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
June 30, 2022
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, Signed by McCoy, Gerald on 06/30/22 Gerald J. McCoy, Chief Engineering Br 2 Division of Reactor Safety 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: DPR-38, DPR-47 and DPR-55 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 Enclosure
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 Report Aspect Section Mitigating Green None (NPP) 71111.21N.
Systems NCV 05 05000269,05000270,05000287/202201 0-01 Open/Closed 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 Report Aspect Section Mitigating Green None (NPP) 71111.21N.
Systems NCV 05 05000269,05000270,05000287/202201 0-02 Open/Closed 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 Report Aspect Section Mitigating Green None (NPP) 71111.21N.0 Systems NCV 5 05000269,05000270,05000287/20220 10-01 Open/Closed 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 Report Aspect Section Mitigating Green None (NPP) 71111.21N.0 Systems NCV 5 05000269,05000270,05000287/20220 10-02 Open/Closed 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 Type Designation Description or Title Revision or
Procedure Date
71111.21N.05 Calculations 32-9252765-002 Calculation OSC-10319, Rev. 15, pages X-455 of 540 06/23/2016
OSC-10055 PSW 125 VDC Relay setting, protective device Rev. 001
coordination and hydrogen generation
OSC-10319 Oconee Units 1, 2, 3 NFPA 805 Coordination Study Rev. 44
OSC-10952 U2, EC91857 NFPA 805 Breaker and Fuse Coordination Rev. 1
(AREVA 32- Study
9192968-001)
OSC-10974 ONS RIS 2005-07 Alternate Compensatory Measure Rev. 6
Calculation
OSC-11082 Protective Device Settings for MCC MXAWC2, MXAWC3, Rev. 5
MXAWC4 and MXAWC2
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 Rev. 6
Fire Risk Evaluation
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 Rev. 11
1, 2, and 3
OSC-9831 Protective Relay Settings Associated with PSW Rev. 11
Switchgear
OSC-9887 NFPA 805 Monitoring Program Scoping Rev. 5
Corrective Action AR 01865817 An evaluation is needed to determine if a breaker 04/23/2014
Documents coordination study is required for PSW Milestone 3
AR 02040023 AREVA Calculation Error for AWC EC 115193 06/22/2016
Corrective Action 02420818 2022 FPTI: Fire Detector Appears to be Painted
Documents 02421758 Generate updated PCS Action list
Resulting from 02422029 2022 FPTI: EC Record Illegible in Fusion
Inspection 02422261 2022 FPTI - WOT instructions do not match design 03/31/2022
documents
Inspection Type Designation Description or Title Revision or
Procedure Date
2422326 2022 FPTI: Concern with Fire PRA modeling of HPI Pump
2422877 2022 FPTI: Inconsistent labeling of breakers in Calcs &
ECs 2423084 2022 FPTI - OSC-10319 Revision Methodology
2423114 2022 FPTI - ArcPlus Revision Methodology
2423868 2022 FPTI - Update Fire PRA Equipment Selection
Calculation
Drawings O -1705-A One Line Diagram, 240/120VAC Station Aux. Circuits Rev. 86
Comp., ICS & REG Supply
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 & Rev. 42
Purification System
O -1721-A-001 Connection Diagram, High Pressure Injection & Rev. 3
Purification System
O -1721-A-002 Connection Diagram, High Pressure Injection Termination Rev. 0
Cab., 2HPICA0090
O -1757-C Connection Diagram, Even Channels, Engineered Rev. 45
Safeguards Terminals, Cabinet 2ESTC2
O -1757-H-001 Connection Diagram, Engineered Safeguards, Cabinet Rev. 0
2PPSCA0018
O -1909 Auxiliary Building, Electrical Equipment Layout Plan Rev. 43
sections. & Details below EL. 809 + 3- Cols. 76 to 85
O -1913 Auxiliary Building, Electrical Equipment Layout, Rev. 43
Penetration Room, Plan Below El. 838 + 0
O -305 A General Arrangement, Auxiliary Building - Units 1 & 2, Rev. 40A
Plan at EL. 809 + 3
O-0906 Auxiliary Building, Electrical Equipment Layout Plan, Rev. 44
below EL. 771 + 0
O-0908-G Auxiliary Building, Electrical Equipment Layout, Plan Rev. 9
below E
- L. 796 + 6, Columns 70-76, PSW Cable tray
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
O-1703-D One Line Diagram, Station Auxiliary Circuits 600V/208V Rev. 65
Inspection Type Designation Description or Title Revision or
Procedure Date
L/C 2X5 & MCC 2XI, 2XL & 2XD
O-1703-E One Line Diagram, Station Auxiliary Circuits, 600V/208V/ Rev. 59
L/C 2X6 & MCC 2XJ, 2XN & 2XP
O-1705 Oneline Diagram, 120VAC & 125VDC Station Aux. Rev. 89
Circuits Instrumentation Vital Buses
O-1721 Connection Diagram, Miscellaneous Equipment, High Rev. 27
Pressure Injection & Purification System
O-1752-A-29 Interconnection Diagram, Motor Control Center No. 2XL, Rev. 18
Units No. 1 thru 5
O-1752-A-36 Interconnection Diagram, Motor Control Center No. 2XN, Rev. 24
Units No. 1 thru 5
O-1752-A-67 Interconnection Diagram, 600V SSF Control Center No. Rev. 4
2XSF Units F05A thru F06D
O-1752-A-68 Interconnection Diagram, 208V SSF Motor Control Center Rev. 6
No. 2XSFA Units F01A thru F02A
O-1766-C Connection Diagram, Misc. Term Cab. 2MTC2 Rev. 38
O-1767-A-060 Connection Diagram, Reactor Building Penetrations, Type Rev. 14
D6 Penetrations, Penetration No. WA3
O-1767-A-062 Connection Diagram, Reactor Building Penetrations, Type Rev. 3
M Penetrations, Penetration No. WA11
O-1790-P Connection Diagram, Computer Cabinet 2G7, Right Side Rev. 34
wall
O-1909-A-001 Auxiliary Building, Electrical Equipment Layout Plan Rev. 2
below EL. 809 + 3
O-6701 One Line Diagram, Station Auxiliary Circuits, 600V PSW Rev. 6
MCC 1XPSW
O-6702 Auxiliary Building, One Line Diagram, Station Auxiliary Rev. 8
Circuits 600V PSW MCC 2XPSWA & 2XPSWB
O-6703 Auxiliary Building, One Line Diagram, Station Auxiliary Rev. 5
O-6721-A Connection Diagram, Valves PSW & High Injection Rev. 0
System
O-6721-B Connection Diagram, Valves PSW & High Injection Rev. 1
System
Inspection Type Designation Description or Title Revision or
Procedure Date
O-6721-C Connection Diagram, Valves PSW & High Injection Rev. 1
System
O-6752-A Interconnection Diagram, Motor Control Center No. Rev. 4
1XPSW Units No. 1, 2, 3, 4 & 5
O-6752-B Interconnection Diagram, Motor Control Center No. Rev. 6
2XPSWA & 2XPSWB, Units 1F, 2F, 3F, 4F & 5F
O-6752-C Interconnection Diagram, Motor Control Center 3XPSWA, Rev. 6
Units F1, F2 and F3
O-6799-B Connection Diagram, Terminal Cab. 2PSWCA0001, Rev. 6
Protected Service Water
O-907 Auxiliary Building, Electrical Equipment Layout Plan, Rev. 52
below EL. 783 + 9- Column line 70 to 76
OEE-242-1 Elementary Diagram, Component Cooling, VLV., 2CC-8, Rev. 9
2CC VA0008
OEE-251 Elementary Diagram, Letdown Cooler A outlet valve, Rev. 7
SSF 2-HP-3 (FS/2/51/3) 2HP VA0003
OEE-251-03 Elementary Diagram, Letdown ISO
- L. Valve 2HP-5, 2 HP Rev. 16
VA0005
OEE-251-0A Elementary Diagram, Letdown Cooler A Outlet Valve, Rev. 5
SSF 2-HP-3 (FS/2/51/3) 2HP VA0003
OEE-251-10 Elementary Diagram, HP Injection Pump, Discha. Rev. 3
Crossov. Valve FS/2/51/55
OEE-251-2 Elementary Diagram,
- R.C. Pump Seal Return Valve, SSF Rev. 9
2-HP-20 (FS/2/51/39) 2HP VA0020
OEE-251-8 Elementary Diagram, RC Pump Seal Return Isol. Vlv. 2- Rev. 11
HP-21, 2HP VA0021
OEE-251-9 Elementary Diagram, Feed and Bleed Controls, Valves Rev. 13
2/51/21 (2HP-V10) (2HP-14) and 2/51/14 (2HP-
V18)(2HP-16)
OEE-265-01 Elementary Diagram, Protected Service Water, STM GEN Rev. 2
3A and 3B Flow Isolation MOV 3PSW-6
OEE-265-02 Elementary Diagram, Protected Service Water, Steam Rev. 3
Generator 2A Flow, Control MOV 2PSW-22
OEE-265-03 Elementary Diagram, Protected Service Water, Steam Rev. 3
Inspection Type Designation Description or Title Revision or
Procedure Date
Generator 2A Throttle MOV 2PSW-24
OEE-265-04 Elementary Diagram, Protected Service Water, Steam Rev. 3
Generator 2B Flow Control, MOV 2PSW-24
OEE-265-05 Elementary Diagram, Protected Service Water, Steam Rev. 3
Generator 2B Throttle MOV 2PSW-25
OEE-265-06 Elementary Diagram, Protected Service Water, HPI Rev. 2
Isolation Valve 2HP-139
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 Rev. 2
Control Loop, Valve 2PSW-22
OFD-101A-2.1 Flow Diagram of High Pressure Injection System Letdown Rev. 51
Section
OFD-101A-2.2 Flow Diagram of High Pressure Injection System Storage Rev. 46
Section
OFD-101A-2.3 Flow Diagram of High Pressure Injection System Rev. 34
Charging Section
OFD-101A-2.4 Flow Diagram of High Pressure Injection System Rev. 48
Charging Section
OFD-101A-2.5 Flow Diagram of High Pressure Injection System SSF Rev. 24
Portion
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, Rev. 2
Steam Generator and HPI Pump motor Cooling Service
Engineering EC 091869 OD300958/EC91869 Unit 3 HPI Pump System Power - Rev. 11
Changes Pre-Outage
EC 091873 OD500922 - PSW Power Feed Installation Rev. 18
EC 115193 Correction - Implement MOD to Add Fuses/Breakers to Rev. 4
loads on MCC AWC3
EC 115193 Correction - Implement MOD to Add Fuses/Breakers to Rev. 0
loads on MCC AWC3
EC 400063 Unit 3, Make Field Changes associated with 3RC66 to Rev. 0
improve the DC Voltage
ONS-2021-011
Inspection Type Designation Description or Title Revision or
Procedure Date
ONS-2021-015
Miscellaneous Oconee Nuclear Station Updated Final Safety Analysis Rev. 28
Report
Safety Evaluation by the Office of Nuclear Reactor December
Regulation, Transition to a Risk-Informed 29, 2010
Performance Based Fire Protection Program in
Accordance with 10 CFR 50.48(c)
Safety Evaluation by the Office of Nuclear Reactor April 28,
Regulation, Oconee Nuclear Station Standby Shutdown 1983
Facility
Oconee Nuclear Station License Amendment Request to 4/14/2010
Adopt NFPA 805
Oconee Selected Licensing Commitments, Section 16.9, Rev. 1
Auxiliary Systems
51-9249010-004 Oconee Nuclear Safe Capability Assessment for Unit 1, 2, 08/15/2016
and 3, Appendix B: Safe Shutdown Equipment List
(SSEL)
Duke-QAPD-001 Quality Assurance Program Description Operating Fleet Rev. A
OELBKAWC302A OELBKAWC302A EDB report, ST 1804 partial equipment 3/30/2022
key, AWC3-2A
OELBKAWC302B OELBKAWC302A EDB report, ST 1804 partial equipment 3/30/2022
key, AWC3-2B
OSC-9659 Oconee Nuclear Safety Capability Assessment for Units Rev. 011
1, 2 and 3
OSS-0254.00-00- (Mech) High Pressure Injection and Purification & Rev. 062
1001 Deborating Demineralizer Systems
OSS-0254.00-00- Protected Service Water System Rev. 003
1053
OSS-0254.00-00- Protected Service Water System Rev. 3
1053
OSS-0254.00-00- (Elec) Design Basis Specification for Fire Detection Rev. 019
2017 System
OSS-0254.00-00- (Mech) Design Basis Specification for Fire Protection Rev. 47
4008
Inspection Type Designation Description or Title Revision or
Procedure Date
OSS-0254.00-00- Design Basis Specification for Fire Protection Rev. 47
4008
OSS-0256.00-00- Design Basis Specification for High Pressure Service Rev. 47
1002 Water System
PD-FP-ALL-1500 Fleet Fire Protection Program Manual Rev. 1
PUL 92-9257585 Product Upgrade list (PUL) 05/19/2016
Procedures AD-EL-ALL-1117 Design Analyses and Calculations Rev. 10
AD-EL-ALL-1132 Preparation and Control of Design Change Engineering Rev. 4
Changes
AD-OP-ALL-0105 Operability Determinations Rev. 6
SD 3.2.14 Fire Protection Program Compensatory Measure Process Rev. 4
Work Orders 20059574 EC115193 Perform Breaker Bench Test for new AWC3 07/12/2016
Bkrs
14