IR 05000529/2024011
| ML24345A055 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 12/11/2024 |
| From: | Greg Warnick NRC/RGN-IV/DORS/EB2 |
| To: | Heflin A Arizona Public Service Co |
| Makor S | |
| References | |
| IR 2024011 | |
| Download: ML24345A055 (1) | |
Text
December 11, 2024
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION - LICENSE RENEWAL INSPECTION REPORT 05000529/2024011
Dear Adam Heflin:
On October 30, 2024, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Palo Verde Nuclear Generating Station and discussed the results of this inspection with Senior Vice Presidents Cary Harbor and Todd Horton and other members of your staff. The results of this inspection are documented in the enclosed report.
No findings or violations of more than minor significance were identified during this inspection.
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, Gregory G. Warnick, Chief Engineering Branch 2 Division of Operating Reactor Safety Docket No. 05000529 License No. NPF-51
Enclosure:
As stated
Inspection Report
Docket Number:
05000529
License Number:
Report Number:
Enterprise Identifier:
I-2024-011-0008
Licensee:
Arizona Public Service
Facility:
Palo Verde Nuclear Generating Station
Location:
Tonopah, AZ
Inspection Dates:
October 21, 2024, to October 30, 2024
Inspectors:
S. Makor, Senior Reactor Inspector
N. Okonkwo, Reactor Inspector
Approved By:
Gregory G. Warnick, Chief
Engineering Branch 2
Division of Operating Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a NRC 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
No findings or violations of more than minor significance were identified.
Additional Tracking Items
None.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel to assess licensee performance and compliance with Commission rules and regulations, license conditions, site procedures, and standards.
OTHER ACTIVITIES
- TEMPORARY INSTRUCTIONS, INFREQUENT AND ABNORMAL
===71003 - Post-Approval Site Inspection for License Renewal The inspectors evaluated the material condition of Palo Verde Nuclear Generating Station (PVNGS) in October 2024 while the plant was shut down for refueling outage 2R25. This allowed the inspectors to evaluate the material condition of inaccessible areas prior to entry into the period of extended operation and to evaluate the licensee implementation of aging management activities. The period of extended operation is the additional 20 years beyond the original 40-year licensed term and begins after midnight on June 1, 2025.
In addition, the inspectors evaluated whether the licensee:
- (1) completed the necessary actions to comply with the license condition and commitments related to aging management; and
- (2) implemented programs that agreed with those approved in the safety evaluation report and described in the updated final safety analysis report. NRC issued the safety evaluation report in NUREG-1915, Safety Evaluation Report Related to the License Renewal of Palo Verde Nuclear Generating Station (ML11095A011). Specific activities evaluated during this inspection are described in the following paragraphs.
Post-Approval Site Inspection for License Renewal===
- (1) A1 - Quality Assurance and Commitment 2 The quality assurance aging management program is an existing program that includes the elements of corrective action, confirmation process, and administrative controls. The PVNGS quality assurance program is applicable to all safety-related and nonsafety-related systems, structures, and components (SSCs) that are subject to aging management activities.
Commitment 2 specified:
- The existing Quality Assurance program is credited for license renewal.
The inspectors interviewed the program owners and sampled revisions to procedures listed in the controlled documents list in the references. The inspectors also reviewed implementing procedures, evaluated program elements, reviewed operating experience, and corrective actions. The inspectors discussed the procedures to be enhanced to include those in-scope nonsafety-related SSCs requiring aging management to address the elements of corrective actions, confirmation process, and administrative controls with the program owner.
The inspectors determined that the licensee had implemented the administrative requirements and met the conditions of Commitment 2.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (2) A1.5 - Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors and Commitment 7 This existing program manages cracking caused by primary water stress corrosion cracking (PWSCC) and loss of material resulting from boric acid wastage in nickel-alloy pressure vessel head penetration nozzles. The affected components include the reactor vessel closure head, upper vessel head penetration nozzles and associated welds. Detection of cracking is accomplished through implementation of a combination of bare metal visual examination (external surface of head) and surface and volumetric examination (underside of head) techniques that are consistent with ASME Code Case N-729-6, subject to the conditions specified in 10 CFR 50.55a. The program conducts bare metal examinations (external surface of head) to detect leakage from PWSCC and other sources by looking for deposition of boric acid.
Commitment 7 specified:
- The existing Nickel-Alloy Penetration Nozzles Welded to the Upper Reactor Vessel Closure Heads of Pressurized Water Reactors is credited for license renewal.
The inspectors reviewed implementing documents, scope of the program, and completed inspections. From this review, the inspectors determined the licensee replaced the reactor vessel heads in all three units with heads fabricated with Alloy 690 nozzle material and welded using Alloy 52 filler material. Since the head replacements, the licensee has not identified any indications on the Alloy 690 nozzles and associated welds. Additionally, no evidence of PWSCC in the vessel head penetration nozzles have been indicated except for vent line indications on Unit 2, which the licensee machined and weld overlaid during refueling outage U2R13. The inspectors determined that the licensee effectively implemented the requirements of this program and met the requirements of Commitment 7.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (3) A1.8 - Steam Generator Tube Integrity and Commitment 10 This existing program evaluates tube integrity in accordance with the structural integrity performance criteria in technical specifications, which encompass and exceeds the requirements of Regulatory Guide 1.121, Bases for Plugging Degraded PWR Steam Generator Tubes. The steam generator management practices are consistent with NEI 97-06, Steam Generator Program Guidelines. No exceptions or enhancements identified for this program.
Commitment 10 specified:
- The existing Steam Generator Integrity program is credited for license renewal with no exceptions or enhancements.
The inspectors interviewed the program owners and sampled revisions to procedures listed in the controlled documents list in the references. The inspectors reviewed the self-assessments/metrics and the program health report with no adverse findings in the 2020 assessment and acceptable green window respectively. The inspectors reviewed relevant operating experience, program elements, remaining pre-period of extended operation actions, and corrective actions.
The inspectors discussed with site personnel Commitment 61 and 62 associated with the Steam Generator Integrity program and the overlap with Commitment 10.
Commitment 61 related to potential PWSCC at divider plate assembly components.
The licensee determined that the potential failure of the steam generator reactor coolant system pressure boundary due to PWSCC cracking of steam generator divider plate bar welds and the divider plate bars in Unit 2 is not a credible concern.
Commitment 62 related to the potential failure of the steam generator Primary-to-Secondary pressure boundary at tube-to-tubesheet welds. The licensee performed general visual inspections of the accessible tubesheet region looking for evidence of cracking during refueling outage 3R24. The inspections identified no tubesheet cracking or rust stains were identified. The inspectors identified no concerns with the existing program and implementation of these commitments to conclude that the licensee met the requirements of Commitment 10.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (4) A1.9 - Open-Cycle Cooling Water System and Commitment 11 This existing program manages loss of material and reduction of heat transfer for components exposed to the raw water in the open-cycle cooling water system.
Commitment 11 specified:
- The existing Open-Cycle Cooling Water System is credited for license renewal and prior to the period of extended operation the program will be enhanced to clarify guidance in the conduct of heat exchanger inspections and piping inspections using nondestructive examination techniques and related acceptance criteria.
The inspectors interviewed the program owners and sampled revisions to procedures listed in the controlled documents list in the references. The inspectors also reviewed implementing procedures, evaluated program elements, reviewed operating experience, testing, work orders, system health reports, and corrective actions.
There were no exceptions for this program and one enhancement. The licensee identified an enhancement to the detection of aging effects and acceptance criteria program elements. The licensee will conduct heat exchanger and piping inspections using nondestructive examination techniques and the related acceptance criteria during onsite procedures before the period of extended operation. The inspectors determined the licensee met Commitment 11 and implemented actions to manage the effects of aging.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (5) A1.10 - Closed-Cycle Cooling Water System and Commitment 12 This existing program manages loss of material, cracking, and reduction in heat transfer for components in closed cycle cooling water systems. The program includes maintenance of system corrosion inhibitor concentrations and chemistry parameters following the guidance of EPRI TR-107396, Closed Cooling Water Chemistry Guideline, to minimize aging, and periodic testing and inspections to evaluate system and component performance. Inspection methods include visual, ultrasonic testing and eddy current testing.
Commitment 12 specified:
- The existing Closed-Cycle Cooling Water System program is credited for license renewal.
The inspectors interviewed the program owners and sampled revisions to procedures listed in the controlled documents list in the references. The inspectors also reviewed implementing procedures, evaluated program elements, reviewed operating experience, and corrective actions.
The inspectors reviewed the updated final safety analysis report (UFSAR)supplement, the aging management program basis document, implementing procedures, work orders, the regulatory commitment tracking system action item (RCTSAI) for the program to verify completed actions for the commitment activities.
The inspectors determined that this program will continue to manage the effects of aging during the period of extended operation. The inspectors determined that the licensee implemented appropriate aging management activities and concluded that the licensee met Commitment 12.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (6) A1.13 - Fire Water System and Commitment 15 This existing program manages the aging of fire water piping, piping components, pump casings, sprinkler heads, tanks, fittings, valve bodies, hose stations, and hydrants. The program focuses on the loss of material caused by general, pitting, crevice, and galvanic corrosion, microbiological corrosion, or biofouling of carbon steel, stainless steel, cast-iron, copper, bronze, brass, galvanized, and ductile iron components in the water-based fire protection systems, including water-based piping that is normally drained. Periodic fire water component inspections, tests and fire main flushing, sprinkler inspections, and flow tests are performed in accordance with National Fire Protection Association (NFPA) 25, Standard for the Inspection, Testing, and Maintenance of Water-Based Fire Protection Systems, 2002 Edition, to ensure that the water-based fire protection systems can perform their intended function.
Commitment 15 specified:
The existing fire water system program is credited for license renewal. Prior to the period of extended operation, the following enhancements will be implemented:
- Review and approve using the nuclear administrative technical manual.
- Procedures will be enhanced to be consistent with the current code of record or NFPA 25-2002 Edition.
- Prior to 50 years replace the sprinklers or enhance or field service test a representative sample and test thereafter every 10 years to ensure that signs of degradation are detected in a timely manner.
- Revise procedures to be consistent with NFPA 25 Sections 7.3.2.1, 7.3.2.2, 7.3.2.3, and 7.3.2. 4.
The inspectors reviewed the UFSAR supplement, aging management program basis document, implementing procedures describing required tests and inspections, work orders, as well as some corrective actions associated with fire water component failures. The inspectors interviewed the program owner and compared the interim staff guidance LR-ISG-2012-02, Aging Management of Internal Surfaces, Fire Water Systems, Atmospheric Storage Tanks, and Corrosion Under Insulation, to plant procedures. The inspectors determined that licensee included enhancements recommended in LR-ISG-2012-02. The inspectors verified the licensee completed the actions related to RCTSAI 3246902. The licensee had enhanced their procedures; however, the RCTSAI remains open until the licensee completes actions to develop plans and work orders for the 50-year field service test representative sprinkler sample or replaces all their sprinkler heads.
Based on the remaining actions related to the sprinkler systems, this program and commitment remain open.
- (7) Reactor Vessels Internals and Commitment 23 This new program addresses reactor vessel internals aging concerns and relies on EPRI MRP-227-A, Materials Reliability Program: Pressurized Water Reactor Internals Inspection and Evaluation Guidelines, and MRP-228, Materials Reliability Program: Inspection Standard for PWR Internals, to manage the aging effects on the reactor vessel internal components, including:
a) Various forms of cracking, including stress corrosion cracking, PWSCC, irradiated-assisted stress corrosion cracking, or cracking due to fatigue/cyclical loading.
b) Loss of material induced by wear.
c)
Loss of fracture toughness due to neutron irradiation embrittlement, or void swelling.
d) Dimensional changes due to void swelling and irradiation growth Loss of preload due to thermal and irradiation-enhanced stress relaxation or creep.
Commitment 23 specified:
- Manage the reactor vessel internals inspections during the period of extended operation through the augmented inservice inspection program.
The inspectors reviewed the UFSAR supplement, the aging management program basis document, implementing procedures, work orders, and the RCTSAI for this program to verify completed actions for the commitment activities. The inspectors interviewed the program owner and discussed the replacement if the Reactor Vessel Heads that now contain Alloy 690 and the schedule for future examinations to be performed for the first time. The inspectors determined that the licensee implemented appropriate aging management activities and concluded that the licensee met Commitment 23.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (8) A1.31 - Masonry Wall Program & Commitment 33 This existing program, implemented as part of the structures monitoring program, manages cracking, missing or broken locks, spalling and mortar deterioration of masonry walls. This program required no enhancements or exceptions and incorporates the guidance provided in Bulletin 80-11, Masonry Wall Design and Information Notice 87-67, Lessons Learned from Regional Inspections of Licensee Actions in Response to NRC IE Bulletin 80-11.
Commitment 33 specified:
- Existing masonry wall program is credited for license renewal. Prior to the period of extended operation, revise procedures to specify ACI 349.3R-96, Evaluation of Existing Nuclear Safety-Related Concrete Structures, as the reference for qualification of personnel to inspect structures under the masonry wall program.
The inspectors reviewed the UFSAR supplement, the aging management program basis document, implementing procedures, work orders, and completed inspection results. The inspectors interviewed the program owner and walked down selected masonry block walls in the turbine building, control building, auxiliary building and pump house building. The inspectors selected a sample of walls to confirm the licensee had appropriately reinforced the walls as required to withstand a seismic event. The inspectors confirmed that the licensee had included specific masonry wall inspection acceptance and personnel qualification criteria in procedure 81DP-0ZZ01, Civil, System, Structure and Component Monitoring Program, specified in ACI 349.3R-96.
The inspectors observed some cracks at the fire water pump house masonry wall in the license renewal program not documented by the licensee. The licensee initiated condition report 24-11778 to analyze the cracks and evaluate the structural integrity of the affected masonry walls. The inspectors also reviewed the RCTSAI for the program to verify actions on the commitment activities. The inspectors determined the licensee had implemented the Commitment 33 actions.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (9) Regulatory Guide 1.127, Inspection of Water-Control Structures Associated with Nuclear Power Plants (X1.S7) and Commitment 35 This program manages the effects of aging of concrete resulting from cracking, spalling, rust bleeding or stains, damaged concrete, abrasion, indication of water infiltration, and settlement issues. For steel components the program manages the effects of aging caused by corrosion The licensee performs periodic visual examinations to monitor the condition of water-control structures and structural components, including structural steel and structural bolting.
Commitment 35 specified:
- Prior to the period of extended operation, enhance procedures to specify that the essential spray ponds inspections include concrete below the water level.
The inspectors reviewed the UFSAR supplement, the aging management program basis document, implementing procedures, work orders, the RCTSAI for the program to verify completed actions for the commitment activities. The inspectors interviewed the program owner, walked down the spray pond and reviewed the results from work orders that performed the walkdown. The inspectors also reviewed the implementing procedure 81DP-0ZZ01 to confirm the licensee specified that the essential spray pond inspections include concrete below the water level. During the walkdown, the inspectors observed cracks in the concrete dividing wall separating spray pond trains A and B of the Unit 2 spray pond, which were not previously identified. The licensee wrote condition report 24-11703, to evaluate these cracks for structural integrity. The inspectors determined that the licensee had implemented the requirements of Commitment 35.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (10) A1.34 - Nickel Alloy Aging Management Program (Plant Specific) and Commitment 36 This existing, plant-specific program manages cracking caused by PWSCC in all reactor coolant pressure boundary locations that contain Alloy 600, with the objective of maintaining plant safety, minimizing the impact on plant availability, and developing and executing short and long-term strategies for Alloy 600 management.
The licensee has replaced all the Alloy 600 pressurizer and hot leg instrument nozzles in each unit with Alloy 690 nozzle material. All pressurizer heater sleeves have been replaced using Alloy 690 materials and welded using Alloy 52 or 52M filler materials. The licensee has also replaced the Reactor Vessel Heads in all three units with heads having Alloy 690 nozzle material and welded using Alloy 52 filler material.
Commitment 36 specified:
- The existing Nickel Alloy Aging Management Program is credited for license renewal.
The inspectors reviewed the UFSAR supplement, aging management program basis document, interviewed the program owner, implementing procedure 4INT-INCO-06, Alloy 600 Management Program Plan, revision 1, work orders, and completed inspection results. The inspectors also reviewed the RCTSAI for the Nickel Alloy Aging Management Program to verify actions required for the commitment activities.
The inspectors verified the licensee implemented the requirements for Commitment
36.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
- (11) A1.36 - Metal Enclosed Bus & Commitment 38 This new program manages the effects of loose connections, embrittlement, cracking, melting, swelling, chipping or discoloration of insulation, loss of material of bus enclosure assemblies, hardening of boots, gaskets, and sealants, and cracking of internal bus supports to ensure that metal-enclosed buses within the scope of license renewal can perform their intended function. The metal enclosed buses within the scope of this program are those used during station blackout recovery.
Commitment 38 specified:
- The metal enclosed bus program is a new program and will be completed before the period of extended operation and once every 10 years thereafter.
Industry and plant-specific operating experience will be evaluated in the development and implementation of this program.
The inspectors reviewed the UFSAR supplement, work documents, the aging management program basis document, as well as revised procedure 82DP-0EE01, Electrical Aging Management, revision 7, which contained procedural steps on how to inspect metal enclosed bus sections to satisfy License Renewal Commitment 38.
The inspectors reviewed the RCTSAI for the program to verify completed actions for the commitment activities.
The inspectors reviewed condition reports, work orders and procedures.
The inspectors reviewed the inspection and test results from the preventive maintenance of the non-seg-bus ducts sections, under work order 5502737.
The inspectors held discussion with the system owner to review the work performed and recorded data during the cleaning and inspection of the bus ducts within the scope using work instruction 244102. The inspectors observed that the work instructions need to be enhanced for clarity and effectiveness. The licensee wrote condition reports 24-11613 and 24-11614 to review the observed condition for correction.
The inspectors noted that some sections of the bus duct within the scope of license renewal are yet to be inspected and tested. Based on our review of the actions implemented for the metal enclosed bus system program and the subsequent tests yet to be performed and verified, the inspectors concluded that Commitment 38 remains open.
- (12) A2.1 - Metal Fatigue of Reactor Coolant Pressure Boundary and Commitment 39 This program manages transient cycles significant to fatigue damage in metal of the reactor coolant pressure boundary. The program monitors and tracks the number of critical thermal and pressure transients for the selected component and ensures the fatigue usage remaining within the allowable limit, thus minimizing fatigue cracking of metal components caused by anticipated cyclic strains in the material.
Commitment 39 specified:
No later than two years prior to the period of extended operation, the following enhancements will be implemented:
- Cumulative usage factor tracking will be implemented for NUREG/CR-6260, Application of NUREG/CR-5999 Interim Fatigue Curves to Selected Nuclear Power Plant Components, locations not monitored by cycle counting (the reactor vessel shell and lower head (juncture) location will be monitored by cycle counting). For locations identified in NUREG/CR-6260 and monitored by cumulative usage factor (CUF), fatigue usage factor action limits will be required for including effects of the reactor coolant environment.
- Include a computerized program to track and manage both cycle counting and fatigue usage factor. FatiguePro will be used for cycle counting and cycle-based fatigue (CBF) monitoring methods. FatiguePro is an EPRI licensed product.
- Monitor plant transients as required by PVNGS Technical Specification 5.5.5.
Calculate the CUFs for a subset of ASME III Class 1 reactor coolant pressure boundary vessel and piping locations, and component locations with Class 1 analyses. The following methods will be used:
- (1) Use CBF and stress-based fatigue (SBF) CUF calculations to monitor fatigue.
- (2) The SBF method will use a fatigue monitoring software program that incorporates a three-dimensional, six-component stress tensor method meeting ASME III NB-3200 requirements.
- Provide action limits on cycles and on CUF that will initiate corrective actions before the licensing basis limits on fatigue effects at any location are exceeded.
o To ensure sufficient cycle count margin to accommodate occurrence of a low-probability transient, corrective actions must be taken before the remaining number of assumed occurrences for any specified transient becomes less than 1.
o CUF action limits will be established to require corrective action when the calculated CUF (from cycle-based or stress-based monitoring) for any monitored location is projected to reach 1.0 within the next 2 or 3 operating cycles. To ensure sufficient margin to accommodate occurrence of a low probability transient, corrective actions will be taken while there is still sufficient margin to accommodate at least one occurrence of the worst-case design transient event (i.e., with the highest fatigue usage per event cycle).
The inspectors reviewed the UFSAR supplement, interviewed the program owner and reviewed the implementing procedure 73ST-9RCO2, Reactor Coolant System Transient And Operational Cycles, for the enhancement implementation. The inspectors reviewed work orders that implemented the surveillance test associated with the monitoring program and sampled condition reports. The inspectors reviewed the RCTSAI for the program to verify actions on the commitment activities. The inspectors determined the licensee implemented the activities associated with Commitment 39.
Based on the review of the procedures, records and discussion with licensee personnel, the inspectors did not identify any findings or violations of more than minor significance for this aging management program.
INSPECTION RESULTS
No findings were identified.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On October 30, 2024, the inspectors presented the NRC inspection results to Senior Vice Presidents Cary Harbor and Todd Horton and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71003
Corrective Action
Documents
CR
09-00293, 19-10681, 19-10681, 24-04304, 20-00379, 24-
00913, 24-03715, 24-04152, 24-04157, 24-04418, 24-06281,
23-09638, 23-09998, 19-02221, 21-02586, 24-01907, 24-
01908, 24-04304, 24-08541, 24-04304, 24-08541, 22-00376,
19-05786, 21-08448, 21-08395, 21-08397
71003
Corrective Action
Documents
Resulting from
Inspection
CR
24-11612, 24-11613, 24-11614, 24-11778, 24-11703
71003
Drawings
13-A-ZJD-0501,
Sh. 1 of 2
Control BLDG & Corridor BLD
- G. Floor Plan at El 74'-0", 102'-
0" and 120'-0"
71003
Drawings
13-A-ZJD-0501,
Sh. 2 of 2
Control BLDG & Corridor BLD
- G. Floor Plan at El 74'-0", 102'-
0" and 120'-0"
71003
Drawings
13-A-ZJD-0509
Control Building Concrete Block Plans at El. 74ft, 100ft and
Wall Elev.
71003
Drawings
13-A-ZJD-0510
Control Building Concrete Block Wall Elevation and Sections
71003
Drawings
13-A-ZJD-0511
Control Building Concrete Block Plans, Sections and Details
71003
Drawings
13-E-ZYP-015,
Sht. 1 of 6
Non-Segregated Bus Arrangement Plan & Sections Unit 1
Train A
71003
Drawings
13-E-ZYP-015,
Sht. 2 of 6
Non-Segregated Bus Arrangement Plan & Sections Unit 1
Train B
71003
Drawings
13-E-ZYP-015,
Sht. 3 of 6
Non-Segregated Bus Arrangement Plan & Sections Unit 2
Train A
71003
Drawings
13-E-ZYP-015,
Sht. 4 of 6
Non-Segregated Bus Arrangement Plan & Sections Unit 2
Train B
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71003
Drawings
13-E-ZYP-015,
Sht. 5 of 6
Non-Segregated Bus Arrangement Plan & Sections Unit 3
Train A
71003
Drawings
13-E-ZYP-015,
Sht. 6 of 6
Non-Segregated Bus Arrangement Plan & Sections Unit 3
Train B
71003
Drawings
LR-PVNGS-FP-
01-M-FPP-002
P&I Diagram Fire Protection System
71003
Drawings
LR-PVNGS-FP-
01-M-FPP-003
P&I Diagram Fire Protection System
71003
Drawings
LR-PVNGS-FP-
01-M-FPP-004
P&I Diagram Fire Protection System (CO2 System)
71003
Drawings
LR-PVNGS-FP-
01-M-FPP-006
P&I Diagram Fire Protection System
71003
Drawings
LR-PVNGS-FP-
A0-M-FPP-001
P&I Diagram Fire Protection System
71003
Drawings
LR-PVNGS-FP-
A0-M-FPP-005
P&I Diagram Fire Protection System
71003
Drawings
Sheet 2E-25
UNIT 2 East Wall Repairs Section E24 - E25
71003
Drawings
Sheet 2E-26
UNIT 2 East Wall Repairs Section E25 - E26
71003
Drawings
Sheet 2N-1
UNIT 2 North Wall Repairs Section N0 - N1
71003
Drawings
Sketch
Simplified Sketch of 5KV Calvert Bus WO 5459176
71003
Miscellaneous
Simplified Sketch of 5KV Calvert Bus WO 5459176
71003
Miscellaneous
Presentation for Com 36 Nickel Alloy Presentation
10/23/24
71003
Miscellaneous
PVGS Challenge Board Presentation - Masonry Walls,
Commitment 33
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71003
Miscellaneous
13-LS-A133
Regulatory Guide 1.127, Inspection of Water-Control
Structures Associated with Nuclear Power Plants - B2.1.33,
NUREG 1801 Program XI.S7
71003
Miscellaneous
13-LS-A134
PVGS Aging Management Program Evaluation Report -
Nickel Alloy Aging Management - B2.1.34, PSNI
71003
Miscellaneous
13-LS-A201
Metal Fatigue of Reactor Coolant Pressure Boundary - B3.1,
NUREG 1801 Program X.M1
71003
Miscellaneous
20-00379-001
EVAL
U2 Auxiliary Building El. 140 Masonry Wall Cracks. Zone
2ZA3C
1/10/2020
71003
Miscellaneous
21-VE-2065
VE bare metal examination of Reactor Head CEDM
penetrations, 2R23 (1 thru 97 including head vent)
71003
Miscellaneous
23-09638-001 -
EVAL
Metal Enclosed Bus Aging Management Program Challenge
Board Actions.
9/26/2023
71003
Miscellaneous
23-09998-001 -
EVAL
Fire Water Aging Management Challenge Board
10/04/2023
71003
Miscellaneous
23-10537-003
UNIT 1 NRC IP-71003 PHASE I Inspection Self-Assessment
of License Renewal Commitment Completion and Readiness
For NRC Closure
11/19/2023
71003
Miscellaneous
23-10537-003
U1 NRC 71003 Phase 1 Assessment
10/17/2023
71003
Miscellaneous
23-VT-1025
VE bare metal examinations of Reactor Vessel Head
penetrations (2-89, Nozzles 1 thru 97 and 2-88-vent line
10/15/23
71003
Miscellaneous
24-02057-002 -
EVAL
Level 3 Evaluation Report 24-02057-002 for Lots of Cracks
on Spray Pond Wall
3/5/2024
71003
Miscellaneous
24-04418-001 -
EVAL
Unable to Perform Visual Inspection of 3ENBNA06 Calvert
Bus Joint 11E
4/25/2024
71003
Miscellaneous
24-06281-001 -
EVAL
Raised Cover on Vertical Section of Calvert Bus 2ENBNA03
6/7/2024
71003
Miscellaneous
24-VT-3057
VE bare metal examinations of Reactor Vessel Head CEDM
and Reactor Head Vent Penetration Welds
71003
Miscellaneous
2R23 RVCH
2R23 RVCH Image Booklet Vessel Head Pictures
71003
Miscellaneous
2R24 5417717
73ST-9RC02 Surveillance Test Reactor Coolant System
10/16/2023
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Transient and Operating Cycle
71003
Miscellaneous
Units 1, 2 & 3 Alloy 600 Management Program Plan
71003
Miscellaneous
AC1 349.3R-96
Evaluation of Existing Nuclear Safety-Related Concrete
Structures Reported by AC1 Committee 349
March 1996
71003
Miscellaneous
CN397-A00003
2019 Re-Survey of ESP Wall Degradation
71003
Miscellaneous
COR 17-2-024
Component Observation Report (COR) for Spray Pond 2 A
Pump House Below Water and Screen
6/2/2020
71003
Miscellaneous
COR 19-2-001
U2 Auxiliary Building EL.140 Masonry Wall Cracks Zone
2ZA3C
1/29/2020
71003
Miscellaneous
COR 19-2-011
2SPB Spray Pond Train B
71003
Miscellaneous
COR 20-2-010
Concrete Structural Steel/Miscellaneous Steel, Fasteners
and Coatings Nuclear Service Spray Ponds and Pump
Houses, Train A and Train B
71003
Miscellaneous
Fire Water
Presentation
Fire Water AMP 13-LS-A113 AMP Evaluation Report By
Hector Datil
10/22/2024
71003
Miscellaneous
IMS 1117
Fire Water 50 year Sprinkler Test (Response on
Commitment #15 Fire water System program)
11/13/2024
71003
Miscellaneous
Letter #102-5937
2-06160-DCM/GAM
4/1/2021
71003
Miscellaneous
MI 244102
Inspect Clean Calvert Bus MEB
71003
Miscellaneous
NRC IN. 87-67
Lessons Learned from Regional Inspections of Licensee
Actions in Response to IE Bulletin 80-11
2/31/1987
71003
Miscellaneous
Picture of crack 1
Picture of crack in the Fire Pump House A train room 1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71003
Miscellaneous
Picture of crack
Picture of crack on the DS Pump Room West wall South
section 2
71003
Miscellaneous
Picture of crack 2
Picture of Crack in the Fire Pump House A train room 2
71003
Miscellaneous
Picture of crack 3
Picture of Crack in the Fire Pump House A train room 3
71003
Miscellaneous
Picture of crack 4
Picture of Crack in the Fire Pump House A train room 4
71003
Miscellaneous
Picture of crack 5
Picture of Crack in the Fire Pump House A train room 5
71003
Miscellaneous
Picture of crack 6
Picture of Crack in the Fire Pump House A train room 6
71003
Miscellaneous
Picture of crack 7
Picture of Crack on the DS Pump Room West wall North
Section 1
71003
Miscellaneous
Picture of crack 8
Picture of crack on the DS Pump Room West wall North
Section 2
71003
Miscellaneous
Picture of crack 9
Picture of crack on DS Pump Room West wall South section
71003
Miscellaneous
RCTSAI 3246902
Regulatory Commitment Tracking System Action Item for
Fire water System Program
11/06/2008
71003
Miscellaneous
RCTSAI 3246926
-
Regulatory Commitment Tracking System Action Item for
Masonry Wall Program
11/06/2008
71003
Miscellaneous
RCTSAI 3246928
-
Regulatory Commitment Tracking System Action Item for RG 1.127, Inspection of Water-Control Structures Associated
with Nuclear Power Plants program
11/06/2008
71003
Miscellaneous
RCTSAI 3246934
Regulatory Commitment Tracking System Action Item for
Metal Fatigue of Reactor Coolant Boundary Program
11/06/2008
71003
Miscellaneous
RCTSAI 3247220
Regulatory Commitment Tracking System Action Item for
Metal Enclosed Bus
11/07/2008
71003
Miscellaneous
RCTSAI 3260208
Regulatory Commitment Tracking System Action Item for
Nickel Alloy Aging Management Program is credited for
2/10/2008
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71003
Miscellaneous
Criteria and Design Features for Inspection of Water Control
Structures Associated with Nuclear Power Plants.
2/2/16
71003
Miscellaneous
STWO 5265352
Surveillance Test WO#52653520 for Miscellaneous UNIT 2
4/12/2022
71003
Miscellaneous
CR ENG EVAL 20-00379-002 (WO 5208388)
1/10/2020
71003
Miscellaneous
CR ENG EVAL 20-00379-002 (WO 5208388)
71003
Miscellaneous
WSL 3440775/0
Non-seg. Phase Bus Duct AT4.16KV Swgr E-PBA-S03 work
instructions
71003
Procedures
81DP-
0ZZ01
Civil System, Structure, and Component Monitoring Program
(Basis)
28a
71003
Procedures
81DP-
0ZZ01
Civil System, Structure, and Component Monitoring Program
71003
Procedures
\\14DP-0FP44
Fire Protection Test Program Requirement
71003
Procedures
Condition Reporting Process (Basis)
48a
71003
Procedures
Condition Reporting Process
71003
Procedures
13-CN-0397
section 6.3
Essential Spray Pond Structure Concrete Repair
71003
Procedures
Fire Protection Test Program Requirement (Basis)
8a
71003
Procedures
Appendix R/FTS Fire Barrier Surveillance (For Walls,
Floors/Ceilings and Raceways (Basis)
15a
71003
Procedures
Appendix R/FTS Fire Barrier Surveillance (For Walls,
Floors/Ceilings and Raceways
71003
Procedures
Appendix A Fire Barrier Surveillance (Basis)
15a
71003
Procedures
Appendix A Fire Barrier Surveillance
71003
Procedures
Alloy 600 Management Program Plan
71003
Procedures
Alloy 600 Management Program Plan
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71003
Procedures
Maintenance Rule
71003
Procedures
Maintenance Rule Monitoring Process
71003
Procedures
Maintenance Rule Monitoring Process (Basis)
1a
71003
Procedures
Maintenance Rule Monitoring Process (Basis)
1a
71003
Procedures
Reactor Coolant System Transient and Operational Cycles
(Basis)
17a
71003
Procedures
Reactor Coolant System Transient and Operational Cycles
71003
Procedures
Visual Examination For Leakage
71003
Procedures
Bare Metal Visual Examination of Reactor Vessel Upper
Head
71003
Procedures
Design Change Process (Basis)
56a
71003
Procedures
Design Change Process
71003
Procedures
Electrical Aging Management
71003
Procedures
Electrical Aging Management (Basis)
7a
71003
Work Orders
WO 5502737, 5548477, 5517543, 5548330, 5402090, 5087628,
208388, 5548360