IR 05000269/2022010
| ML22179A298 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/30/2022 |
| From: | Gerald Mccoy Division of Reactor Safety II |
| To: | Snider S Duke Energy Carolinas |
| References | |
| IR 2022010 | |
| Download: ML22179A298 (17) | |
Text
SUBJECT:
OCONEE NUCLEAR STATION - TRIENNIAL FIRE PROTECTION INSPECTION REPORT 05000269/2022010 AND 05000270/2022010 AND 05000287/2022010
Dear Mr. Snider:
On May 19, 2022, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Oconee Nuclear Station. On June 30, 2022, 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.
Two findings of very low safety significance (Green) are documented in this report. Two of these findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violations or the significance or severity of the violations documented in this inspection report, you should provide a response within 30 days of the date of this inspection report, with the basis for your denial, to the U.S. Nuclear Regulatory Commission, ATTN:
Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; the Director, Office of Enforcement; and the NRC Resident Inspector at Oconee Nuclear 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, Gerald J. McCoy, Chief Engineering Br 2 Division of Reactor Safety June 30, 2022 Signed by McCoy, Gerald on 06/30/22 Docket Nos. 05000269 and 05000270 and 05000287 License Nos. DPR-38 and DPR-47 and DPR-55
Enclosure:
As stated
Inspection Report
Docket Numbers:
05000269, 05000270 and 05000287
License Numbers:
Report Numbers:
05000269/2022010, 05000270/2022010 and 05000287/2022010
Enterprise Identifier:
I-2022-010-0041
Licensee:
Duke Energy Carolinas, LLC
Facility:
Oconee Nuclear Station
Location:
Seneca, SC
Inspection Dates:
March 21, 2022 to April 08, 2022
Inspectors:
P. Braaten, Senior Reactor Inspector
L. Jones, Senior Reactor Inspector
J. Montgomery, Senior Reactor Inspector
D. Terry-Ward, Construction Inspector
Approved By:
Gerald J. McCoy, Chief
Engineering Br 2
Division of Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a triennial fire protection inspection at Oconee Nuclear 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
Inadequate Design Change Results In Inadequate Defense-In-Depth Recovery Action Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000269,05000270,05000287/202201 0-01 Open/Closed None (NPP)71111.21N.
The inspectors identified a Green finding and associated non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2 and 3 Renewed Facility Operating License Condition 3.D for the licensees failure to verify the adequacy of the design change of the Nuclear Safety Capability Assessment (NSCA). Specifically, fire protection engineers failed to ensure that a defense-in-depth (DID) recovery action added to the NSCA would stop a high-pressure injection (HPI)pump that had spuriously started.
Failure to Develop Accurate Safe Shutdown Equipment List Leads to Inaccurate Probabilistic Risk Assessment Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000269,05000270,05000287/202201 0-02 Open/Closed None (NPP)71111.21N.
The inspectors identified a Green finding and associated NCV for the licensees failure to develop a comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event. Specifically, the sites safe shutdown equipment list (SSEL)contained in the NSCA, which informed the sites Fire probabilistic risk assessment (PRA)equipment selection calculation did not account for the fact that the HPI pumps were required to be off to reach and maintain hot standby.
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.
REACTOR SAFETY
71111.21N.05 - Fire Protection Team Inspection (FPTI) Structures, Systems, and Components (SSCs) Credited for Fire Prevention, Detection, Suppression, or Post-Fire Safe Shutdown Review (IP Section 03.01)
The inspectors verified that components and/or systems will function as required to support the credited functions stated for each sample.
- (1) Unit 3 Cable Room Suppression System (FA AB/FZ 101)
- (3) Unit 2 High-Pressure Injection System
- (4) Protected Service Water System
Fire Protection Program Administrative Controls (IP Section 03.02) (1 Sample)
The inspectors verified that the selected control or process is implemented in accordance with the licensees current licensing basis. If applicable, ensure that the licensees FPP contains adequate procedures to implement the selected administrative control. Verify that the selected administrative control meets the requirements of all committed industry standards.
- (1) NFPA 805 Monitoring Program
Fire Protection Program Changes/Modifications (IP Section 03.03) (2 Samples)
The inspectors verified the following:
a. Changes to the approved FPP do not constitute an adverse effect on the ability to safely shutdown.
b. The adequacy of the design modification, if applicable.
c. Assumptions and performance capability stated in the SSA have not been degraded through changes or modifications.
d. The FPP documents, such as the Updated Final Safety Analysis Report, fire protection report, FHA, and SSA were updated consistent with the FPP or design change.
e. Post-fire SSD operating procedures, such as abnormal operating procedures, affected by the modification were updated.
- (1) ONS-2021-011 - Plant Impact Review associated with EC 414403, Install additional letdown isolation system for X-HP-5
- (2) ONS-2021-015 - Plant Impact Review associated with multiple ECs for a Unit 1 SSC Control Console Mod
INSPECTION RESULTS
Inadequate Design Change Results In Inadequate Defense-In-Depth Recovery Action Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000269,05000270,05000287/20220 10-01 Open/Closed None (NPP)71111.21N.0 The inspectors identified a Green finding and associated non-cited violation (NCV) of Oconee Nuclear Station Units 1, 2 and 3 Renewed Facility Operating License Condition 3.D for the licensees failure to verify the adequacy of the design change of the Nuclear Safety Capability Assessment (NSCA). Specifically, fire protection engineers failed to ensure that a defense-in-depth (DID) recovery action added to the NSCA would stop a high-pressure injection (HPI)pump that had spuriously started.
Description:
While reviewing the licensees safe shutdown analysis, OSC-9669, Oconee Nuclear Safety Capability Assessment for Units 1, 2, and 3, the team identified that the required state of the HPI pumps is 'off' when shutting down utilizing the safe shutdown facility (SSF). The analysis also identified that for fires in the Auxiliary Building, there are potential fire impacts to cables associated with the high-pressure injection pumps.
These fire impacts were identified as variances from deterministic requirements (VFDR) and were required to be addressed in order to meet the sites performance criteria. VFDR-AB-016 identified that the 2A HPI pump may be subject to a fire-induced spurious start. The site dispositioned the variance by stating that the site met the risk requirements and credited a recovery action in order to maintain defense-in-depth.
The credited recovery action was added to the safe shutdown analysis via EC 403491 in December 2016. The purpose of this EC was to update applicable design and licensing basis documents to complete documentation alignment to NFPA 805. The DID recovery action credited was to remove the 7kV/4kV control power fuses. This action was implemented in procedure AP/2/A/1700/050, Challenging Plant Fires. This procedure and action directed operators to remove only the control power fuses for all three HPI pumps.
By removing control power fuses, the pump would lose its capability to be controlled remotely, but the current state of the pump would be unchanged. This action, therefore, prevented any subsequent fire-induced spurious operation, but the action would not alter the state and mitigate a fire-induced spurious start that had already occurred. Therefore, the team determined that this action was not effective in mitigating the consequences of a fire induced spurious operation.
Corrective Actions: NCR 02421758
Performance Assessment:
Performance Deficiency: The failure to verify the adequacy of the design change of the NSCA was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify the adequacy of the design of NSCA reduced the defense-in-depth aspect of the fire protection program as described in Section 1.2 of NFPA 805.
Significance: The inspectors assessed the significance of the finding using Appendix F, Fire Protection and Post - Fire Safe Shutdown SDP. Using IMC 0609, Appendix F, Attachment 1, the inspectors determined the issue was of very low safety significance (Green) because the finding did not adversely affect the ability to reach and maintain hot shutdown/hot standby conditions using the credited safe shutdown success path (Question 1.4.7-C)
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Oconee Nuclear Station Units 1, 2 and 3 Renewed Facility Operating License Condition 3.D required the licensee to implement and maintain in effect all provisions of the approved FPP that complied with 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, as specified in the revised licensee's amendment request dated April 14, 2010. Section 4.7.3 of the revised license amendment request stated that the site maintained the Fire Protection Quality Assurance program that was in place prior to transition to NFPA 805. This QA Program is documented in Fleet Fire Protection Program Manual PD-FP-ALL-1500. Section 5.6.3 of PD-AP-ALL-1500 stated that design activities shall be accomplished in accordance with procedures that ensure the applicable design requirements are included and that appropriate reviews are conducted.
Contrary to the above, since December 7, 2016, the design activity was not accomplished in accordance with procedures that ensured the applicable design requirements were included and that appropriate reviews are conducted. Specifically, the design review by the fire protection engineers failed to ensure that the DID recovery action added to the NSCA would mitigate an HPI pump that was running.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Develop Accurate Safe Shutdown Equipment List Leads to Inaccurate Probabilistic Risk Assessment Cornerstone Significance Cross-Cutting Aspect Report Section Mitigating Systems Green NCV 05000269,05000270,05000287/20220 10-02 Open/Closed None (NPP)71111.21N.0
The inspectors identified a Green finding and associated NCV for the licensees failure to develop a comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event. Specifically, the sites safe shutdown equipment list (SSEL)contained in the NSCA, which informed the sites Fire probabilistic risk assessment (PRA)equipment selection calculation did not account for the fact that the HPI pumps were required to be off to reach and maintain hot standby.
Description:
The licensees NSCA for a postulated fire in the Auxiliary Building (to include the main control room) credits the use of the SSF as a dedicated shutdown strategy. The SSF utilizes the Reactor Coolant Makeup (RCMU) Pump to provide reactor coolant system (RCS)seal cooling to achieve and maintain Mode 3. Because of the low flowrate of the RCMU pump, the success of the SSF requires the HPI pumps to be off, to not go solid in the RCS.
For NFPA 805, the licensees credited safe and stable state is Mode 4. This is because the SSF cannot maintain the plant at Mode 3 indefinitely. Therefore, after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the licensee credits re-establishing control via the main control room to bring the plant to Mode 4. This involves starting a HPI pump. For a fire in the AB, the NSCA identifies VFDR-AB-016 which states, in part, "fire damage to cables may prevent tripping the HPI pump or result in spurious pump start. This is notable because the normal operational alignment of the plant would have a HPI pump running.
The methodology for the inclusion of equipment and basic events in the fire PRA is documented in OSC-8978, Rev. 6, "ONS Fire PRA - Equipment Selection Report". This report reflects that equipment from the post fire SSEL from Revision 4 of the NSCA. The SSEL from Rev. 4 of the NSCA does not reflect the need for the HPI pumps to be 'off' in order to reach and maintain Mode 3it only reflects the need for the HPI pumps to be 'on' in order to reach and maintain Mode 4. As a result, the PRA equipment selection calculation does not contain basic events for all applicable failure modes for each HPI.
Corrective Actions: NCR 02423868, NCR 02422326
Performance Assessment:
Performance Deficiency: The licensees failure to develop a comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event was a performance deficiency.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to develop a comprehensive list of systems and equipment led to inaccurate plant fire risk assessments because the effect of a running HPI pump on the success criteria of reaching and maintaining Mode 3 from the SSF was not considered as a part of the risk analysis.
Significance: The inspectors assessed the significance of the finding using Appendix F, Fire Protection and Post - Fire Safe Shutdown SDP. Using IMC 0609, Appendix F, Attachment 1, the inspectors determined the issue was associated with the post-fire safe shutdown finding category, and could not screen the issue using the 3 questions in Step 1.4.7. Using Step 1.5, inspectors determined that the plant had a fire PRA capable of adequately evaluating the risk associated with the finding, the licensees risk-based evaluation for this fire finding indicate a delta-CDF of less than 1E-6, and the evaluation result was accepted by a regional Senior Reactor Analyst (SRA).
The SRA independently assessed the adequacy and results of the licensees fire PRA using the guidance in Appendix K, Maintenance Risk Assessment and Risk Management SDP.
Using IMC 0609, Appendix K, the SRA calculated the incremental core damage probability deficit (ICDPD) and the incremental large early release probability (ILERPD) deficit due to the performance deficiency and entered this value into figure 1 of IMC 0609 Appendix K, treating the issue as an inadequate risk assessment. The total calculated risk deficit was 3.24E-8. As a result, both Appendix F questions 1.5.1 A and B were answered YES, which screened the finding to Green.
Cross-Cutting Aspect: Not Present Performance. No cross-cutting aspect was assigned to this finding because the inspectors determined the finding did not reflect present licensee performance.
Enforcement:
Violation: Oconee Nuclear Station Units 1, 2 and 3 Renewed Facility Operating License Condition 3.D required the licensee to implement and maintain in effect all provisions of the approved FPP that complied with 10 CFR 50.48(c), National Fire Protection Association Standard NFPA 805, as specified in the NRC safety evaluation report (SER) dated December 29, 2010. NFPA 805 Section 2.4.2.1 stated, in part, that a comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event shall be developed.
Contrary to the above, since December 29, 2010, the licensee failed to develop a comprehensive list of systems and equipment and their interrelationships to be analyzed for a fire event. Specifically, the sites SSEL contained in the NSCA, which informed the sites Fire PRA equipment selection calculation did not account for the fact that the HPI pumps were required to be off to reach and maintain hot standby.
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
On June 30, 2022, the inspectors presented the triennial fire protection inspection results to Steven M. Snider and other members of the licensee staff.
On April 7, 2022, the inspectors presented the initial FPTI inspection results to Steven Snider and other members of the licensee staff.
On May 19, 2022, the inspectors presented the updated FPTI inspection results to Steven Snider and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2-9252765-002
Calculation OSC-10319, Rev. 15, pages X-455 of 540
06/23/2016
OSC-10055
PSW 125 VDC Relay setting, protective device
coordination and hydrogen generation
Rev. 001
OSC-10319
Oconee Units 1, 2, 3 NFPA 805 Coordination Study
Rev. 44
OSC-10952
(AREVA 32-
9192968-001)
U2, EC91857 NFPA 805 Breaker and Fuse Coordination
Study
Rev. 1
OSC-10974
ONS RIS 2005-07 Alternate Compensatory Measure
Calculation
Rev. 6
OSC-11082
Protective Device Settings for MCC MXAWC2, MXAWC3,
MXAWC4 and MXAWC2
Rev. 5
OSC-11549
ONS Fire PRA FRE Input Calculation
Rev. 3
OSC-11914
ONS Fire PRA - Circuit Failure Analysis
Rev. 0
OSC-8978
ONS Fire PRA - Equipment Selection Report
Rev. 6
OSC-9314
NFPA 805 Transition Risk-Informed Performance-Based
Fire Risk Evaluation
Rev. 6
OSC-9375
ONS Fire PRA - Fire Scenario Report
Rev. 9
OSC-9376
ONS Fire PRA - Cable Selection Report
Rev. 5
OSC-9377
ONS Fire PRA - Model Development Report
Rev. 5
OSC-9659
Oconee Nuclear Safety Capability Assessment for Units
1, 2, and 3
Rev. 11
OSC-9831
Protective Relay Settings Associated with PSW
Switchgear
Rev. 11
Calculations
OSC-9887
NFPA 805 Monitoring Program Scoping
Rev. 5
An evaluation is needed to determine if a breaker
coordination study is required for PSW Milestone 3
04/23/2014
Corrective Action
Documents
AREVA Calculation Error for AWC EC 115193
06/22/2016
2420818
22 FPTI: Fire Detector Appears to be Painted
2421758
Generate updated PCS Action list
2422029
22 FPTI: EC Record Illegible in Fusion
71111.21N.05
Corrective Action
Documents
Resulting from
Inspection
2422261
22 FPTI - WOT instructions do not match design
documents
03/31/2022
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2422326
22 FPTI: Concern with Fire PRA modeling of HPI Pump
2422877
22 FPTI: Inconsistent labeling of breakers in Calcs &
22 FPTI - OSC-10319 Revision Methodology
2423114
22 FPTI - ArcPlus Revision Methodology
2423868
22 FPTI - Update Fire PRA Equipment Selection
Calculation
O -1705-A
One Line Diagram, 240/120VAC Station Aux. Circuits
Comp., ICS & REG Supply
Rev. 86
O -1711-B
Connection Diagram, Unit Control Board 2UB1
Rev. 87
O -1711-D
Connection Diagram, Unit Control Board 2UB1
Rev. 57
O -1721-A
Connection Diagram Valves, High Pressure Injection &
Purification System
Rev. 42
O -1721-A-001
Connection Diagram, High Pressure Injection &
Purification System
Rev. 3
O -1721-A-002
Connection Diagram, High Pressure Injection Termination
Cab., 2HPICA0090
Rev. 0
O -1757-C
Connection Diagram, Even Channels, Engineered
Safeguards Terminals, Cabinet 2ESTC2
Rev. 45
O -1757-H-001
Connection Diagram, Engineered Safeguards, Cabinet
2PPSCA0018
Rev. 0
O -1909
Auxiliary Building, Electrical Equipment Layout Plan
sections. & Details below EL. 809 + 3-Cols. 76 to 85
Rev. 43
O -1913
Auxiliary Building, Electrical Equipment Layout,
Penetration Room, Plan Below El. 838 + 0
Rev. 43
O -305 A
General Arrangement, Auxiliary Building - Units 1 & 2,
Plan at EL. 809 + 3
Rev. 40A
O-0906
Auxiliary Building, Electrical Equipment Layout Plan,
below EL. 771 + 0
Rev. 44
O-0908-G
Auxiliary Building, Electrical Equipment Layout, Plan
below E
- L. 796 + 6, Columns 70-76, PSW Cable tray
Rev. 9
O-1422-X-029
Instrument Detail, Letdown Isolation Valve Control, 2HP-5
Rev. 16
O-1422-X-47
Instrument Detail, RCP Seal Return, Isolation 2HP-21
Rev. 7
Drawings
O-1703-D
One Line Diagram, Station Auxiliary Circuits 600V/208V
Rev. 65
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
L/C 2X5 & MCC 2XI, 2XL & 2XD
O-1703-E
One Line Diagram, Station Auxiliary Circuits, 600V/208V/
L/C 2X6 & MCC 2XJ, 2XN & 2XP
Rev. 59
O-1705
Oneline Diagram, 120VAC & 125VDC Station Aux.
Circuits Instrumentation Vital Buses
Rev. 89
O-1721
Connection Diagram, Miscellaneous Equipment, High
Pressure Injection & Purification System
Rev. 27
O-1752-A-29
Interconnection Diagram, Motor Control Center No. 2XL,
Units No. 1 thru 5
Rev. 18
O-1752-A-36
Interconnection Diagram, Motor Control Center No. 2XN,
Units No. 1 thru 5
Rev. 24
O-1752-A-67
Interconnection Diagram, 600V SSF Control Center No.
2XSF Units F05A thru F06D
Rev. 4
O-1752-A-68
Interconnection Diagram, 208V SSF Motor Control Center
No. 2XSFA Units F01A thru F02A
Rev. 6
O-1766-C
Connection Diagram, Misc. Term Cab. 2MTC2
Rev. 38
O-1767-A-060
Connection Diagram, Reactor Building Penetrations, Type
D6 Penetrations, Penetration No. WA3
Rev. 14
O-1767-A-062
Connection Diagram, Reactor Building Penetrations, Type
M Penetrations, Penetration No. WA11
Rev. 3
O-1790-P
Connection Diagram, Computer Cabinet 2G7, Right Side
wall
Rev. 34
O-1909-A-001
Auxiliary Building, Electrical Equipment Layout Plan
below EL. 809 + 3
Rev. 2
O-6701
One Line Diagram, Station Auxiliary Circuits, 600V PSW
MCC 1XPSW
Rev. 6
O-6702
Auxiliary Building, One Line Diagram, Station Auxiliary
Circuits 600V PSW MCC 2XPSWA & 2XPSWB
Rev. 8
O-6703
Auxiliary Building, One Line Diagram, Station Auxiliary
Rev. 5
O-6721-A
Connection Diagram, Valves PSW & High Injection
System
Rev. 0
O-6721-B
Connection Diagram, Valves PSW & High Injection
System
Rev. 1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
O-6721-C
Connection Diagram, Valves PSW & High Injection
System
Rev. 1
O-6752-A
Interconnection Diagram, Motor Control Center No.
1XPSW Units No. 1, 2, 3, 4 & 5
Rev. 4
O-6752-B
Interconnection Diagram, Motor Control Center No.
2XPSWA & 2XPSWB, Units 1F, 2F, 3F, 4F & 5F
Rev. 6
O-6752-C
Interconnection Diagram, Motor Control Center 3XPSWA,
Units F1, F2 and F3
Rev. 6
O-6799-B
Connection Diagram, Terminal Cab. 2PSWCA0001,
Protected Service Water
Rev. 6
O-907
Auxiliary Building, Electrical Equipment Layout Plan,
below EL. 783 + 9-Column line 70 to 76
Rev. 52
OEE-242-1
Elementary Diagram, Component Cooling, VLV., 2CC-8,
2CC VA0008
Rev. 9
OEE-251
Elementary Diagram, Letdown Cooler A outlet valve,
SSF 2-HP-3 (FS/2/51/3) 2HP VA0003
Rev. 7
OEE-251-03
Elementary Diagram, Letdown ISO
- L. Valve 2HP-5, 2 HP
VA0005
Rev. 16
OEE-251-0A
Elementary Diagram, Letdown Cooler A Outlet Valve,
SSF 2-HP-3 (FS/2/51/3) 2HP VA0003
Rev. 5
OEE-251-10
Elementary Diagram, HP Injection Pump, Discha.
Crossov. Valve FS/2/51/55
Rev. 3
OEE-251-2
Elementary Diagram,
2-HP-20 (FS/2/51/39) 2HP VA0020
Rev. 9
OEE-251-8
Elementary Diagram, RC Pump Seal Return Isol. Vlv. 2-
HP-21, 2HP VA0021
Rev. 11
OEE-251-9
Elementary Diagram, Feed and Bleed Controls, Valves
2/51/21 (2HP-V10) (2HP-14) and 2/51/14 (2HP-
V18)(2HP-16)
Rev. 13
OEE-265-01
Elementary Diagram, Protected Service Water, STM GEN
3A and 3B Flow Isolation MOV 3PSW-6
Rev. 2
OEE-265-02
Elementary Diagram, Protected Service Water, Steam
Generator 2A Flow, Control MOV 2PSW-22
Rev. 3
OEE-265-03
Elementary Diagram, Protected Service Water, Steam
Rev. 3
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Generator 2A Throttle MOV 2PSW-24
OEE-265-04
Elementary Diagram, Protected Service Water, Steam
Generator 2B Flow Control, MOV 2PSW-24
Rev. 3
OEE-265-05
Elementary Diagram, Protected Service Water, Steam
Generator 2B Throttle MOV 2PSW-25
Rev. 3
OEE-265-06
Elementary Diagram, Protected Service Water, HPI
Isolation Valve 2HP-139
Rev. 2
OEE-265-07
Elementary Diagram, Protected Service Water, HPI Flow
Throttle Valve 2HP-140
OEE-265-13
Elementary Diagram, PSW System/ S/G 2A Supply
Control Loop, Valve 2PSW-22
Rev. 2
OFD-101A-2.1
Flow Diagram of High Pressure Injection System Letdown
Section
Rev. 51
OFD-101A-2.2
Flow Diagram of High Pressure Injection System Storage
Section
Rev. 46
OFD-101A-2.3
Flow Diagram of High Pressure Injection System
Charging Section
Rev. 34
OFD-101A-2.4
Flow Diagram of High Pressure Injection System
Charging Section
Rev. 48
OFD-101A-2.5
Flow Diagram of High Pressure Injection System SSF
Portion
Rev. 24
OFD-131A-1.1
Flow Diagram of Protected Service Water (PSW) System
Rev. 3
OFD-131A-2.2
Flow Diagram of Protected Service Water (PSW) System,
Steam Generator and HPI Pump motor Cooling Service
Rev. 2
OD300958/EC91869 Unit 3 HPI Pump System Power -
Pre-Outage
Rev. 11
OD500922 - PSW Power Feed Installation
Rev. 18
Correction - Implement MOD to Add Fuses/Breakers to
loads on MCC AWC3
Rev. 4
Correction - Implement MOD to Add Fuses/Breakers to
loads on MCC AWC3
Rev. 0
Unit 3, Make Field Changes associated with 3RC66 to
improve the DC Voltage
Rev. 0
Engineering
Changes
ONS-2021-011
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
ONS-2021-015
Oconee Nuclear Station Updated Final Safety Analysis
Report
Rev. 28
Safety Evaluation by the Office of Nuclear Reactor
Regulation, Transition to a Risk-Informed
Performance Based Fire Protection Program in
Accordance with 10 CFR 50.48(c)
December
29, 2010
Safety Evaluation by the Office of Nuclear Reactor
Regulation, Oconee Nuclear Station Standby Shutdown
Facility
April 28,
1983
Oconee Nuclear Station License Amendment Request to
Adopt NFPA 805
4/14/2010
Oconee Selected Licensing Commitments, Section 16.9,
Auxiliary Systems
Rev. 1
51-9249010-004
Oconee Nuclear Safe Capability Assessment for Unit 1, 2,
and 3, Appendix B: Safe Shutdown Equipment List
(SSEL)
08/15/2016
Duke-QAPD-001
Quality Assurance Program Description Operating Fleet
Rev. A
OELBKAWC302A
OELBKAWC302A EDB report, ST 1804 partial equipment
key, AWC3-2A
3/30/2022
OELBKAWC302B
OELBKAWC302A EDB report, ST 1804 partial equipment
key, AWC3-2B
3/30/2022
OSC-9659
Oconee Nuclear Safety Capability Assessment for Units
1, 2 and 3
Rev. 011
OSS-0254.00-00-
1001
(Mech) High Pressure Injection and Purification &
Deborating Demineralizer Systems
Rev. 062
OSS-0254.00-00-
1053
Protected Service Water System
Rev. 003
OSS-0254.00-00-
1053
Protected Service Water System
Rev. 3
OSS-0254.00-00-
2017
(Elec) Design Basis Specification for Fire Detection
System
Rev. 019
Miscellaneous
OSS-0254.00-00-
4008
(Mech) Design Basis Specification for Fire Protection
Rev. 47
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
OSS-0254.00-00-
4008
Design Basis Specification for Fire Protection
Rev. 47
OSS-0256.00-00-
1002
Design Basis Specification for High Pressure Service
Water System
Rev. 47
PD-FP-ALL-1500
Fleet Fire Protection Program Manual
Rev. 1
PUL 92-9257585
Product Upgrade list (PUL)
05/19/2016
AD-EL-ALL-1117
Design Analyses and Calculations
Rev. 10
AD-EL-ALL-1132
Preparation and Control of Design Change Engineering
Changes
Rev. 4
AD-OP-ALL-0105
Rev. 6
Procedures
SD 3.2.14
Fire Protection Program Compensatory Measure Process
Rev. 4
Work Orders
20059574
EC115193 Perform Breaker Bench Test for new AWC3
Bkrs
07/12/2016