment

(2) Shipping package 22-RW-010 of an LSA-II shipment

(3) Shipping package 22-SH-027 of an SCO-II shipment

(4) Shipping package 23-RW-002 of an LSA-II shipment

(5) Shipping package 23-SH-016 of a Type A package

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)

(3) Unit 3 (October 1, 2022, through September 30, 2023)

BI02: RCS Leak Rate Sample (IP Section 02.11) (3 Samples)

(1) Unit 1 (October 1, 2022, through September 30, 2023)

(2) Unit 2 (October 1, 2022, through September 30, 2023)

(3) Unit 3 (October 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) Units 1, 2, and 3, effectiveness of corrective actions related to charging pump discharge pulsation dampeners, on November 20, 2023

(2) Unit 3, effectiveness of corrective actions following the inadvertent closing of the main turbine control valve 4 at power due to damage to a control cable during pigeon abatement activities, on December 13, 2022

(3) Units 1, 2, and 3, effectiveness of corrective actions related to staging the FLEX spent fuel pool make up pump in incorrect pad during Unit 1 outage, 1R24, on December 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 that might be indicative of a more significant safety issue. The inspectors performed an in-depth review of the licensees modifications quality tracking process and assessed the licensees problem identification threshold, evaluations, and corrective actions related to this process. Overall, the inspectors determined that the licensee has adequately followed site procedures, and no more than minor concerns were identified.

71153 - Follow Up of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)

The inspectors evaluated the following licensee event reports (LERs):

(1) LER 05000528/2023-001-00 and LER 05000528/2023-001-01, Unit 1 Reactor Trip Following a Main Turbine Trip Agencywide Documents Access and Management System (ADAMS) accession numbers ML23158A305 and ML23256A370. The inspectors reviewed the original and updated LER submittals and determined that the cause of the condition described in the LER was reasonably within the licensees ability to foresee and correct and therefore was reasonably preventable. A performance deficiency and finding were identified and are described in the inspection results section of this report. No violation of NRC requirements was identified.

INSPECTION RESULTS

Failure to Follow Procedure for an Operability Determination of Recirculation Actuation Signal Sump B Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000529/2023004-01 Open/Closed

[H.14] -

Conservative Bias 71111.15 The inspectors identified a Green finding and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to accomplish an activity affecting quality in accordance with the prescribed procedures by failing to follow procedure 40DP-9OP26, Operations Operability Determination Process, revision 48. Specifically, the licensee utilized a previous prompt operability determination to support operability of recirculation actuation signal sump B, without verifying it adequately covered the current condition.

Description:

On May 9, 2023, during the Unit 2 refueling outage 2R24, the licensee identified a steam/water leak from the hinge pin assembly cover of steam generator 2 economizer feedwater header check valve, SGE-V-006. The nonborated steam/water leakage was located directly above recirculation actuation signal sump B and started to fill the sump with water. On May 11, 2023, the reactor was made critical following completion of refueling outage 2R24, and the licensee performed an operability determination declaring steam generator 2 economizer feedwater header check valve SGE-V-006 operable. The operability determination also stated recirculation actuation signal sump B would need a compensatory measure to be periodically pumped down to the east containment sump for the water to be pumped out. On May 16, 2023, the reactor returned to full power. From May 15-19, 2023, the licensee attempted to repair the steam leak by injecting leak sealant into the leaking check valve hinge pin assembly cover but was unsuccessful. On May 22, 2023, the unit was shut down to repair the steam leak by welding the leaking hinge pin assembly cover to the valve body. On May 28, 2023, the reactor was made critical following repairs and returned to full power on May 31, 2023.

The inspectors questioned the licensee on the use of a compensatory measure for recirculation actuation signal sump B and noted the licensee did not state whether recirculation actuation signal sump B was operable or inoperable in this condition. On June 16, 2023, the licensee updated the operability determination to remove the compensatory measure statement. The inspectors again noted no operability call regarding recirculation actuation signal sump B and questioned the licensee if recirculation actuation signal sump B had been operable or inoperable with the nonborated water leakage into the sump. On June 24, 2023, the licensee updated the operability determination to address the impact of the nonborated leakage into recirculation actuation signal sump B and declared the sump operable, referencing information from engineering evaluation 23-05203-003 completed on May 11, 2023.

The inspectors reviewed engineering evaluation 23-05203-003 and noted one of the inputs and assumptions was from prompt operability determination 20-12291-003, dated October 3, 2020, that evaluated essentially the same condition (nonborated water leakage from a steam generator economizer feedwater valve into a recirculation actuation signal sump) in Unit 1 at approximately 500 effective full power days (end of core life). The prompt operability determination evaluated, in part, how the presence of the nonborated water in the recirculation actuation signal sump may result in a reactor coolant system boric acid dilution effect after a loss of coolant accident. The prompt operability determination concluded the nonborated water would not cause an inadvertent criticality event or return-to-power and was only applicable through the end of the then-current Unit 1 operating cycle. The inspectors questioned the licensee on whether prompt operability determination 20-12291-003 was appropriate for the condition in Unit 2 since the document stated its conclusion was only applicable through the end of the then-current Unit 1 operating cycle and since Unit 2 had different fuel than Unit 1.

On September 8, 2023, during their review in response to the inspectors questions, the licensee wrote condition report 23-09185 stating engineering evaluation 23-05203-003 and the operability determination dated June 24, 2023, contained nonconservative assumptions regarding core burnup. Condition report 23-09185 further noted the documents utilized prompt operability determination 20-12291-003 that was nonconservative because the Unit 2 low-burnup core had higher reactivity and required a higher boric acid concentration in the reactor coolant system to maintain shutdown margin following a loss-of-coolant accident. On September 26, 2023, the licensee completed engineering evaluation 23-09185-002 using the updated Unit 2 information and concluded the leakage from steam generator 2 economizer feedwater header check valve, SGE-V-006, into recirculation actuation signal sump B was acceptable in all modes of operation provided the water did not completely fill the sump. On September 27, 2023, the licensee wrote an operability determination to incorporate the conclusions of engineering evaluation 23-09185-002 and declared the recirculation actuation signal sump B operable.

The inspectors reviewed the condition reports, updated engineering evaluations, and new operability determination and concluded the documents had addressed the inspectors concerns. The inspectors also validated the licensee had appropriate controls established so the water never completely filled recirculation actuation signal sump B, and therefore maintained design margin to the calculated containment maximum flood level.

Corrective Actions: The licensee shut down the unit and drained steam generator 2 to weld the hinge pin assembly cover to the valve body to stop the leakage into recirculation actuation signal sump B. The licensee also performed a new engineering evaluation and operability determination that demonstrated reasonable assurance that recirculation actuation signal sump B was operable with this condition.

Corrective Action References: Condition reports 23-05203, 23-05751, and 23-09185

Performance Assessment:

Performance Deficiency: The failure to follow procedure 40DP-9OP26, Operations Operability Determination Process, revision 48, was a performance deficiency. Specifically, the licensee utilized a previous prompt operability determination to support operability of recirculation actuation signal sump B without verifying it adequately covered the current condition.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee utilized a previous prompt operability determination to support operability of recirculation actuation signal sump B but did not recognize the previous document evaluated a condition near the end of core life whereas the current operability determination was evaluating an essentially identical condition at the beginning of core life utilizing different fuel, and the licensee performed a new operability determination.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the Mitigating Systems and Barrier Integrity cornerstones were affected.

Using Exhibit 2, Mitigating Systems Screening Questions, the finding was screened as having very low safety significance (Green) because the finding did not represent a deficiency affecting design or qualification of a mitigating structure, system, or component; did not involve a single-train TS system; did not represent the loss of probabilistic risk assessment (PRA) function one train of a multi-train system for greater than its TS allowed outage time; did not represent the 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 the loss of a PRA system and/or function as defined in the PRIB or the licensees PRA for greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; and, did not represent the loss of the PRA function of one or more non-TS trains of equipment designated as risk-significant in accordance with the licensees maintenance rule program for greater than 3 days.

Additionally, the finding did not involve external events mitigating systems, the reactor protection system, fire brigade, or flexible coping strategies.

Using Exhibit 3, Barrier Integrity Screening Questions, the finding was screened as having very low safety significance (Green) because the finding did not involve control manipulations that unintentionally added positive reactivity that challenged fuel cladding integrity (e.g.,

inadvertent boron dilution, cold water injection, two or more inadvertent control rod movements, recirculation pump speed control), did not result in a mismanagement of reactivity by operator(s) that challenged fuel cladding integrity (e.g., reactor power exceeding the licensed power limit, inability to anticipate and control changes in reactivity during crew operations), did not result in the mismanagement of the foreign material exclusion or reactor coolant chemistry control program that challenged fuel cladding integrity (e.g., loose parts, material controls), and did not result from fuel handling errors, a dropped fuel assembly, a misplaced fuel bundle, or crane operations over the core or anywhere in the refueling pathway that challenged fuel cladding integrity or resulted in a release of radionuclides.

Additionally, the finding did not involve potential non-compliance with regulatory requirements for protection of the reactor pressure vessel against fracture (e.g., pressure-temperature limits or pressurized thermal shock issues) and did not involve reactor containment; the control room, auxiliary, reactor, or spent fuel pool building; or the spent fuel pool.

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee rationalized the use of a previous prompt operability determination from Unit 1 for recirculation actuation signal sump B in Unit 2 even though the previous document stated its conclusion was only applicable through the end of the then-current Unit 1 operating cycle.

Enforcement:

Violation: Title 10 CFR Part 50, appendix B, criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. Procedure 40DP-9OP26, Operations Operability Determination Process, revision 48, provides guidelines and instructions for evaluating the operability of structures, systems, or components, when a condition is identified potentially impacting a specified safety function of the system, structure, or component, an activity affecting quality.

Step 4.1.1.N, requires, in part, that where operability has been established for an essentially identical condition under consideration via a condition report or another station document, the reference document shall have been verified.

Contrary to the above, from May 11 to September 27, 2023, the licensee failed to accomplish an activity affecting quality in accordance with procedure 40DP-9OP26. Specifically, when operability had been established for an essentially identical condition under consideration via a condition report or another station document, the reference document was not verified. The licensee utilized a previous prompt operability determination to support operability of recirculation actuation signal sump B but did not recognize the previous document evaluated a condition near the end of core life whereas the current operability determination was evaluating an essentially identical condition at the beginning of core life utilizing different fuel.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Inadequate Valve Factors for Motor-operated Valves Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000528,05000529,05000530/2023004-02 Open/Closed

[H.12] - Avoid Complacency 71111.21M The inspectors identified a Green finding and associated NCV of 10 CFR 50.55a(b)(3)(ii)when the licensee failed to establish a program to ensure that motor-operated valves (MOVs)continue to be capable of performing their design basis safety functions.

Description:

The inspectors reviewed the licensee's document 13-MS-B075, "MOV Valve Factors Study," revision 0. Several issues were noted during the review of this document:

• Group GL-2 did not have any documented differential pressure test data to establish a valve factor. In accordance with self -imposed standard MPR-2524-A, JOG MOV Periodic Verification Program Summary, two or more valves are required to be tested in a group to properly establish a valve factor.

• MOV valve factors study, 13-MS-B075, did not discuss or analyze the effects of side loading on high flow globe valves.

• While there was uncertainty applied to valve factors within groups, it is not clear that the uncertainty meets the two standard deviations criteria set by MPR-2524-A.

The licensee revised the 13-MS-B075 document to revision 1 to incorporate the deficient items identified by the inspectors. Six valves in each unit were found to have inadequate valve factors. As a result, higher valve factors were used when the valve regrouping was completed. Once the new valve factors were determined, they were used in the MOV thrust and torque calculations. The condition was adverse because the higher valve factor resulted in a lower margin. However, all valves retained positive margin and no operability concerns were identified.

Corrective Actions: The licensee entered the issue in their corrective action program. A revision was issued to the MOV Valve Factors Study to incorporate changes identified by the inspectors.

Corrective Action References: 2022-13437

Performance Assessment:

Performance Deficiency: The licensee failed to establish a program to ensure that MOVs continue to be capable of performing their design basis functions in accordance with 10 CFR 50.55a(b)(3)(ii). Specifically, the licensee did not use an appropriate means to establish valve factors for safety-related MOVs which called into question the ability of the valves to perform their safety functions.

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, six valves per unit had inadequate valve factors. When the appropriate valve factors were assigned, the margin for each valve was adversely impacted.

Although, all the affected valves had positive margin and no operability concerns were identified, the licensee had to re-perform a number of valve design calculations to demonstrate that they could meet their design basis functions.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors used Exhibit 2 - Mitigating Systems Screening Questions to assess the significance of the issue. The inspectors determined that the issue screened to Green because the degraded condition did not represent a loss of the PRA system or function of a Technical Specification (TS) system for greater than its TS allowed outage time. Specifically, six valves in each unit were found to have adverse valve factors. When the valve design calculations were performed with the appropriate valve factors, all of the affected valves had positive margin and no operability concerns were identified.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. The inspectors determined that the issue was present performance since it occurred within the past 3 years. Specifically, the issue could have been identified through critical reviews of the MOV program and calculations.

Enforcement:

Violation: Title 10 CFR 50.55a(b)(3)(ii) requires, in part, that licensees establish a program to ensure that MOVs continue to be capable of performing their design basis functions.

Contrary to the above, from October 23, 2012, to September 14, 2023, the licensee failed to establish a program to ensure that MOVs continue to be capable of performing their design basis functions. Specifically, the MOV Valve Factors Study contained several errors which caused six valves in each unit to have adverse valve factors. When the appropriate valve factors were assigned, the margin for each valve was adversely impacted.

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:

05000528,05000529,05000530/2022010-02.

Failure to Follow Security Guide for Use of Air Rifles for Pigeon Control Results in Inadvertent Closure of a Main Turbine Control Valve at Power Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000530/2023004-03 Open/Closed

[H.12] - Avoid Complacency 71152A The inspectors reviewed a self-revealed Green finding for the failure to follow the security desktop instruction Sec DI-02, Use of Air Rifles for Pigeon Control, revision 5, to verify that there was no hazard to plant equipment prior to using an air rifle for pigeon control.

Specifically, on July 18, 2023, the licensees failure to verify that there was no hazard to plant equipment prior to using an air rifle for pigeon control resulted in damage to a main turbine control valve signal cable while operating at 100 percent power, causing the valve to close and a partial load reject to 87 percent power.

Description:

The pigeon abatement program at Palo Verde is an initiative implemented to control the pigeon population on site. Pigeons have been known to use the buildings at the site to roost, which has resulted in areas of the plant being subject to industrial and health challenges from a large bird population. Security desktop instruction Sec DI-02, Use of Air Rifles for Pigeon Control, revision 5, provides guidance to the operations department and designated marksmen of the security department on the use of air rifles for pigeon control on site. Pigeon control activities are performed on all three units at varying frequencies and times of day based on pigeon population and activity.

On July 18, 2023, the Unit 3 main control room experienced an unexpected closing of the main turbine control valve number four resulting in a subsequent partial load rejection and reduction of reactor power to 87 percent. During subsequent troubleshooting, the licensee discovered a short on the control cable for control valve number four that was determined to be damaged by an air rifle lead pellet from a missed shot by a site security officer in the Unit 3 breezeway during pigeon abatement activities. The security officer later reported that there was an instance of a missed target while in the breezeway, and the final location of the pellet was unknown to the officers.

The inspectors reviewed the licensees cause evaluation and determined the designated marksman performing pigeon abatement activities in the Unit 3 breezeway on July 18, 2023, did not follow step 3.8 of the security desktop instruction. Step 3.8 states that the designated marksman shall be responsible for ensuring/validating prior to each shot, that there are no hazards to plant personnel or equipment. In taking a shot at a pigeon in the overhead of the breezeway, the designated marksman did not recognize the potential risk that a missed shot could have on several cable trays containing various electrical cabling between the control building and the turbine building. Further review of the licensee cause evaluation determined that when the program was initially put into practice in 2004, the designated marksman would be accompanied by an auxiliary operator to provide the shooter with an awareness of plant equipment in the vicinity of the activity and the potential impact to the equipment if a shot was missed. This practice was not included as a requirement in the desktop instruction and was discontinued when the leader performing the abatement showed sufficient proficiency. This history of successful performance without an auxiliary operator providing more in-depth knowledge of plant equipment resulted in a mindset that did not recognize potential hazards of conducting this activity.

Corrective Actions: The licensee fixed the control cable and performed a cause evaluation.

The licensee revised the security desktop guide to specify locations where air rifles could be used in the power block.

Corrective Action References: Condition reports 23-07440, 23-07442, 23-07500, 23-07520, 23-12760, and 23-12173

Performance Assessment:

Performance Deficiency: The failure to follow security desktop instruction Sec DI-02, Use of Air Rifles for Pigeon Control, revision 5, was a performance deficiency. Specifically, prior to shooting, the designated marksman did not ensure that there was no hazard to cableways in the overhead of the breezeway.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, on July 18, 2023, the licensees failure to verify that there was no hazard to plant equipment prior to using an air rifle for pigeon control resulted in damage to a main turbine control valve signal cable while operating at 100 percent power, causing the valve to close and a partial load reject to 87 percent power.

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 1, Initiating Events Screening Questions, the finding was screened as having very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.

Cross-Cutting Aspect: H.12 - Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction tools. Specifically, while using an air rifle to perform pigeon abatement activities, the officers relied on past successes and did not identify the potential to damage plant equipment if a shot was missed.

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

Failure to Maintain FLEX Spent Fuel Pool Makeup Pump Consistent with the Requirements of 10 CFR 50.155(b)(1)

Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000528/2023004-04 Open/Closed

[H.9] - Training 71152A The inspectors identified a Green finding and associated NCV of 10 CFR 50.155(b)(1),

Mitigation of Beyond-Design-Events, in that the licensee failed to maintain and implement mitigation strategies that support the pre-deployment and readiness of FLEX spent fuel makeup pumps to support maintaining or restoring spent fuel pool cooling capabilities.

Specifically, in support of the Unit 1 refueling outage 1R24, the licensee failed to stage the FLEX spent fuel pool makeup pump on the designated seismically designed pad and staged it in a location not supported by the licensees mitigation strategies.

Description:

The FLEX spent fuel pool makeup pump was not staged on the correct designated seismic pad during the Unit 1 refueling outage 1R24 activities per the Final Integrated Plan, dated December 24, 2015, which documents the licensees completion of commitments to comply with NRC Order EA-12-049 (which is now codified under 10 CFR 50.155) and states that [Arizona Public Service] will pre-stage critical FLEX equipment prior to each outage at designated seismically designed pads to reduce deployment time and risk.

The FLEX spent fuel makeup pump that was designated equipment relied on for the mitigation of beyond-design-basis events was incorrectly staged and tied down to the FLEX pad #4, which is designated for the FLEX steam generator injection pump. Per the station's initial evaluation, the pad is not explicitly calculated to be sufficient for the FLEX spent fuel makeup pump and is therefore considered to be unavailable for FLEX use as FLEX availability is defined in terms of the equipment being protected from a FLEX seismic event.

As a compensatory measure, the station staged a FLEX steam generator injection pump on FLEX pad #3 as a substitute for the FLEX spent fuel pool makeup pump. Furthermore, this is recognized in the licensees FLEX equipment status control procedure, 14DP-0BD02, PVNGS FLEX Equipment Status Control, revision 4, where out of service is defined as equipment not capable of performing its intended function, or critical equipment not stored correctly (tied down) at the FLEX storage facility or outage unit.

On October 5, 2023, prior to the Unit 1 refueling outage 1R24, two of the sites four FLEX spent fuel pool makeup pumps were staged in support of outage shutdown risk assessment (SRA) and lower mode FLEX strategies. One pump was staged near the demineralized water storage tank, for SRA purposes, and the other pump was tied down to FLEX pad #4. On October 18, 2023, after the resident staff walked down equipment with personnel from the Operations department with regard to FLEX equipment staging, it was identified that the FLEX spent fuel pool makeup pump was pre-staged on the incorrect pad, FLEX pad #4. Per the licensee's evaluation, the routine task work order instructions incorrectly specified to stage the FLEX spent fuel makeup pump used for lower mode FLEX strategies on FLEX pad

1. 4 vice FLEX pad #3. The licensee exited the FLEX strategy when they entered Mode 5 and had reloaded the fuel into the reactor vessel on November 2, 2023.

Corrective Actions: This issue was entered into the licensees corrective action program. The licensee implemented compensatory measures to stage a FLEX steam generator injection pump on FLEX pad #3. The licensee generated an action item to revise the routine task work order to ensure the FLEX spent fuel makeup pump is staged on the correct pad. The licensee will revise the seismic calculations to analyze the FLEX spent fuel pool makeup pump on FLEX pad #4 and include justification that the existing hydraulic analysis bounds this configuration to support changing their FLEX strategies such that the FLEX spent fuel pool makeup pump can be staged on pad #4.

Corrective Action References: Condition Report 23-10834

Performance Assessment:

Performance Deficiency: The failure to stage the FLEX spent fuel pool makeup pump on the designated seismically designed pad was a performance deficiency. Specifically, in support of the Unit 1 refueling outage 1R24, the licensee failed to stage the FLEX spent fuel pool makeup pump on the designated seismically designed pad and staged it in a location not supported by the licensees mitigation strategies.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Configuration Control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, in support of the Unit 1 refueling outage 1R24, the licensee failed to stage the FLEX spent fuel pool makeup pump on the designated seismically designed pad and staged it in a location not supported by the licensees mitigation strategies.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the Barrier Integrity Cornerstone was affected. Using Exhibit 3, Barrier Integrity Screening Questions, the finding was screened as having very low safety significance (Green) because the finding did not adversely affect decay heat removal capabilities from the spent fuel pool causing the pool temperature to exceed the maximum analyzed temperature limit specified in the site-specific licensing basis, did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the spent fuel pool that caused mechanical damage to fuel clad AND a detectible release of radionuclides, did not result in a loss of spent fuel pool water inventory decreasing below the minimum analyzed level limit specified in the site-specific licensing basis, and did not affect the spent fuel pool neutron absorber, fuel bundle misplacement (i.e., fuel loading pattern error) or soluble boron concentration.

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, since the submission of the Final Integrated Plan on December 24, 2015, the FLEX spent fuel makeup pump has been periodically staged on the incorrect pad during outages and walked down by personnel from several departments such as the Fire department, Operations, and Engineering Department, unaware of the exact FLEX requirements.

Enforcement:

Violation: Title 10 CFR 50.155(b)(1) requires, in part, that each licensee shall develop, implement, and maintain mitigation strategies for beyond-design basis external events, capable of being implemented site-wide and include maintaining or restoring spent fuel pool cooling capabilities.

Contrary to the above, from October 5, 2023, through November 2, 2023, the licensee failed to develop, implement, and maintain mitigation strategies for beyond-design basis external events, capable of being implemented site-wide and include maintaining or restoring spent fuel pool cooling capabilities. Specifically, in support of the Unit 1 refueling outage 1R24, the licensee failed to stage the FLEX spent fuel pool makeup pump on the designated seismically designed pad and staged it in a location not supported by the licensees mitigation strategies.

Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.

Failure to Follow Systematic Troubleshooting Procedure Results in Turbine Trip and Subsequent Complicated Reactor Trip Cornerstone Significance Cross-Cutting Aspect Report Section Initiating Events Green FIN 05000528/2023004-05 Open/Closed EA-23-092

[H.14] -

Conservative Bias 71153 The inspectors reviewed a self-revealed Green finding when the licensee failed to follow procedure 01DP-9ZZ01, Systematic Troubleshooting, revision 12, to provide high confidence that the direct cause of system/equipment degradation had been corrected and that the system/equipment can be restored to normal operation. Specifically, the licensee did not follow the systematic troubleshooting procedure to provide high confidence that the direct cause of the control oil hydraulic fluid reservoir hi-hi/lo-lo level alarm had been corrected.

Description:

Control oil pumps A and B provide high-pressure fluid for the operation of the main turbine control valves, stop valves, combined intercept valves, as well as the trip devices in the trip and overspeed protection circuits. During normal operations, one pump operates with the second pump in standby. If the running pump trips and the standby pump does not auto-start, control oil header pressure will continue to slowly drop and result in a main turbine trip. Both pumps pull hydraulic fluid from a common control oil hydraulic fluid reservoir. The control oil reservoir has one normal high/low level indicating switch, CON-LIS-13, and one emergency high/low level switch, CON-LS-14. The normal high/low level indicating switch, CON-LIS-13, provides level indication of the control oil pump system reservoir and hi and lo level alarms in the main control room. The emergency high/low level switch, CON-LS-14, is internal to the reservoir and provides hi-hi and lo-lo level alarms to the main control room.

On June 20, 2022, the Unit 1 control room received a control oil hydraulic fluid reservoir hi-hi/lo-lo level alarm. The licensee followed the alarm response procedure, and an auxiliary operator reported hydraulic fluid level within normal operating range at +2.5 inches above normal per local level gauge, CON-LIS-13. On June 21, 2022, maintenance performed troubleshooting on the emergency high/low level switch, CON-LS-14, with no abnormalities noted. On June 30, 2022, the licensee performed additional troubleshooting on CON-LS-14, which appeared to be operating correctly. The licensee decided to not support additional troubleshooting efforts because of a perceived risk to online operations. Utilizing previous operating experience and the belief that the normal high/low level indicating switch, CON-LIS-13, was providing accurate indication, the licensee decided to jumper out the hi-hi/lo-lo level alarm until the next refueling outage and installed the jumper on July 5, 2022.

Between July 2022 and March 2023, Unit 1 experienced four instances when the standby control oil pump experienced a delay in developing discharge pressure during pump starts, three instances for control oil pump A and one instance for control oil pump B, respectively.

On April 8, 2023, at 9:41 p.m. control oil pump A tripped on an electrical protection ground fault, and control oil pump B automatically started. The running amps for control oil pump B indicated low, and the control oil system pressure continued to degrade. This delay in developing pressure led to a main turbine trip on low control oil header pressure at 9:43 p.m.

The turbine trip resulted in a reactor power cut back, but an automatic reverse power relay actuation did not occur to trip the generator switchyard output breakers.

At 9:44 p.m. operators manually opened the main generator switchyard output breakers per abnormal operating procedure 40AO-9ZZ08, Load Rejection, revision 35. Since the breakers were manually opened, a fast bus transfer did not occur, resulting in the de-energization of the 13.8 kV non-class buses, NAN-S01 and NAN-S02, which resulted in the loss of power to all four reactor coolant pumps and therefore a reactor trip and loss of forced circulation. Due to the loss of the 13.8kV non-class buses, the condensate pumps de-energized which caused the main feedwater pumps to trip on low suction pressure. Operators manually started the motor-driven auxiliary feedwater pump to feed both steam generators.

At 10:02 p.m. operators completed standard post trip actions and entered emergency operating procedure 40EP-9EO07, Loss of Offsite Power - Loss of Forced Circulation, revision 31. No emergency plan classification was required. At 10:04 p.m. operators manually actuated a main steam isolation signal per the loss of forced circulation emergency operating procedure, due to the loss of the circulating water flow to the main condenser. This isolated the main steam header and prevented the use of the steam bypass control system, requiring the use of atmospheric dump valves for heat removal. At 11:08 p.m. operators restored 13.8 kV power to NAN-S01 through NAN-S03 from offsite power. On April 9, 2023, at 1:24 a.m.

operators restored 13.8 kV power to NAN-S02 through NAN-S04 from offsite power. On April 9, 2023, at 3:10 a.m. operators restarted reactor coolant pumps 1A and 2A, restoring forced flow circulation of the reactor coolant system.

Following the event, the licensee determined that the control oil reservoir was actually at -19 inches below normal operating level. At this level, both control oil pumps air bleed valves return lines are exposed to air, causing the standby pump to lose its prime, which is why the control oil pump B failed to develop discharge pressure. The licensee also found the normal high/low level indicating switch, CON-LIS-13, float arm assembly disconnected from the mounted level gauge because the float arm assembly set screw had loosened and fallen out.

Per the licensees cause evaluation, this failure had occurred coming out of the Unit 1 refueling outage 1R24 in April 2022 when the disconnected float arm assembly caused the normal high/low level indicating switch, CON-LIS-13, to display an erroneous +2.5 inches above normal fluid level.

The inspectors determined the licensee used procedure 30DP-9MP27, Toolpouch and Minor Maintenance Process, revision 0, and did not use procedure 01DP-9ZZ01, Systematic Troubleshooting, revision 12, when troubleshooting the control oil reservoir hi-hi/lo-lo alarm condition in June 2022 before jumpering out the hi-hi/lo-lo alarm on July 5, 2022. Step 4.1.2.F of procedure 30DP-9MP27 states toolpouch maintenance can be used to perform non-intrusive inspections and data gathering in support of Systematic Troubleshooting using Appendix D and Appendix H (Red Sheet) per 01DP-9ZZ01, Systematic Troubleshooting. The inspectors determined more intrusive troubleshooting was warranted to determine the actual control oil reservoir level by direct measurement, and that the licensee did not demonstrate a conservative bias when given conflicting indications from two level instruments. The inspectors noted the licensee had other opportunities to utilize the systematic troubleshooting procedure when both control oil pumps experienced delays in developing discharge pressure during pump starts between July 2022 through March 2023.

Corrective Actions: The licensee added approximately 275 gallons of hydraulic fluid to the control oil reservoir, removed the jumper for the emergency high/low level switch lo-lo alarm, replaced the broken normal high/low level indicating switch, and performed a root cause evaluation.

Corrective Action References: Condition reports 23-03479 and 23-03648

Performance Assessment:

Performance Deficiency: The failure to follow procedure 01DP-9ZZ01, Systematic Troubleshooting, revision 12, was a performance deficiency. Specifically, when investigating the control oil hydraulic fluid reservoir hi-hi/lo-lo level alarm, the licensee failed to follow Step 1.1.2, which states, the use of this procedure provides a systematic approach to data collection, failure analysis and a test/measurement plan that results in high confidence that the direct cause of system/equipment degradation has been corrected and that the system/equipment can be restored to normal operation.

Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee did not recognize that the control oil reservoir normal high/low level indicating switch was broken resulting in an invalid indication. The licensee incorrectly used this information to jumper out the emergency high/low level switch lo-lo alarm, which was correctly alarming due to low control oil level. Actual low control oil reservoir level caused the loss of control oil header pressure, resulting in a turbine trip and subsequent reactor trip.

Significance: The inspectors assessed the significance of the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The initiating events screening questions for transients in Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated January 1, 2021, do not have a question specific to turbine trips. Section 0609A-03, Guidance, states that the screening questions cover a wide range of instances and scenarios but are not intended to be all inclusive. In the absence of a screening question for turbine trips, the inspection staff and senior reactor analysts concluded that performance of a detailed risk evaluation was necessary.

Further, the licensees response to the turbine trip de-energized both 13.8kV non-class buses and caused a loss of power to the reactor coolant pumps that resulted in a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of condenser). Per Inspection Manual Chapter 0609, appendix A, a finding that causes a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g., loss of condenser) requires a detailed risk evaluation.

Methodology - Because this occurrence was an event and not a degraded condition, the senior reactor analyst used the guidance in Chapter 8, Initiating Events Analyses, Volume 1, Internal Events, of the Risk Assessment of Operational Events Handbook, Revision 2.02.

Specifically, Section 8.2, Case 1 - Initiating Event Only, was used because the performance deficiency did not cause the unavailability of additional mitigating systems.

Assumptions - The following assumptions with their required model modifications were made to conduct the analysis:

• Switchyard conditions effect on the fast bus transfer - The analyst assumed that the fast bus transfer would not work if another Palo Verde Unit was in an outage at the time of a turbine trip on Unit 1.

• Fire water feeding of steam generators - The analyst considered fire water feeding of steam generators was a viable option for averting core damage, though plant recovery would take additional time prior to resuming normal operations.

The senior reactor analyst noted that the event was best characterized as a loss of condenser heat sink event initiated by a turbine trip initiating event. Noting that the Palo Verde SPAR model did not have a turbine trip initiating event, the analyst added a turbine trip initiating event tree to the model which would lead to an OK end state if the fast bus transfer properly functioned and would transfer to the loss of condenser heat sink event tree if the fast bus transfer failed to function. A top event which evaluated whether conditions in the Palo Verde switchyard would lead to failure of the fast bus transfer feature was used.

Under the switchyard conditions top event, the analyst created a fault tree which contained the failures which would lead to the unavailability of the fast bus transfer feature. These failures included logic which combined independent failures of both bus NAN-S01 and bus NAN-S02 and the percentage of the time that the Palo Verde switchyard was in a condition which would inhibit operation of the fast bus transfer.

For the fault tree above, the analyst created a basic event for the time the Palo Verde switchyard was in a condition to inhibit the fast bus transfer. The licensee provided information in Action Item 23-03479-065 that detailed 5 factors which affected the percentage of time the fast bus feature would not work. The analyst reviewed the information and judged that the time other Palo Verde units were in outage was the dominant factor and used that value the licensee had derived (15.4% of a given year) for the basic event probability.

Model Modifications - The analyst made several modifications to fault trees AFW-SG1-N and AFW-SG2-N, which evaluate the probability of failure of flow to the steam generators from the main feedwater system and, the non-safety train of the auxiliary feedwater system, and the condensate system, to better reflect the availability of these systems.

First, the analyst added logic to represent the licensees strategy to depressurize a steam generator and feed the steam generator with a fire water pump. The licensee had previously added a hard-piped modification to accomplish this in 2019. The analyst reviewed applicable flow calculations, thermal-hydraulic computations, operator procedures, and steam generator design information, to confirm the viability of this strategy. The flow path tapped in at the discharge of the non-safety auxiliary feedwater pump, so the analyst inserted strategy logic under fault tree AFW-FSP-SG1N, which contains non-safety auxiliary feedwater pump failures.

The new fault tree contains probabilistic failures to run or start and be in maintenance of Palo Verdes two engine-driven fire water pumps using template data for engine-driven pumps and generic demand and rate alpha factors from the SPAR model. Also, a human failure event quantified at 2.60E-2 was created using the SPAR-H human reliability analysis methodology.

In evaluating this event the analyst considered all performance shaping factors nominal except high stress for diagnosis and action and low experience/training for action.

Additionally, the analyst reviewed the human failure events under sub-fault tree AFW-SG1N-INJ-ISO and AFW-SG2N-INJ-ISO which evaluate the isolation of feed water injection valves, since these valves were isolated by the operators during the event. The analyst used the SPAR-H methodology to re-quantify basic events AFW-XHE-XM-MSIS, Operator Fails to Reopen Feed Injection Valves After MSIS (PSA), and AFW-XHE-XM-FWIV2HR, Control Room Operators Fail to Direct an AO to the MSSS Building for Manual Control of the Feed Water Isolation Valves, based on new information related to these events discovered during this review.

Basic event AFW-XHE-XM-MSIS was re-quantified to be 2.0E-3 by applying using nominal values for all performance shaping factors except using extra time for available time for diagnosis. This determination was made after noting that the manipulation time was under five minutes with nearly one hour of time available.

Basic Event AFW-XHE-XM-FWIV2HR was re-quantified to be 1.1E-3 by applying using nominal values for all performance shaping factors except using extra time for available time for diagnosis and action. This determination was made after noting that for the applicable conditions of a loss of instrument air, backup nitrogen would be available for nearly seven hours which would give additional time to dispatch an auxiliary operator.

Finally, basic event HPI-XHE-XM-RECIRC, Operator Fails to Start/Control High Pressure Recirculation - PWR, was set to IGNORE to better model the fact that Palo Verde accomplishes this event via automatic circuitry and not manual manipulation by operators.

Conditional Core Damage Probability (CCDP) - The analyst made the previously described modifications to the Palo Verde SPAR model, Version 8.80, and conducted runs on SAPHIRE, Version 8.2.8, to estimate the CCDP. Then the analyst set the turbine trip initiating event frequency to 1.0 and the initiating event frequencies for all other initiators to 0.0, as well as setting house event HE-MSIS, MSIS Event Has Occurred, to TRUE to assess the event.

This yielded an estimate of CCDP of 4.2E-7.

The analyst obtained the initiating event frequency for turbine trips at pressurized water reactors (1.5E-1/year) from the Idaho National Laboratory. This value was applied to obtain a baseline CCDP of 6.3E-8. The analyst subtracted this baseline CCDP value from the event CCDP to obtain an estimate of 3.6E-7 for the event, characterizing the issue with very low safety significance.

Conditional Large Early Release Probability (CLERP) - Using Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, the analyst screened all large early release sequences as being of very low safety significance except consequential steam generator tube ruptures. The analyst then applied the guidance in Volume 5, Risk Analysis of Containment-Related Events (LERF), of the Risk Assessment of Operational Events Handbook, Revision Draft, January 2018, to evaluate the increase in CLERP from a consequential steam generator tube rupture. The analyst considered all applicable accident sequences to be characterized by having high reactor coolant system pressure, at least one dry steam generator, and low steam generator pressure. The analyst also noted that Palo Verde, Unit 1, had low leakage reactor coolant pump seals which would inhibit reactor coolant loop clearing. Applying these attributes to Table 2.4-2, Recommended p(csgtr) values for Accident Sequences for an Unfavorable SG (CE SG-like), the analyst used a LERF factor of 0.2 to estimate CLERP from the performance deficiency of 7.2E-8/year.

External Events - The CCDP and CLERP from external events were negligible because the probability of any external events coincident with this turbine trip initiating event would be extremely low.

Sensitivities - The following sensitivities were run to estimate the potential variations in significance resulting from changes in influential assumptions.

• Fire water feeding of steam generators more successful - The analyst applied the licensees value for operator failure to feed the steam generators with fire water (5.54E-3) which had a lower failure rate than the analyst derived from using the SPAR-H methodology and estimated a CCDP of 3.0E-7.

• No credit for fire water feeding of steam generators - The analyst gave no credit for the fire water feeding strategy to obtain what the analyst believes to be an overly conservative CCDP of 3.0E-6.

• Switchyard conditions which inhibit fast bus transfer less frequent - The analyst applied the licensees proposed value of the fraction of the year in which the fast bus transfer would be unavailable due to switchyard conditions of 1.5E-2, which reflects the dependency of another unit being in outage and Rudd line outages. These assumptions produce a CCDP of 4.4E-8.

Switchyard conditions which would produce a White CCDP - The analyst varied the probability of the basic event which represented the time the fast bus transfer would have been inhibited to obtain a CCDP of 1.0E-6. The probability would be 4.4E-1 or 161 days.

Conclusion - The analyst estimated the increase in conditional core damage probability (CCDP) from the performance deficiency to be 3.6E-7/year and the increase in conditional large early release probability (CLERP) to be 7.2E-8/year. These results are of very low safety significance (Green).

Cross-Cutting Aspect: H.14 - Conservative Bias: Individuals use decision making-practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, when confronted with two conflicting indications for control oil reservoir level, the licensee believed the indicator showing sufficient hydraulic fluid level and removed the more conservative (lo-lo) level alarm without considering the possibility of a failed indicator or measuring actual reservoir level. The licensee also did not identify that operator rounds recorded the same control oil level from the normal high/low level indicating switch, CON-LIS-13, since May 2022.

Enforcement:

Inspectors did not identify a violation of regulatory requirements associated with this finding.

The inspectors verified no proprietary information was retained or documented in this report.

• On November 3, 2023, the inspectors presented the Palo Verde Nuclear Generating Station, Unit 1, inservice inspection results to Cary Harbor, Vice President, Nuclear Regulatory & Oversight, and other members of the licensee staff.

• On November 21, 2023, the inspectors presented the occupational and public radiation safety inspection results to Todd Horton, Senior Vice President, Site Operations, and other members of the licensee staff.

• On December 6, 2023, the inspectors presented the exit meeting for the closure of URI 05000528,05000529,05000530/2022010-02 inspection results to Cary Harbor, Vice President, Nuclear Regulatory & Oversight, and other members of the licensee staff.

• On February 2, 2024, the inspectors presented the integrated inspection results to Adam Heflin, Executive Vice President and Chief Nuclear Officer, and other members of the licensee staff.

DOCUMENTS REVIEWED

Inspection

Procedure

Type

Designation

Description or Title

Revision or

Date

71111.04

Corrective Action

Documents

23-00750, 23-05952, 23-08254, 23-08258, 23-09825, 23-

10073, 20-07410, 20-03186

71111.04

Drawings

01-M-EWP-001

P & I Diagram Essential Cooling Water System

71111.04

Drawings

03-M-AFP-001

P & I Diagram Auxiliary-Feedwater System

71111.04

Procedures

40DP-9OPA1

Turbine Building Upper Level Watch Station rounds

71111.04

Procedures

40OP-9AF01

Essential Auxiliary Feedwater System

71111.04

Procedures

40OP-9AF01

Essential Auxiliary Feedwater System

71a

71111.04

Procedures

40OP-9EW01

Essential Cooling Water System (EW) Train A

71111.04

Procedures

73ST-9AF02

Auxiliary Feedwater A - Inservice Test

71111.04

Work Orders

4511539, 5433597, 5458966

71111.05

Corrective Action

Documents

23-10610, 19-04199, 23-10612, 23-10794, 23-10795, 23-

2317, FSCCR 5581849, 23-12417, 23-12400, 23-12101

71111.05

Drawings

01-M-FPP-006

P & I Diagram Fire Protection System

71111.05

Fire Plans

Pre-fire Strategies Manual

71111.05

Miscellaneous

5565756

Transient Combustible Control Permit

71111.05

Miscellaneous

5571949

FSCCR

71111.05

Miscellaneous

5571974

FSCCR

71111.05

Procedures

14DP-0FP01

Firewatch Requirements

71111.05

Procedures

14DP-0FP02

Fire System Impairments and Notifications

71111.05

Procedures

14DP-0FP02

Fire System Impairments and Notifications

71111.05

Procedures

14FT-1FP05

Unit 1 Wet Pipe, Deluge and Pre-Action System Inspection

71111.05

Procedures

14FT-9FP08

CO2 Fire Suppression System Functional Test

71111.05

Procedures

14FT-9FP74

Fire Water Containment Isolation Valve Surveillance

71111.05

Procedures

14OP-9FP01

Fire Water Suppression System

71111.05

Procedures

40DP-9ZZ17

Control of Doors, Hatches and Floor Plugs

71111.05

Procedures

40OP-9HJ02

Control Building HVAC (Smoke Removal)

71111.08P

Corrective Action

Documents

CR

2-03437, 22-03659, 22-04542, 22-06709, 23-02996, 23-

2997, 23-02994, 23-03531

71111.08P

Corrective Action

CR

23-10380, 23-10447

Documents

Resulting from

Inspection

71111.08P

Miscellaneous

4INT-ISI-1

4th Inspection Interval - Inservice Inspection Program

Summary Manual - PVGS Unit 1

71111.08P

NDE Reports

20-VT-1008

Pressure Test Report - VE of PZR Instrument Connections -

5-37 location 7 (TE-101)

10/11/2020

71111.08P

NDE Reports

2-VT-1036

Pressure Test Report - VE Bare Metal of PZR Instrument

Nozzles

04/10/2022

71111.08P

NDE Reports30-881

Liquid Penetrant Examination Report - Pressurizer Nozzle

Weld and Weld Pad

10/13/2020

71111.08P

Procedures

70TI-9ZC01

Boric Acid Walkdown Leak Detection

71111.08P

Procedures

73DP-9WP01

Welder and Procedure Qualification

71111.08P

Procedures

73DP-9WP04

Welding and Brazing Control

71111.08P

Procedures

73DP-9WP05

Weld Filler Material Control

71111.08P

Procedures

73DP-9ZC01

Boric Acid Corrosion Control Program

71111.08P

Procedures

73TI-0ZZ02

Ultrasonic Thickness Measurement

71111.08P

Procedures

73TI-0ZZ13

Radiographic Examination

71111.08P

Procedures

73TI-9RC09

Bare Metal Visual Examination of Reactor Vessel Upper

Head

71111.08P

Procedures

73TI-9ZZ05

Dry Magnetic Particle Examination

71111.08P

Procedures

73TI-9ZZ07

Liquid Penetrant Examination

71111.08P

Procedures

73TI-9ZZ10

Ultrasonic Examination of Welds in Ferritic Components

71111.08P

Procedures

73TI-9ZZ17

Visual Examination of Welds, Bolting and Components

71111.08P

Procedures

73TI-9ZZ18

Visual Examination of Component Supports

71111.08P

Procedures

73TI-9ZZ22

Visual Examination For Leakage

71111.08P

Procedures

PDI-UT-1

PDI Generic Procedure for the Ultrasonic Examination of

Ferritic Pipe Welds

G

71111.08P

Self-Assessments SWMS # 17-

09564

Simple Self Assessment - Welding Program Self

Assessment

06/30/2017

71111.08P

Self-Assessments SWMS # 22-

06769

Simple Self Assessment Report - Unit1, 4th Interval, First

Period ISI

07/08/2022

71111.08P

Work Orders

WO/Task

5308159/0, 5458822/0, 5458998/0, 5459333/0, 5459333/7,

5459333/9,

71111.11A

Miscellaneous

15DP-0OT04

Appendix G

23 NRC Annual Operating Test Results

71111.11Q

Miscellaneous

NLR235050301.2

Loss of Vacuum/LOOP/LOFC

08/25/2023

71111.11Q

Miscellaneous

NLR235050301.3

Fire Alarms with Spurious Equipment Actuation

11/16/2023

71111.12

Corrective Action

Documents

23-09752, 23-09720, 23-09696, 23-09721, 23-09827, 23-

09796, 23-09708, 22-09417, 22-13323, 23-00901, 23-

01797, 23-02470, 23-02728, 23-03662, 23-03944, 23-

04749, 23-04878, 23-10768, 23-10502, 23-11634, 23-

10162, 23-04824, 22-10063, 23-08071, 23-10058, 23-

11689, 23-03747, 23-05432, 23-05635, 23-05990, 23-

06642, 23-10837

71111.12

Drawings

01-E-SFF006

Control Wiring Diagram Reactor Control System CEDMCS

Cabinet C4 Sheet 10

71111.12

Drawings

01-E-SJF-003

Control Wiring Diagram Plant Distributed Control System

Ovation Infrastructure Network Sheet 1

71111.12

Drawings

01-E-SJF-003

Control Wiring Diagram Plant Distributed Control System

Ovation Infrastructure Network Sheet 3

71111.12

Drawings

01-E-SJF-003

Control Wiring Diagram Plant Distributed Control System

Ovation Infrastructure Power Sheet 2

71111.12

Miscellaneous

PVNGS Operations Quality Assurance Program Description

(QAPD)

71111.12

Miscellaneous

MR 2.0 - Performance Monitoring Bases Document

Worksheet - Diesel Generator PMB

71111.12

Miscellaneous

Material engineering evaluation 04569, Replace Pulse Tone

Piston with type 304 stainless steel in place of 303 stainless

steel

71111.12

Miscellaneous

MR 2.0 Performance Monitoring Bases Document

Worksheet - ESFAS

71111.12

Miscellaneous

13-EN-0306

Installation Cable Splicing and Terminations

71111.12

Miscellaneous

EVAL-PVNGS-

03185

Maintenance Rule Event

71111.12

Miscellaneous

EVAL-PVNGS-

FT-03144

Unit 2 Automatic Reactor Trip

71111.12

Miscellaneous

EVAL-PVNGS-

Unit 2 Automatic Reactor Trip

FT-03144

71111.12

Miscellaneous

EVAL-PVNGS-

SB-03063

Maintenance Rule (a)(1) Evaluation

71111.12

Miscellaneous

N001-0701-00023

INSTALL AND ASSY 5 GAL 3010 PULSETONE V-VE-8944

14SE79 C/L 28

71111.12

Miscellaneous

SB PMB

MR 2.0 - Performance Monitoring Bases Document

Worksheet - Reactor Protection System

71111.12

Miscellaneous

SPEC 13-PN-

0410

General Maintenance Practices

71111.12

Procedures

2DP-0ZZ03

Inspection and Testing of Plant Maintenance and

Modifications

71111.12

Procedures

30DP-0AP01

Maintenance Work Order Writers Guide

71111.12

Procedures

30DP-9MP01

Conduct of Maintenance

71111.12

Procedures

36MT-9SA01

Balance of Plant Engineered Safety Features Actuation

System Functional Test

71111.12

Procedures

36ST-9SA05

FBEVAS, CREFAS, and CRVIAS 18 month Functional Test

71111.12

Procedures

40OP-9SA02

De-Energization of BOP ESFAS

71111.12

Procedures

60DP-0QQ17

Nuclear Assurance Inspections

71111.12

Procedures

60DP-0QQ21

Qualification and Certification of QA/QC Inspection

Personnel

71111.12

Procedures

70DP-0MR01

Maintenance Rule

71111.12

Procedures

70DP-0MR01

Maintenance Rule

71111.12

Procedures

70DP-0MR02

Maintenance Rule Monitoring Process

71111.12

Procedures

85DP-9MS01

Conduct of Modifications

71111.12

Work Orders

5454663, 5454637, 5454605, 5459882, 5454594, 5454597,

5454665, 5209620, 5363801, 5418854, 5418921, 5571374,

5573609, 5569781, 5477186, 5554974, 5516678, 5526592,

5536211, 5537359, 5539341

71111.13

Corrective Action

Documents

23-06556, 23-06579, 23-09468, 23-10854, 23-10966, 21-

06842, 21-07979

71111.13

Miscellaneous

Unit 2 Control Oil Pump Strategy Night Order

06/20/2023

71111.13

Miscellaneous

ACMP, B CO Fluctuating and low discharge pressure

combined with reduced reliability

71111.13

Miscellaneous

Protected Equipment for CONP01B OOS

10/27/2023

71111.13

Miscellaneous

Fire Department Turnover Document - Work Mechs with

Zone

08/06/2023

71111.13

Miscellaneous

PVGS Operator Information Manual

April 2020

71111.13

Miscellaneous

PRA Review Fire Protection Impairments

08/06/2023

71111.13

Miscellaneous

Phoenix RM U3 Train-B BOP-ESFAS Sequencer What-If

08/07/2023

71111.13

Miscellaneous

Protected Equipment for B BOP ESFAS Sequencer

INOPERABLE

10/04/2023

71111.13

Miscellaneous

Protected Equipment For both SBOGS DG B INOPERABLE

in Unit 3

10/04/2023

71111.13

Miscellaneous

Equipment Status (EST) Tag Log, Unit 3

10/04/2023

71111.13

Miscellaneous

Phoenix Risk Monitor - PVGS Unit 3 Current Risk Summary

Report

10/04/2023

71111.13

Miscellaneous

Risk Management and Readiness Plan - U3 B BOP

ESFAS Downpower

71111.13

Miscellaneous

High Risk Activity Mitigation Summary - Work Week 2340

71111.13

Miscellaneous

Unit 3 Control Room Logs - 12/12/2023

2/12/2023

71111.13

Miscellaneous

WO 534495 & 5346907 Risk Management and Readiness

Plan

71111.13

Miscellaneous

Risk Management and Readiness Plan for WO 5573137:

Troubleshoot and Correct the Continuous Gripper High

Voltage for Subgroup 3 and Indication of Blown Fuse

10/19/2023

71111.13

Miscellaneous

PVGS Unit 2 Phoenix Risk Monitor Plant State as of

10/19/2023 12:24 PM

10/19/2023

71111.13

Miscellaneous

PVGS Unit 2 Phoenix Risk Monitor Plant State as of

10/19/2023 10:50 PM

10/19/2023

71111.13

Miscellaneous

Risk Management and Readiness Plan for WO 5573669:

Troubleshoot/Repair Instrument Air Moisture Separator

SGA-UV-172

10/22/2023

71111.13

Miscellaneous

PVGS Unit 2 Phoenix Risk Monitor Plant State as of

10/21/2023 12:43 AM

10/21/2023

71111.13

Miscellaneous

PVGS Unit 2 Phoenix Risk Monitor Plant State as of

10/22/2023 1:53 PM

10/22/2023

71111.13

Miscellaneous

13-JS-B002

Air-Operated Valve (AOV) Risk Ranking and Categorization

71111.13

Miscellaneous

13-VTD-M322-

00010

Miller Type F12 1/2 Basic Compressed Air Filter [PUB. #NI-

170]

71111.13

Procedures

01DP-0AP57

Management of Critical Evolutions and Infrequently

Performed Tests or Evolutions

71111.13

Procedures

2DP-0RS01

Online Integrated Risk

71111.13

Procedures

40DP-9RS01

Operations Department Online Nuclear Risk Determination

Modes 1 and 2

71111.13

Procedures

40DP-9RS03

Risk Management Actions

71111.13

Procedures

40OP-9CH01

CVCS Normal Operations

71111.13

Procedures

40OP-9CO01

Electro-Hydraulic Control System

71111.13

Procedures

40OP-9OP01

Manual Operation of Air Operated Valves

71111.13

Procedures

40OP-9SA02

De-Energization of BOP ESFAS

71111.13

Procedures

40OP-9SF01

Control Element Drive Mechanism Control System

(CEDMCS) Operation

71111.13

Procedures

40TI-9SF01

Pressurizer Level Control System Tuning

71111.13

Procedures

70DP-0RA05

Assessment and Management of Risk When Performing

Maintenance in Modes 1 and 2

71111.13

Procedures

73DP-9ZZ20-02

AOV Risk Ranking and Categorization

71111.13

Work Orders

5554974, 5573225, 5461826, 5573137, 5573669, 5346907,

5354495

71111.15

Calculations

03-MC-SI-0502

ESF Pump Suction Lines - Train B

71111.15

Calculations

13-JC-EW-0201

TLU for Essential Chillers Cooling Water Local Flow

Indicators (F-0063 & F-0064)

71111.15

Calculations

13-MC-EC-0252

EC System Water Requirements and Chiller Sizing

71111.15

Calculations

13-MC-EC-0252

EC System Water Requirements and Chiller Sizing

71111.15

Calculations

13-MC-EC-0252

EC System Water Requirements and Chiller Sizing

71111.15

Calculations

13-MC-EW-0305

EW System Hydraulic Calculation

71111.15

Calculations

13-MC-PC-501

Fuel Pool Cooling & Cleanup System-Control to Fuel

Building

71111.15

Calculations

13-MC-SI-0250

Safety Injection, Containment Spray, and Shutdown Cooling

System Pump NPSH Evaluations

71111.15

Calculations

13-MC-SI-0506

Safety Injection System Train B

71111.15

Calculations

13-NC-ZC-0237

Maximum Passive Heat Sink for Hydrogen Generation &

ECCS Evaluation

71111.15

Corrective Action

20-12291, 23-05203, 23-05220, 23-05751, 23-06486, 23-

Documents

06639, 23-06673, 23-09185, 23-09453, 23-09591, 19-

03662, 19-04858, 19-14353, 19-14374, 23-03429, 23-

05881, 18-18088, 23-03012, 23-06274, 23-06913, 23-

09192, 23-08071, 23-10058, 22-09600, 23-11149, 23-

06705, 23-09690, 23-11758, 23-11775, 20-16080, 20-

16801, 23-12058, 23-12088, 23-12089, 23-09689, 23-

2388, 23-12394, 23-12870, 23-12802, 20-06122, 20-16655

71111.15

Drawings

01-M-RCP-001

P & I Diagram Reactor Coolant System

71111.15

Drawings

03-C-ZAS-0620

Shutdown Heat Exchanger Room A - EXCEL Platform 3

71111.15

Drawings

13-C-OOA-0003

Typical Stair and Miscellaneous Details

71111.15

Drawings

13-C-SPS-0375

Nuclear Service Spray Ponds Plan

71111.15

Drawings

13-C-SPS-0376

Nuclear Service Spray Ponds Sections and Details - Sheet

71111.15

Drawings

13-C-ZCS-0301

Containment Internals Partial Concrete Plan at El. 80'-0

Areas CAC and CAD

71111.15

Miscellaneous

System Training Manual Volume 72 Radioactive Waste

Drain System (RD)

71111.15

Miscellaneous

Operations Logs May 1 - May 25, 2023

09/27/2023

71111.15

Miscellaneous

U2C25 Core Data Book

71111.15

Miscellaneous

N001-0205-00197; Boric Acid Precipitation Reanalysis for

Palo Verde Nuclear Generating Station Units 1, 2, and 3

71111.15

Miscellaneous

PVGS Operator Information Manual

April 2020

71111.15

Miscellaneous

SA Design Basis Manual

71111.15

Miscellaneous

Engineering evaluation 21-03935-003

71111.15

Miscellaneous

Engineering evaluation 22-05045-002

71111.15

Miscellaneous

PVNGS DBM - Reactor Coolant System

71111.15

Miscellaneous

13-MM-0723

Water Chillers Quality Class Q and R

71111.15

Miscellaneous

13-MS-B108

Emergency Diesel Generator Upgrades Evaluations

71111.15

Miscellaneous

13-NS-C113

Additional Heat Sink Quantities for ECCS Performance

Analysis (13-NC-ZC-0237)

71111.15

Miscellaneous

2013-00269

Engineering Document Change

71111.15

Miscellaneous

DG-1153

Design Change Request

71111.15

Miscellaneous

N001-0101-00262

Design Reports for Borg Warner ASME Class 1 and Class 2

Valves

71111.15

Miscellaneous

RCGVS-PE-VR07

Valve Design Requirements for Reactor Coolant Gas Vent

System

71111.15

Miscellaneous

Specification 13-

JN-0699

Procurement of Nuclear Service ASME B & PV Code Class

Solenoid valves

71111.15

Miscellaneous

VTD-G063-00002

General Atomic (Sorrento Division) Instruction Manual for

BOP ESFAS

71111.15

Procedures

33ST-9HJ05

Control Room Air In-Leakage Tracer Gas Test

71111.15

Procedures

40DP-9OP26

Operations Operability Determination Process

71111.15

Procedures

40DP-9OP26-01

Operations Operability Determination Process

Administrative Guideline

71111.15

Procedures

40DP-9ZZ17

Control of Doors, Hatches and Floor Plugs

71111.15

Procedures

40OP-9DG01

Emergency Diesel Generator A

71111.15

Procedures

40OP-9EW01

Essential Cooling Water System (EW) Train A

71111.15

Procedures

40OP-9EW02

Essential Cooling Water System (EW) Train B

71111.15

Procedures

40OP-9HJ01

Control Building HVAC (HJ)

71111.15

Procedures

40OP-9SA01

BOP ESFAS Modules Operation

71111.15

Procedures

40OP-9ZZ05

Power Operations

151

71111.15

Procedures

73DP-0AP08

Control Room Envelope Habitability Program

71111.15

Procedures

73DP-9XI01

Pump and Valve Inservice Testing Program

71111.15

Procedures

73DP-9ZZ12

Motor Operated Valve (MOV) Program

71111.15

Procedures

73ST-9EW02

Essential Cooling Water Pumps - Comprehensive Pump

Test

71111.15

Procedures

73ST-9XI24

Reactor and Pressurizer vent Valves - Inservice Test

71111.15

Procedures

74ST-9SI01

Safety Injection Tank Boron Surveillance Test

71111.15

Procedures

81DP-0CC15

Engineering Work Orders and Evaluations

71111.15

Work Orders

5534545, 3798546, 4432112, 5554974, 3094267, 5456841,

5578491, 5582422, 5466262, 5427435, 5472350, 5450088

71111.20

Calculations

13-JC-SH-0200

Subcooled Margin Monitor Setpoint and Uncertainty

Calculation

71111.20

Corrective Action

Documents

23-10482, 23-10502, 23-10372, 23-10367, 23-10373, 23-

10338, 23-10336, 23-10254, 23-10262, 23-10206, 23-

10174, 23-10231, 23-10244, 23-10251, 23-10595, 23-

10624, 23-10604, 23-10638, 23-10959, 23-10962, 23-

11302, 23-11266, 23-08289, 23-09840, 23-10116, 23-

10165, 23-10187, 23-10277, 23-10280, 23-10976, 23-

09607, 23-10270, 23-10274, 23-12188

71111.20

Miscellaneous

APS Correspondence 162-06829-EFS, CRDR 9-Q226,

Inadequate Definition of Loops Filled-INPO OE 7189

06/02/1995

71111.20

Miscellaneous

ABB Combustion Engineering CENPSD-770, Analysis for

Lower Mode Functional Recovery Guidelines

71111.20

Miscellaneous

Appendix G, Protected Equipment Scheme, U1R24 PZR

Manway on -SDC Entry conditions Met

11/03/2023

71111.20

Miscellaneous

Unit ONE 24th Refueling Outage Shutdown Risk

Assessment Final Report

71111.20

Miscellaneous

Specific Maneuver Plan: EOC Shutdown 91% to 20% Unit 1

Cycle 24

71111.20

Miscellaneous

Shutdown Safety Function Assessment Pressurizer Manway

On, SDC Entry Conditions Met

10/07/2023

71111.20

Miscellaneous

Palo Verde Nuclear Generating Station (PVNGS) Unit 1

Core Operating Limits Report

71111.20

Miscellaneous

Unit 1 Cycle 24 72ST-9RX14, Shutdown Margin, Data

10/07/2023

71111.20

Miscellaneous

Anticipated Critical Position, 1 over M Evaluation

11/11/2023

71111.20

Miscellaneous

4758718

Engineering Evaluation

71111.20

Procedures

01DP-0AP17

Managing Personnel Fatigue

71111.20

Procedures

36MT-1RC03

Refueling Water Level Indicating System Instrumentation

Calibration - Train A

71111.20

Procedures

36ST-9SI10

Remote Shutdown Monitoring System Instrumentation

Calibration for the Train B SI System

71111.20

Procedures

40AO-9ZZ23

Loss of SFP level or Cooling

71111.20

Procedures

40EP-9EO11

Lower Mode Functional Recovery

71111.20

Procedures

40OP-9AP19

Lower Mode Functional Recovery Technical Guideline

71111.20

Procedures

40OP-9SI01

Shutdown Cooling Initiation

71111.20

Procedures

40OP-9ZZ02

Initial Reactor Startup Following Refuelings

71111.20

Procedures

40OP-9ZZ04

Plant Startup Mode 2 to Mode 1

71111.20

Procedures

40OP-9ZZ10

Mode 3 to Mode 5 Operations

71111.20

Procedures

40OP-9ZZ11

Mode Change Checklist

106

71111.20

Procedures

40OP-9ZZ16

RCS Drain Operations

71111.20

Procedures

40OP-9ZZ23

Outage GOP

71111.20

Procedures

40OP-9ZZ26

Tracking Containment Penetrations

71111.20

Procedures

40ST-9RC01

RCS and Pressurizer Heatup and Cooldown Rates

71111.20

Procedures

40ST-9RC03

RCS Loop/SDC Tain Weekly Breaker Alignment and Power

Availability Surveillance in MODEs 3, 4, 5, and 6

71111.20

Procedures

40ST-9ZZ10

Post Accident Monitoring Instrumentation Channel Checks

71111.20

Procedures

70DP-0RA01

Shutdown Risk Assessments

71111.20

Procedures

2IC-9RX03

Core Reloading

71111.20

Procedures

2OP-9RX01

Calculation of Estimated Critical Condition

71111.20

Procedures

2ST-9RX14

Shutdown Margin - Modes 3, 4, and 5

71111.20

Procedures

78OP-9FX01

Refueling Machine Operations

71111.24

Calculations

01-EC-MA-0221

AC Distribution

71111.24

Calculations

13-EC-PE-0124

Diesel Generator Frequency Meter Loop E-PEN-SI-

G01/G02 Uncertainty Calculation

71111.24

Calculations

13-JC-EW-0203

Essential Cooling Water System (ECWS) Pump Discharge

Pressure Instrument (EWN-PI-0009 & EWN-PI-0010)

Uncertainty Calculation

71111.24

Calculations

13-MC-AF-0800

Auxiliary Feedwater ESF Function Response Times

71111.24

Calculations

13-MC-SP-0307

SP/EW System Thermal Performance Design Bases

Analysis

71111.24

Calculations

13-NC-ZC-0206

LOCA Containment P-T Transient Analysis

71111.24

Corrective Action

Documents

23-09943, 23-09948, 23-10107, 20-14414, 23-11321, 23-

10198, 22-11747, 23-09771, 23-09783, 23-10207, 23-

222, 23-10244, 23-10377, 23-10414, 23-10524, 23-

10656, 23-11190, 23-11975, 23-11938

71111.24

Drawings

01-M-RCP-001

P and I Diagram Reactor Coolant System

71111.24

Drawings

01-M-SIP-002

P and I Diagram Safety Injection and Shutdown Cooling

System

71111.24

Drawings

13-N001-1101-

00086

Containment Spray Pump Section V-CE-4576 18JA78

71111.24

Drawings

N001-1306-00084

ESFAS Aux Relay Cabinet Wire List

71111.24

Miscellaneous

Report of Calibration: Heise HQS-2 Heise Pressure Module

OP3112

07/31/2023

71111.24

Miscellaneous

Essential Cooling Water System Design Basis Manual

71111.24

Miscellaneous

PVNGS DBM - Safety Injection System

71111.24

Miscellaneous

01-M-HJP-001

CONTROL BUILDING HVAC P & I Diagram

71111.24

Miscellaneous

01-P-ZJC-308

Control Building HVAC Details

71111.24

Miscellaneous

13-JC-ZZ-0504

Motor Operated valve Torque Calculation for 13JSIAH0657,

13JSIBH0658, 13JSIAH0678, and 13JSIBH0679

71111.24

Miscellaneous

PVN2301R0-L

ISG Testing at Palo Verde Unit 1

71111.24

Miscellaneous

VTD-R411-0004

RUSKIN MANUFACTURING COMPANY OPERATION-

MAINTENANCE INSTRUCTIONS AND SPARE PARTS

LIST FOR BUBBLETIGHT DAMPERS

71111.24

Procedures

31MT-9SI03

Containment Spray Pump Disassembly and Assembly

71111.24

Procedures

2MT-9DG03

Train A Integrated Safeguards (ISG) Testing

71111.24

Procedures

2ST-9SB02

Month Surveillance Test for Westinghouse Type DS-416

Reactor Trip Breakers

71111.24

Procedures

33MT-9HF01

Fuel and Auxiliary Building Normal Ventilation System Train

A Pneumatic and Electrical Jumper Installation

71111.24

Procedures

33ST-9HJ01

Control Room AFU Airflow Capacity and Pressurization Test

71111.24

Procedures

36ST-9SA01

ESFAS Train A Subgroup Relay Functional Test

71111.24

Procedures

36ST-9SA03

ESFAS Train A Subgroup Relay Shutdown Functional Test

71111.24

Procedures

36ST-9SB52

RTSG Shunt and Undervoltage Trip Functional Test

71111.24

Procedures

39MT-9ZZ05

Refurbishment of Limitorque SMB-000 Motor Operated

Valves

71111.24

Procedures

40AL-9RK2B

Panel B02B Alarm Responses

71111.24

Procedures

40OP-9DG01

Emergency Diesel Generator A

71111.24

Procedures

40OP-9DG03

Conduct of Train A Integrated Safeguards (ISG) Testing

71111.24

Procedures

40OP-9EC01

Essential Chilled Water Train A

71111.24

Procedures

40OP-9EW01

Essential Cooling Water System (EW) Train A

71111.24

Procedures

40OP-9OP43

M&TE Gauge Installation

71111.24

Procedures

40OP-9SI03

Safety Injection Tank Operations

71111.24

Procedures

40ST-9DG01

Diesel Generator A Test

71111.24

Procedures

73DP-9CL02

Containment Leakage Rate Testing Program

71111.24

Procedures

73DP-9XI05

Check Valve Condition Monitoring Program

71111.24

Procedures

73DP-9ZZ12

Motor Operated Valve (MOV) Program

71111.24

Procedures

73ST-9SI15

Containment Spray Pumps - Comprehensive Test

71111.24

Procedures

73ST-9CL01

Containment Leakage Type B and C Testing

71111.24

Procedures

73ST-9DG01

Class 1E Diesel Generator and Integrated Safeguards Test

Train A

71111.24

Procedures

73ST-9DG01

Class 1E Diesel Generator and Integrated Safeguards Test

Train A

34a

71111.24

Procedures

73ST-9EC01

Essential Chilled Water Pumps - Inservice Test

71111.24

Procedures

73ST-9EW01

Essential Cooling Water Pumps - Inservice Test

71111.24

Procedures

73ST-9SI03

Leak Test of SI/RCS Pressure Isolation Valves

71111.24

Procedures

73ST-9XI03

SI Train A Valves - Inservice Test

71111.24

Procedures

73ST-9XI05

AF Valves - Inservice Test

71111.24

Procedures

73ST-9XI05

AF Valves - Inservice Test

35a

71111.24

Procedures

Containment

Leakage Type B

and C Testing

Containment Leakage Type B and C Testing

2a

71111.24

Work Orders

5454594, 5209620, 5456512, 5372007, 5541528, 5371996,

5377723, 5403932, 5456509, 5456511, 5290782, 5456849,

5456935, 5305407, 5456843, 5456828, 5459253, 5570107,

5458803, 5456814, 5479577, 5456802

71114.06

Calculations

13-NC-CH-0310

Post Accident Radiation Levels at RU-150 and RU-151

71114.06

Corrective Action

Documents

23-09503, 23-09504, 23-09749, 23-09818, 23-09921, 23-

10098, 23-11673

71114.06

Miscellaneous

Emergency Preparedness Scenario 2305 Drill

09/08/2023

71114.06

Miscellaneous

Emergency Preparedness 2305 & 2306 Drill Report - Drill

Dates Sep 19th and Sep 26th 2023

11/03/2023

71114.06

Miscellaneous

Core Damage Assessment User Manual

71114.06

Miscellaneous

EP-0801H

[Emergency Action Level] Classification Matrix Hot

Conditions ([Reactor Coolant System] > 210°F)

71114.06

Miscellaneous

EP-0803H

[Protective Action Recommendation] Flowchart

71114.06

Miscellaneous

EP-0804G

Release Evaluation Flowchart

71114.06

Miscellaneous

EP-0930B

[Emergency Action Level] Classification Matrix All

Conditions

71124.05

Calibration

Records

Thermo fisher scientific ipcm-12 calibration cover sheet

SN#12025

09/07/2023

71124.05

Calibration

Records

Thermo fisher scientific PM12 calibration cover sheet SN#

216

2/08/2022

71124.05

Calibration

Records

Thermo fisher scientific FH40G-L internal detector

calibration sheet, SN#25611

06/21/2023

71124.05

Calibration

Records

Calibration verification data sheet for fastscan whole body

counter #2

06/10/2022

71124.05

Calibration

Records

Calibration data sheet for Eberline RM-20 SN#1612

07/07/2023

71124.05

Calibration

Records

Alpha beta sample counter SN#339533

10/14/2023

71124.05

Calibration

Records

Thermo scientific model RO-20 ion chamber SN# 11367

03/01/2023

71124.05

Calibration

Records

Thermo scientific model RO-20 ion chamber SN# 1426

2/02/2023

71124.05

Calibration

Records

Small article monitor SN#669

01/27/2023

71124.05

Corrective Action

Documents

CR # XX-XXXXX

21-06265, 21-07409, 21-09419, 21-09942, 22-00171,

2-10743, 23-07845, 23-08098, 32-08852, 23-09225,

23-09369, 23-09628

71124.05

Miscellaneous

Technical specifications and offsite dose calculation manual

radiation monitors information reference chart

10/13/2023

71124.05

Self-Assessments 2023-004

Nuclear assurance department audit 2023-004 radiation

safety

06/30/2023

71124.05

Work Orders

Work order

1. XXXXXXX

5459402, 5459503, 5259139, 5111180, 5449700, 5110593,

273241

71124.08

Corrective Action

Documents

21-00603, 21-00604, 21-01091, 21-01100, 21-02040, 21-

03350, 21-07861, 21-08704, 21-10518. 22-03452, 22-

05490, 22-08014, 22-08573, 22-13359, 23-01794, 23-

01795, 23-01940

71124.08

Corrective Action

Documents

Resulting from

Inspection

23-12205

71124.08

Miscellaneous

Palo Verde Nuclear Generation Station Part 37 Plan for The

Protection of Category 1 and Category 2 Quantities of

Radioactive Material

71124.08

Miscellaneous

19-03441-001

Methodology for Detecting Hard-to-Detect Radionuclides in

11/03/2023

Waste Oil

71124.08

Miscellaneous

218-04875

23 Annual Source Inventory

2/14/2023

71124.08

Miscellaneous

NBA11C000115

Radiation Protection Technician Training Program:

Packaging Radioactive Material

09/09/2021

71124.08

Miscellaneous

NBA16C000106

Radiation Protection Technician Training Program:

Processing Filters

71124.08

Miscellaneous

NBA19C000213

Radiation Protection Technician Training Program: Shipping

Radioactive Material

05/13/2021

71124.08

Miscellaneous

NBA20C000107

Radiation Protection Technician Training Program:

Operating DAWPS Facility

09/03/2019

71124.08

Procedures

20DP-0SK37

Palo Verde 10 CFR Part 37 Security Program Overview

71124.08

Procedures

20DP-0SK77

Security Miscellaneous Testing

71124.08

Procedures

20SP-0SK25

Security Patrols

71124.08

Procedures

20SP-0SK50

Law Enforcement Response Plan

71124.08

Procedures

75RP-9RP15

Control and Storage of Radioactive Material and Radioactive

Wastes

71124.08

Procedures

75RP-9RP26

Radioactive Source Control

71124.08

Procedures

75ST-9ZZ02

Radioactive Source Leak Test Surveillance

71124.08

Procedures

76DP-0AP12

Low Level Radioactive Material Storage Facility Overview

71124.08

Procedures

76DP-0AP12

Low Level Radioactive Material (LLRM) Storage Facility

Overview

71124.08

Procedures

76DP-0RP03

Radwaste Process Control Program

71124.08

Procedures

76RP-0RW03

Waste Stream Sampling and Database Maintenance

71124.08

Procedures

76RP-0RW04

Receipt of Radioactive Material

71124.08

Procedures

76RP-0RW05

Packaging and Classification of Radioactive Waste

71124.08

Procedures

76RP-0RW06

Packaging of Radioactive Material

71124.08

Procedures

76RP-0RW07

Shipping Radioactive Materials

71124.08

Procedures

76RP-0RW08

High Integrity Container Receipt, Handling, Use, and

Closure

71124.08

Radiation

Surveys

0-M-20230927-2

Low Level Radioactive Material Storage Facility: monthly

survey

09/27/2023

71124.08

Radiation

Surveys

0-M-20230929-2

Dry Active Waste Processing and Storage Facility - Storage

Area Monthly Routine Survey

09/29/2023

71124.08

Radiation

Surveys

0-M-20230929-3

Dry Active Waste Processing and Storage Facility - Storage

Area Monthly Routine Survey

09/29/2023

71124.08

Radiation

Surveys

1-M-20200807-1

Radwaste 112'/100' High Level Storage Area

08/07/2020

71124.08

Radiation

Surveys

1-M-20231017-6

Radwaste Truck Bay/ Low Level Storage Area 100' -

Routine Monthly

10/17/2023

71124.08

Radiation

Surveys

2-M-20170507-2

Radwaste 112' High Level Storage Area Verification CR#

17-06898

05/07/2017

71124.08

Radiation

Surveys

3-M-20231017-2

Radwaste Truck Bay/ Low Level Storage Area

10/17/2023

71124.08

Self-Assessments 2021-004

Nuclear Assurance Department (NAD) Audit Report:

Radiation Safety

08/02/2022

71124.08

Self-Assessments 2023-004

Nuclear Assurance Department (NAD) Audit - Radiation

Safety

07/30/2023

71124.08

Self-Assessments CR-22-00214

Security Identified - Requesting a Simple Self-Assessment

on 10CFR 37.55(a)

71124.08

Self-Assessments CR-23-00383

Security Identified - Conduct a Simple Self-Assessment on

10CFR Part 37

71124.08

Work Orders

5378617

75ST-9ZZ02, Radioactive Source Leak Test Surveillance

01/17/2023

71124.08

Work Orders

5433339

75ST-9ZZ02, Radioactive Source Leak Test Surveillance

07/27/2023

71151

Miscellaneous

Units 1, 2, and 3 PI Summary Report BI01-02 10/2022-

09/2023

10/19/2023

71151

Miscellaneous

Units 1, 2, and 3 RCS leakage data 10/2022-9/2023

71151

Procedures

40DP-9LC01

Performance Indicator - Barrier Integrity Cornerstone: RCS

Leak Rate

71151

Procedures

74DP-0LC01

RCS Activity Performance Indicator

71151

Procedures

93DP-0LC09

NRC ROP PI Data Collection, Verification and Submittal

71152A

Corrective Action

Documents

15-01314, 16-04187, 19-01360, 22-06393, 23-10957, 23-

10502, 23-11634, 23-10162, 23-07440, 23-07442, 23-

07500, 23-07520, 23-12173, 23-12760, CRDR 3144389, 23-

10834, 23-11614

71152A

Miscellaneous

CRDR 4636321

71152A

Miscellaneous

Material engineering evaluation 04569, Replace Pulse Tone

Piston with Type 304 Stainless Steel in place of 303

Stainless Steel

71152A

Miscellaneous

ERET 2952421

71152A

Miscellaneous

CRAI 4645846

71152A

Miscellaneous

PVGS Diverse and Flexible Coping Strategies (FLEX)

Program

71152A

Miscellaneous

13-C-ZYS-674

Miscellaneous Yar Structures Sheet 21

71152A

Miscellaneous

19-01360-065

Engineering Evaluation

71152A

Miscellaneous

AN1000-A00025

FLEX Equipment Deployment Slabs/Pads and Equipment

Tie-Downs Evaluation (Seismic and Wind)

71152A

Miscellaneous

N001-0701-00023

INSTALL AND ASSY 5 GAL 3010 PULSETONE V-VE-8944

14SE79 C/L 28

71152A

Miscellaneous

NM1000-A00032

Spent Fuel Pool Cooling FLEX Pump NPSH Availability

71152A

Miscellaneous

SPEC 13-PN-

0410

General Maintenance Practices

71152A

Procedures

14DP-0BD02

PVNGS FLEX Equipment Status Control

71152A

Procedures

20DP-0SK20

Conduct of Security

71152A

Procedures

30DP-9MP03

Mechanical System Cleanliness

71152A

Procedures

40DP-9OP02

Conduct of Operations

71152A

Procedures

40MG-9ZZ07-002

FLEX Support Guidelines MODE 5, 6, or Defueled

71152A

Procedures

40OP-9CH13

Charging Pump Suction Stabilizer and Discharge Pulsation

Dampener Operation

71152A

Procedures

Sec DI-02

Use of Air Rifles for Pigeon Control

71152A

Procedures

Sec DI-02

Use of Air Rifles for Pigeon Control

71152A

Work Orders

5571374, 5573609, 5569781, 55551182, 5431980,

5431989, 5573223

71152S

Corrective Action

Documents

23-03332, 23-03380, 23-04015, 23-04085, 23-04095, 23-

05193, 23-05203, 23-06120, 23-06159, 23-06225, 23-

06328, 23-06541, 23-06596, 23-07044, 23-07074, 23-

07081, 23-07760, 23-07972, 23-08288, 23-08623, 23-

08693, 23-08851, 23-09316, 23-09453, 23-10425, 23-

10533, 23-11514, 23-12196, 23-12808, 23-12816

71152S

Miscellaneous

Nuclear Assurance Department Audit Report 2023-002 -

Engineering Programs

71152S

Miscellaneous

Nuclear Assurance Department Audit Report 2023-003 -

Fire Protection

71152S

Miscellaneous

Nuclear Assurance Department Audit Report 2023-006 -

Design Control & Configuration Management

71152S

Miscellaneous

2017-Current Summary of Detail engineering Change

Quality Metric

11/16/2023

71152S

Miscellaneous

23 10 October - Modifications Engineering PI Report

11/16/2023

71152S

Miscellaneous

NA-02-C25-2022-

2

Unit 2 Cycle 25 Framatome Checklist (Proprietary)

71152S

Procedures

01DP-0AP12

Condition Reporting Process

71152S

Procedures

01DP-0AP12

Condition Reporting Process

71152S

Procedures

01DP-0AP12-01

Condition Reporting Administrative Guideline

71152S

Procedures

01DP-0AP12-01

Condition Reporting Administrative Guideline

71152S

Procedures

01DP-0AP59-07

Trend Analysis Administrative Guideline

71152S

Procedures

01DP-0AP59-10

Self-Assessment and Benchmarking Administrative

Guideline

71152S

Procedures

60DP-0QQ19

Internal Audits

71153

Corrective Action

Documents

17-01395, 19-11892, 23-03479, 23-05265, 23-05983,

2967441, 3201838, 3202178, 3599300

71153

Drawings

01-E-MAB-008

Elementary Diagram: Main Generation System Generator &

Transformer Primary Protection Unit Tripping

71153

Drawings

01-E-NAB-012

Elementary Diagram: 13.8kV Non-class 1E Power System

Buses 1E-NAN-S01 & 1E-NAN-S02 Unit Aux XFMR

Breakers Control

71153

Drawings

01-E-NAB-014

Elementary Diagram: 13.8kV Non-class 1E Power System

Buses 1E-NAN-S03 & 1E-NAN-S04 Bus Tie Breakers

Control

71153

Drawings

01-J-ZZL-020

Control Logic Diagram: Turbine Trip Logic

71153

Drawings

01-J-ZZL-021

Control Logic Diagram: Reactor Trip Logic

71153

Drawings

13-E-MAD-001

Logic Diagram: Generator Tripping

71153

Miscellaneous

PVNGS Design Basis Manual - 13.8kV AC Non-class 1E

Power System

71153

Miscellaneous

PVNGS Design Basis Manual - Electrical Topical

71153

Miscellaneous

PVNGS Design Basis Manual - Main Generation System

71153

Miscellaneous

PVNGS Design Basis Manual - Offsite Power

71153

Miscellaneous

PVNGS System Training Manual - 13.8kV AC Non-class 1E

Power System

71153

Miscellaneous

PVNGS System Training Manual - Main Generation System

71153

Miscellaneous

PVNGS System Training Manual - Main Generator

Excitation and Regulation System

71153

Miscellaneous

Plant Review Board Meeting 23-002 Presentation

04/12/2023

71153

Miscellaneous

Plant Review Board Initial Documentation

04/11/2023

71153

Miscellaneous

Plant Transient Review Assessment

04/09/2023

71153

Miscellaneous

13-EN-0306

Installation Specification for Cable Splicing and Terminations

for the Palo Verde Generating Station Units 1, 2, and 3

Quality Class Q, QAG, and NQR

71153

Miscellaneous

PR-0201

Maintenance Program Requirement Description

71153

Miscellaneous

VTD-G080-00016

General Electric Polyphase Power Directional Relay for Anti-

motoring Protection Type GGP53C [PUB. #GEK-34117G]

71153

Procedures

01DP-9ZZ01

Systematic Troubleshooting

71153

Procedures

01DP-9ZZ01

Systematic Troubleshooting

2a

71153

Procedures

30DP-9MP01

Conduct of Maintenance

71153

Procedures

30DP-9MP01

Conduct of Maintenance

71153

Procedures

30DP-9MP27

Toolpouch and Minor Maintenance Process

71153

Procedures

30DP-9MP27

Toolpouch and Minor Maintenance Process

71153

Procedures

30DP-9MP27

Toolpouch and Minor Maintenance Process

1a

71153

Procedures

40AO-9ZZ03

Loss of Cooling Water

71153

Procedures

40AO-9ZZ03

Loss of Instrument Air

71153

Procedures

40AO-9ZZ05

Loss of Charging or Letdown

71153

Procedures

40AO-9ZZ08

Load Rejection

71153

Procedures

40AO-9ZZ12

Degraded Electrical Power

71153

Procedures

40AO-9ZZ12

Degraded Electrical Power

71153

Procedures

40EP-9EO01

Standard Post Trip Actions

71153

Procedures

40EP-9EO02

Reactor Trip

71153

Procedures

40EP-9EO07

Loss of Offsite Power / Loss of Forced Circulation

71153

Procedures

40EP-9EO10-051

Appendix 51: Electric Plant Single Line Diagram

71153

Procedures

93DP-0LC12

Administrative Control and Compliance of NERC Standards

2