IR 05000269/2019002
| ML19226A082 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 08/14/2019 |
| From: | Frank Ehrhardt Division Reactor Projects II |
| To: | Burchfield J Duke Energy Carolinas |
| Reeder D | |
| References | |
| IR 2019002 | |
| Download: ML19226A082 (27) | |
Text
ust 14, 2019
SUBJECT:
OCONEE NUCLEAR STATION - NRC INTEGRATED INSPECTION REPORT 05000269/2019002, 05000270/2019002, AND 05000287/2019002
Dear Mr. Burchfield:
On June 30, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Oconee Nuclear Station. On July 25, 2019, 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. These findings involved violations of NRC requirements. We are treating these violations as non-cited violations (NCVs) consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violations or 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.
If you disagree with a cross-cutting aspect assignment in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the U.S. Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001; with copies to the Regional Administrator, Region II; and the NRC Resident Inspector at Oconee. 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 10 CFR 2.390, Public Inspections, Exemptions, Requests for Withholding.
Sincerely,
/RA/
Frank J. Ehrhardt, Chief Reactor Projects Branch 1 Division of Reactor Projects Docket Nos. 05000269, 05000270, and 05000287 License Nos. DPR-38, DPR-47, and DPR-55 Enclosure:
Inspection Report 05000269/2019002, 05000270/2019002, and 05000287/2019002 cc w/ encl: Distribution via LISTSERV
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting an integrated 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. NRC and self-revealed findings, violations, and additional items are summarized in the table below.
List of Findings and Violations Failure to Verify Adequacy of Design of LPSW System for Auxiliary Building Flooding Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green Not Present 71111.06 Systems NCV Performance 05000269,05000270,05000287/2019002-01 (NPP)
Open/Closed Inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to verify the adequacy of design of the low pressure service water (LPSW) system. Specifically, design review calculation OSC-8671, Auxiliary Building Flood Design Values, failed to ensure safety-related equipment would not be adversely impacted by failure of non-seismic LPSW piping in the auxiliary building.
Failure to Monitor or Demonstrate Control of Performance Under the Maintenance Rule Cornerstone Significance Cross-Cutting Report Aspect Section Mitigating Green [H.1] - 71111.12 Systems NCV Resources 05000269,05000270,05000287/2019002-02 Open/Closed The inspectors identified a Green NCV of 10 CFR 50.65(a)(2) when the licensee failed to monitor the performance or demonstrate effective control of performance of the standby shutdown facility (SSF).
Additional Tracking Items Type Issue Number Title Report Section Status LER 05000287/2018-002-00 Actuation of the Keowee 71153 Closed Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses.
LER 05000287/2018-002-01 Actuation of the Keowee 71153 Closed Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses.
PLANT STATUS
===Unit 1 operated at or near 100 percent rated thermal power (RTP) for the entire inspection period.
Unit 2 operated at or near 100 percent RTP for the entire inspection period.
Unit 3 operated at or near 100 percent RTP for the entire inspection period.
INSPECTION SCOPES
Inspections were conducted using the appropriate portions of the inspection procedures (IPs) in effect at the beginning of the inspection unless otherwise noted. Currently approved IPs with their attached revision histories are located on the public website at http://www.nrc.gov/reading-rm/doc-collections/insp-manual/inspection-procedure/index.html. Samples were declared complete when the IP requirements most appropriate to the inspection activity were met consistent with Inspection Manual Chapter (IMC) 2515, Light-Water Reactor Inspection Program - Operations Phase. The inspectors performed plant status activities described in IMC 2515 Appendix D, Plant Status and conducted routine reviews using IP 71152, Problem Identification and Resolution. 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
71111.01 - Adverse Weather Protection
Summer Readiness (IP Section 03.01)
- (1) The inspectors evaluated summer readiness of offsite and alternate alternating current (AC) power systems.
71111.04 - Equipment Alignment
Partial Walkdown (IP Section 03.01)
The inspectors evaluated system configurations during partial walkdowns of the following systems/trains:
- (1) Protected service water (PSW) system on May 7, 2019, during annual SSF system outage
- (2) Keowee Hydro Station alignment on May 7, 2019, during annual SSF system outage
- (3) 230kV switchyard yellow bus and degraded voltage protection circuitry restoration on May 16, 2019, following extended system outage, maintenance and modification
- (4) Emergency AC power sources (Keowee Hydro Unit 1, SSF, CT-5, and PSW) on June 12, 2019, while Keowee Hydro Unit 2 was out of service
71111.05Q - Fire Protection
Quarterly Inspection (IP Section 03.01)
The inspectors evaluated fire protection program implementation in the following selected areas:
- (1) Unit 1 Auxiliary Building Elevation 809', Fire Zone 109 on May 11, 2019
- (2) Unit 2 Auxiliary Building Elevation 809', Fire Zone 104 on May 11, 2019
- (3) Unit 2 Auxiliary Building Elevation 822', Fire Zone 111 on May 24, 2019
- (4) Unit 3 Auxiliary Building Elevation 796', Fire Zone 89, 86, and 87 on May 24, 2019
- (5) Unit 3 Auxiliary Building Elevation 809', Fire Zone 100, 98, and 99 on May 24, 2019
71111.06 - Flood Protection Measures
Internal Flooding (IP Section 02.02a)
The inspectors evaluated internal flooding mitigation protections in the:
- (1) Unit 1, 2, and 3 auxiliary building basement high pressure injection (HPI) and low pressure injection (LPI) hatch areas
71111.07A - Heat Sink Performance
Annual Review (IP Section 02.01)
The inspectors evaluated readiness and performance of:
- (1) 3A recirculated cooling water heat exchanger
71111.11Q - Licensed Operator Requalification Program and Licensed Operator Performance
===Licensed Operator Performance in the Actual Plant/Main Control Room (IP Section 03.01) (1 Sample)
- (1) Main control room observation during a downpower to 98 percent reactor power to conduct PT/3/A/0290/003, Turbine Valve Movement, on May 31, 2019 Licensed Operator Requalification Training/Examinations (IP Section 03.02)===
- (1) The inspectors observed and evaluated licensed operator requalification training on May 22, 2019.
71111.12 - Maintenance Effectiveness
Routine Maintenance Effectiveness Inspection (IP Section 02.01)
The inspectors evaluated the effectiveness of routine maintenance activities associated with the following equipment and/or safety significant functions:
- (2) NCR 2253656, Cartridge missing from auxiliary building flood flow limiting valve 3LPSW-845 Quality Control (IP Section 02.02) (2 Samples)===
The inspectors evaluated maintenance and quality control activities associated with the
===following equipment performance activities:
- (1) SSF Outage Work Orders (WOs) including: WO 20142395 - 0CCW-277 Manual Override Screw Needs Replacement; WO 20326318 - Replace AHU 0-42 Compressor #1 Timer; WO 20260071 - PM Rebuild SSF Diesel Engine Service Water Pump; WO 20230354 - PM Switchgear OTS1
- (2) WO 20225494-02, Unit 1A HPI Flow Transmitter 7A replacement on May 28, 2019
71111.13 - Maintenance Risk Assessments and Emergent Work Control
Risk Assessment and Management Sample (IP Section 03.01)
The inspectors evaluated the risk assessments for the following planned and emergent work activities:
- (1) 1CCW-13 mechanical/electrical preventative maintenance resulting in station yellow risk condition on April 9, 2019
- (2) Oconee Nuclear Site elevated yellow risk during planned maintenance outage on the SSF on May 7, 2019
- (3) Emergent work on the yellow bus due to degraded CCVTs on May 17, 2019
- (4) Risk assessment and work controls for transport of 2SF-2 across Unit 1/2 spent fuel pool on May 30, 2019
- (5) Unit 2 low pressure turbine rotor lifts resulting in Yellow Station Risk on June 28, 2019
71111.15 - Operability Determinations and Functionality Assessments
Operability Determination or Functionality Assessment (IP Section 02.02)
The inspectors evaluated the following operability determinations and functionality assessments:
- (2) NCR 2266198, AP/1-2/A/1700/036 entry due to temporary chiller degradation
- (3) NCR 2271174, 3A Core Flood Tank vent valve will not fully close
- (4) NCR 2272266, SSF sump pump 1 discharge check valve CCW-312 found stuck open during operator rounds
71111.18 - Plant Modifications
===Temporary Modifications and/or Permanent Modifications (IP Section 03.01 and/or 03.02) (1 Sample)
The inspectors evaluated the following temporary or permanent modifications:
- (1) Evaluation Engineering Change (EC) 412504 supporting replacement and rigging of 2SF-2 for transport across spent fuel pool
71111.19 - Post-Maintenance Testing
Post Maintenance Test Sample (IP Section 03.01)===
The inspectors evaluated the following post maintenance tests:
- (1) WO 20260071-05, SSF Diesel Engine Service Water Pump test following replacement on May 7, 2019
- (2) WO 20324170, Standby Shutdown Facility Diesel-Generator operational test after SSF outage on May 9, 2019
- (3) WO 20225494-03, Testing of replacement Unit 1A HPI Flow Transmitter 7A on May 28, 2019
- (4) WO 20309525-01, 3A Essential Siphon Vacuum (ESV) train test following adjustments of pump starter relay
- (5) WO 20295098, testing of 1RC-165 following repairs
71111.22 - Surveillance Testing
===The inspectors evaluated the following surveillance tests:
Inservice Testing (IP Section 03.01)===
- (1) PT/1/A/0600/013, Motor Driven Emergency Feedwater Pump Test on the 1B MDEFW pump on April 3, 2019
- (2) PT/1/A/0251/001, Low Pressure Service Water Pump Test on the B LPSW pump on April 4, 2019
- (3) WO 20319437-03, PT/2/A/0261/10, Essential Siphon Vacuum System Test, on June 13, 2019 Routine (IP Section 03.01) (2 Samples)===
- (1) OP/0/A/1600/010, Operation of the SSF diesel-generator, on April 23, 2019
===(2) IP/2/A/0315/014 A, TXS RPS Interposing Relay Test and Control Rod Drive Breaker Trip Timing Test on Unit 2, on May 8, 2019
71114.06 - Drill Evaluation
===Select Emergency Preparedness Drills and/or Training for Observation (IP Section 03.01) (1 Sample)
- (1) Inspectors observed an emergency preparedness drill on May 15, 2019, that involved a loss of core monitoring instrumentation, a feedwater pump trip transient, elevated reactor coolant pump vibrations and fuel damage, a secondary-side steam leak, loss of all offsite and emergency AC power capability, and failure to power SSF equipment and PSW unavailable.
OTHER ACTIVITIES - BASELINE
===71151 - Performance Indicator Verification The inspectors verified licensee performance indicators submittals listed below:
BI02: RCS Leak Rate Sample (IP Section 02.11)===
- (1) Unit 1 (April 1, 2018 - March 31, 2019)
- (2) Unit 2 (April 1, 2018 - March 31, 2019)
- (3) Unit 3 (April 1, 2018 - March 31, 2019)
MS07: High Pressure Injection Systems (IP Section 02.06) ===
- (1) Unit 1 (April 1, 2018 - March 31, 2019)
- (2) Unit 2 (April 1, 2018 - March 31, 2019)
- (3) Unit 3 (April 1, 2018 - March 31, 2019)
MS10: Cooling Water Support Systems (IP Section 02.09) (3 Samples)===
- (1) Unit 1 (April 1, 2018 - March 31, 2019)
- (2) Unit 2 (April 1, 2018 - March 31, 2019)
- (3) Unit 3 (April 1, 2018 - March 31, 2019)
71152 - Problem Identification and Resolution
Semiannual Trend Review (IP Section 02.02)
- (1) The inspectors reviewed the licensees corrective action program for potential adverse trends in human and equipment performance that might be indicative of a more significant safety issue.
71153 - Followup of Events and Notices of Enforcement Discretion Event Report (IP Section 03.02)
The inspectors evaluated the following licensee event reports (LERs) which can be accessed at https://lersearch.inl.gov/LERSearchCriteria.aspx:
- (1) LER 05000287/2018-002-00, Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses
- (2) LER 05000287/2018-002-01, Actuation of the Keowee Hydroelectric Station Due to Loss of AC Power to the Unit 3 Main Feeder Buses
INSPECTION RESULTS
Observation: Semi-Annual Trend Review 71152 Inspectors identified a negative trend in human and equipment performance associated with inspection and maintaining readiness of emergency response equipment such as fire extinguishers, fire hose reels, fire brigade turnout gear, and emergency lighting stored in emergency equipment lockers. Nuclear Condition Reports 2273853, 2273814, 2271810, 2267309, 2260693, 2255541, 2253667, 2252769, and 2193141 were referenced. Because these various issues were identified by personnel from outside organizations and departments or were discovered by the fire brigade once the equipment was attempted to be used during a drill, the issues indicated there was a weakness in the routine inspection and maintenance of the equipment. One other notable trend was identified by the licensee in NCR 2276467 regarding subtle indicators that operator procedure use and adherence may be declining from station goals and standards. The licensee is developing an action plan to improve performance in this area.
Failure to Verify Adequacy of Design of LPSW System for Auxiliary Building Flooding Cornerstone Significance Cross- Report Cutting Section Aspect Mitigating Green NPP 71111.06 Systems NCV 05000269,05000270,05000287/2019002-01 Open/Closed Inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when the licensee failed to verify the adequacy of design of the LPSW system. Specifically, design review calculation OSC-8671, Auxiliary Building Flood Design Values, failed to ensure safety-related equipment would not be adversely impacted by failure of non-seismic LPSW piping in the auxiliary building.
Description:
Updated Final Safety Analysis Report Chapter 3.4.1.1.1, Current Flood Protection Measures for Turbine and Auxiliary Buildings stated that the auxiliary building could be subject to flooding from a single break in the non-seismic portions of the LPSW system and that these portions are isolated, or flow limited to allow operators sufficient time to identify and isolate the source. The section also stated that operator response times were tested to ensure flood mitigation can occur before safety related equipment is adversely affected. Design review calculation OSC-8671, Revision 7, established acceptance criteria related to auxiliary building flooding by verifying that safety-related equipment was adequately protected from internal flooding based on the ability of the operators to identify and isolate a postulated break in 45 minutes.
On March 7, 2019, inspectors identified several non-conservative evaluations within OSC-8671 for the LPSW system. These evaluations determined flooding rates for certain break locations but did not consistently consider flow from both ends of a failed pipe and the multiple sources that could supply the failure location. When considered, safety-related equipment could be adversely affected before the design basis operator response time of 45 minutes expired. Specifically: 1) The Unit 3 high pressure injection (HPI) pumps were estimated to be affected by flooding after 19 minutes, 2) The Unit 1/2 HPI pumps were estimated to be affected by flooding after 36 minutes, and 3) The Unit 3 A reactor building spray (RBS) pump (in the Unit 3 A low pressure injection (LPI) pump room) was estimated to be affected by flooding after 35 minutes.
Although the design basis was not met for these areas, time-critical operator action validations per periodic test PT/0/A/0120/033, Time Critical Action Verification, showed that operators have executed the LPSW isolation mitigating actions of abnormal procedure AP/1-2/A/1700/030, Auxiliary Building Flood, and AP/3/A/1700/030 in time frames of approximately 33 minutes and 22 minutes, respectively. For the Unit 1/2 HPI pump room and Unit 3 A LPI pump room, the time validations demonstrated that additional margin would likely be available and would allow operators time to identify and isolate additional components that were not specifically identified in the abnormal procedure. However, for the Unit 3 HPI pump room, there was a reasonable doubt on the capability of operators to isolate certain LPSW breaks before equipment would be adversely affected because the time validation results were 3 minutes longer than the time to impact the pumps. The Unit 3 HPI pump room vulnerability was determined to exist only during summer weather conditions that required the plant to operate all four condenser circulating water (CCW) pumps with the second condensate cooler in service.
Corrective Actions: The licensee took compensatory measures to isolate certain lines to improve operator response time margins. Procedure changes were made to identify additional isolation valves and ladders were installed where necessary to access the new components. On Unit 3, administrative controls were implemented to restrict alignment of the second condensate cooler until alternative measures could be developed.
Corrective Action References: NCRs 02261347 and 02266374
Performance Assessment:
Performance Deficiency: The failure to verify the adequacy of design of the LPSW system per 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. Specifically, design review calculation OSC-8671 failed to ensure safety-related equipment would not be adversely impacted by failure of non-seismic LPSW piping in the auxiliary building.
Screening: The performance deficiency was more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inspectors identified different postulated failure locations that could result in flooding rates that would individually impact either the three Unit 3 HPI pumps, the six Unit 1 & 2 HPI pumps, or the 3A RBS pump under design basis conditions.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. A regional SRA conducted a quantitative risk evaluation using SAPHIRE Version 8.1.8 and Oconee SPAR model Version 8.60. In completing the evaluation, the SRA assumed that operators would not be able to isolate auxiliary building flooding from breaks of non-seismic LPSW piping in Unit 3 before affecting safety-related equipment, were it to occur under certain seasonal conditions, and would have barely adequate time to do so for Units 1 and 2. The SRA also assumed that FLEX equipment was available. The result was a change in core damage frequency of less than 1E-6/year, which was of very low safety significance (Green). The dominant sequence was an auxiliary building flood with operator failure to isolate, various failures of the SSF, and failure to implement FLEX. Remaining mitigation capability would include operator capability to trip the RCPs, isolate the seal bleed off line, and maintain the RCS subcooled.
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: 10 CFR Part 50, Appendix B, Criterion III, Design Control, required, in part, that design control measures shall provide for checking the adequacy of design, such as by the performance of design reviews. Contrary to the above, since October 29, 2014, the design review performed for auxiliary building flooding did not verify adequacy of the design because calculation OSC-8671 failed to ensure safety-related equipment would not be adversely impacted by failure of non-seismic LPSW piping in the auxiliary building.
Enforcement Action: This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.
Failure to Monitor or Demonstrate Control of Performance Under the Maintenance Rule Cornerstone Significance Cross- Report Cutting Section Aspect Mitigating Green [H.1] - 71111.12 Systems NCV Resources 05000269,05000270,05000287/2019002-02 Open/Closed The inspectors identified a Green NCV of 10 CFR 50.65(a)(2) when the licensee failed to monitor the performance or demonstrate effective control of performance of the standby shutdown facility.
Description:
On November 7, 2018, testing during a refueling outage revealed that the 1B RCS letdown cooler outlet isolation valve, 1HP-4, operated satisfactorily from the main control room, but it could not be operated from the SSF control room because of a relay failure. Engineers performed a maintenance rule evaluation in NCR 2241697 and determined that the inability to close the isolation valve from the SSF was a functional failure. On March 1, 2019, inspectors identified that the functional failure was incorrectly applied to the high pressure injection system rather than the SSF. One of the SSF functions monitored by the maintenance rule program was the ability of RCS boundary valves to isolate and prevent RCS inventory loss. Because of other existing SSF failures within the monitoring period, the additional functional failure caused the performance criteria for the SSF to be exceeded and required an a(1) evaluation to be performed. Subsequently, because four out of five events were maintenance preventable, the Unit 1 SSF function was changed to maintenance rule a(1) status on April 17, 2019. If the failure of 1HP-4 had been correctly classified originally, the SSF would have more promptly entered a(1) status after the fifth failure in December 2018.
Corrective Actions: The licensee entered the error with the failure classification into their correction action program and performed the required evaluations.
Corrective Action References: NCR 2260902
Performance Assessment:
Performance Deficiency: The failure to monitor the performance or demonstrate effective control of performance of the SSF was a violation of 10 CFR 50.65(a)(2) and a performance deficiency.
Screening: The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency was associated with multiple functional failures that adversely affected the reliability of the safe shutdown facility.
Significance: The inspectors assessed the significance of the finding using Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the finding screened as very low safety significance (Green) because it was a process error and did not represent a degraded condition that involved an actual loss of function of a single train for greater than its TS allowed outage time.
Cross-Cutting Aspect: H.1 - The inspectors determined that the performance deficiency had a cross cutting aspect of resources in the human performance area because the apparent cause of the performance deficiency was that leaders did not ensure that resources, such as the maintenance rule manager database and design basis documents, contained accurate information for determining which maintenance rule functions were attributable to components that affected multiple systems. This led to a failure being applied to an incorrect function. This vulnerability was originally identified during a May 25, 2016, Maintenance Rule Expert Panel meeting. However, the associated assignment in NCR 2022432 was closed without action based on an assumption that system engineers would be able to determine the correct function. In this case, the presence of inaccurate information created doubt amongst the involved engineers and manager and led to an incorrect determination.
Enforcement:
Violation: 10 CFR Part 50.65(a)(1), requires, in part, that holders of an operating license shall monitor the performance or condition of structures, systems, or components (SSCs) within the scope of the rule as defined by 10 CFR 50.65(b), against licensee-established goals, in a manner sufficient to provide reasonable assurance that such SSCs are capable of fulfilling their intended functions.
10 CFR 50.65(a)(2) states, in part, that monitoring specified in 10 CFR 50.65(a)(1) is not required where it has been demonstrated that the performance or condition of an SSC is being effectively controlled through the performance of appropriate preventive maintenance, such that the SSC remains capable of performing its intended function.
Contrary to the above, from December 2018 through March 2019, the licensee failed to demonstrate that the performance of the SSF had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals. Specifically, the license did not identify and account for a maintenance preventable functional failure of the SSF system. This resulted in the licensee failing to account for four total maintenance preventable functional failures of the SSF occurring from January 2018 to December 2018. The failures demonstrated that the performance of the SSF was not being effectively controlled through the performance of appropriate preventive maintenance and, as a result, that goal setting and monitoring was required.
Enforcement Action: This violation is being treated as an NCV, 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 July 25, 2019, the inspectors presented the integrated inspection results to J. Ed Burchfield, Jr. and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection Type Designation Description or Title Revision or
Procedure Date
71111.01 Corrective Action NCR 2215114,
Documents 2216927,
26849,
230316,
230662,
231905,
257173, 2272078
Miscellaneous Nuclear Switchyard Reliability Working Group Action
Register, Open Items for Oconee Nuclear Station, as of
June 2019
Procedures System Health Report for Switchyard Power Systems Q1-2019
AP/1/A/1700/034 Degraded Grid 13
CSD-EG-ALL- Nuclear Switchyard Operating Guidelines 4
2000.0
CSD-EG-ALL- Nuclear Switchyard Interface Agreement 1
2000.1
OP/0/A/1106/040 Generator Voltage Schedule 75
71111.04 Drawings K-0700 One Line Diagram - Relays & Meters - 13.8 - 230KV 43
O-6700 One-Line Diagram - Main PSW Switchgear - 13.8/4.16KV 15
System
O-702-A One Line Diagram - 6900V & 4160V Auxiliary Sys. 37
O-702-A-002 One Line Diagram 6900V & 4160V Station Auxiliary System 21
O-702-B One Line Diagram - 4160 and 600V Essential Load 26
Centers - Auxiliary Power Systems - Standby Shutdown
Facility
Engineering 405339
Changes
Miscellaneous Unit 1 and 2
operator narrative
logs, nightshift
dated May 16,
2019
Inspection Type Designation Description or Title Revision or
Procedure Date
Procedures AD-OP-ALL-0201 Protected Equipment 5
AD-WC-ALL-0410 Work Activity Integrated Risk Management (Critical Activity 8
Plan)
OMP 5-05 100 KV Backup / Emergency Power System 2
OP/0/A/1106/019 Keowee Hydro at Oconee 106
OP/0/A/1107/016 Removal and Restoration of 230kV Switchyard Buses 17
E
OP/0/A/1650/001 PSW System 4
OP/0/A/1650/002 PSW Ventilation 4
OP/0/A/1650/005 PSW AC Power 18
OP/0/A/1650/006 PSW DC Power 9
PT/0/A/0620/009 Keowee Hydro Operation 54
Work Orders 20239003,
239004
71111.05Q Calculations OSC-9588 Oconee Fire Hazard Ignition Source Tabulation 8
Miscellaneous AD-EG-ALL-1520 Transient Combustible Control 11
AD-EG-ALL-1523 Duties of a Fire Watch 8
AD-EG-ALL-1523 Temporary Ignition Source Control 1
CSD-ONS-PFP- Unit 1 Auxiliary Building Elevation 809' 0
CSD-ONS-PFP- Unit 2 Auxiliary Building Elevation 809' 0
CSD-ONS-PFP- Unit 2 Auxiliary Building Elevation 822' 0
CSD-ONS-PFP- Unit 3 Auxiliary Building & Reactor Building Elevation 796' 0
CSD-ONS-PFP- Unit 3 Auxiliary Building Elevation 809' 0
O-0-SOG-9000- (Standard Operating Guideline) Equipment Locations 0
010
O-0-SOG-9000- (Standard Operating Guideline) Fires Located Within a 0
016 Contaminated RCA/RCZ
OSS-0254.00-00- Design Specification for Fire Protection 43
4008
Inspection Type Designation Description or Title Revision or
Procedure Date
Procedures MP/0/A/1705/019 Fire Protection- SLC - Related Fire Doors- HELB Doors - 27
Annual and Bi-Monthly Inspections
MP/0/A/1705/032 Fire Hose Stations - Common Areas - SLC Related - 43
Inspections
MP/0/A/1705/032 Fire Extinguishers - Auxiliary Building - Monthly Inspection 2
A
71111.06 Calculations OSC-10790 Oconee Nuclear Station Internal Flooding Analysis 3
OSC-7719 Study of Consequences of Pipe Break Scenarios of HPSW 3
& LPSW Piping located inside the ONS Aux Building
OSC-8671 Auxiliary Building Flood Design Values 7
Corrective Action NCR 02266374, 02261347, 2264236
Documents
Drawings O-2510B Piping Layout Miscellaneous Piping Turbine Room 22
Mezzanine Plan
O-2510J Piping Layout Miscellaneous Piping EL. 796'-6" Auxiliary 21
Building Plan & Sections
O-2510L Piping Layout Miscellaneous Piping EL. 809'+3" Auxiliary 23
Building Plan
O-2510N Piping Layout Miscellaneous Piping EL. 822'-0" Auxiliary 9
Building Plan & Sections
OFD-107C-3.1 Flow Diagram of Liquid Waste Disposal System (Low 6
Activity Waste Tank Portion)
OFD-124A-1.2 Flow Diagram of Low Pressure Service Water System 56
Turbine Building
OFD-124B-3.6 Flow Diagram of Low Pressure Service Water System 26
(LPSW) Auxiliary Building Air Handling Units
OFD-125B-1.6 Flow Diagram of Low Pressure Service Water System 43
Turbine Building
Miscellaneous Completed PT/0/A/0120/033 dated 22 March 2017
Procedures AP/1-2/A/1700/030 Auxiliary Building Flood 21
AP/3/A/1700/030 Auxiliary Building Flood 19
PT/0/A/0120/033 Time Critical Action Verification 7
71111.07A Procedures MP/0/A/1100/006 Heat Exchanger-Flat Plate-Disassembly, Cleaning, Repair 26
and Reassembly
Inspection Type Designation Description or Title Revision or
Procedure Date
TE-MN-ALL-0002 Foreign Material Exclusion Level and Controls Screening 1
Work Orders 20187590
71111.11Q Miscellaneous OFD-122A-3.1 Flow Diagram of Main Steam System (Main Steam 33
Headers 3A and 3B)
OFD-122B-3.1 Flow Diagram of HP and LP Turbine Exhaust & Steam Seal 25
System (HP Turbine Inlet, Steam Seal Header & Feedwater
Pump Turbine Steam Seals)
OFD-122B-3.2 Flow Diagram of HP & LP Turbine Exhaust & Steam Seal 16
System (HP Turbine Extraction & Moisture Separator
Reheaters)
OP-OC-SAE-R274 Simulator Exercise Guide (SAE-R274) 00a
Procedures AD-TQ-ALL-0420 Conduct of Simulator Training and Evaluation 15
OP/3/A/1102/004 Operation at Power 133
PT/3/A/0290/003 Turbine Valve Movement 23
71111.12 Corrective Action 2022432, 260902,
Documents 2241697, 174814,
2192815, 245792,
249515, 198027,
253656, 893113,
2093988,
2172442,
237927,
274665
Drawings 0-422X-3 Instrument Details Emergency HP Injection Flow 18
Transmitter
OFD-101A-1.3 Flow Diagram of High Pressure Injection System (Charging 34
Section)
OFD-101A-1.4 Flow Diagram of High Pressure Injection System (Charging 49
Section)
OM 251. - 0635 - O/L Valcor 2 3-Way Flow Regulator, Duke Item DMV-769,
001 Version D
Miscellaneous MRule Manager
Maintenance Rule Program Notebook 19
Day and Zimmerman Certification of Qualification
Inspection Type Designation Description or Title Revision or
Procedure Date
Day and Zimmerman Vision Acuity Record
Day and Zimmerman Continued Satisfactory Performance
Document
Maintenance Rule Expert Panel Meeting Minutes 25 May 2016
EVAL-2019-ON- Maintenance Rule a(1) Evaluation
SSF+-00002114
Issue Ticket
2736800
Issue Ticket
49636800
Issue Ticket
61736800
Issue Ticket
75736800
Issue Ticket
81736800
OM-0245- IB/Griswold, Automatic Flow Control Valves 0
2415.001
OSS-0254.00-00- (MECH) High Pressure Injection and Purification & 58
1001 Deborating Demineralizer Systems
OSS-0254.00-00- (MECH) Design Basis Specification for Standby Shutdown 55
1004 Facility Reactor Coolant Makeup System
OSS-0254.00-00- (Mech) Design Basis Specification for the Standby 25
1009 Shutdown Facility HVAC System
Procedures AD-EG-ALL-1210 Maintenance Rule Program 1
AD-QC-ALL-0108 Nuclear Oversight Vendor QC Acceptance 2
IP/0/A/0075/008 Rosemount Pressure Transmitter Assembly, Initial 43
Calibration, and Mounting Instructions
IP/0/A/0101/001 Low Risk Maintenance Configuration Control 18
IP/0/A/0200/029A Sealing Cable Entrance Fittings on Class 1E Devices 35
IP/0/A/2001/003 C Inspection and Maintenance of Type HK Metal-Clad 27
Switchgear, Associated Bus, and Disconnects
IP/0/A/2001/003 K Inspection and Maintenance of 600 Volt K-Line Breakers 33
IP/0/A/3011/016 Motor Control Center, Distribution Center, and Power 12
Inspection Type Designation Description or Title Revision or
Procedure Date
Panelboard Preventive Maintenance
IP/0/A/5090/001 Tube Fitting and Tubing Installation 25
IP/1/A/0202/001D High Pressure Injection System Emergency HP Injection 5
Flow Instrument Calibration
MP/0/A/1200/108 Valve-Removal and Installation of Flanged Valves or Wafer 65
Valves to Piping
MP/0/A/1200/168 Valve - Valcor - 3 Way - Disassembly, Repair, 11
Reassembly, and Adjustment
MP/0/A/1300/035 Pump - Ingersoll-Rand - 4X11 AN Type - SSF - Diesel 16
Engine Service Water Pump - Removal, Disassembly,
Repair, and Assembly
MP/0/A/3007/019 SSF - Air Conditioning Unit - Maintenance 24
A
Work Orders WO 20225494,
2143025,
20123122,
20123261,
20142395,
260071,
20326318,
230354,
273757,
223158,
2018032,
212073
71111.13 Corrective Action NCR 02272633 230KV Yellow Bus CCVTs Doble Results
Documents
Miscellaneous Engineered Lift Plan, Install new SF-2 Valve in Unit 1/2 April 4, 2019
and 3 Spent Fuel Pools
Critical Activity Unit 2 and Unit 3 SF-2 Replacement April 16,
Plan 2019
Oconee Nuclear 230 kV Switching Station Degraded Grid Protection
Station UFSAR,
2/31/2016 -
Inspection Type Designation Description or Title Revision or
Procedure Date
Chapter 8.2.1.3.1
OSS-0254.00-00- Design Basis Specification (MECH) Condenser Circulating 42
1003 Water (CCW) System
OSS-0254.00-00- (Elect) 230kV Switchyard System 23
2004
SLC 16.8.8 External Grid Trouble Protection System
Technical Emergency Power Switching Logic (EPSL) 230 kV
Specification Switchyard Degraded Grid Voltage Protection (DGVP)
Bases 3.3.19
Procedures AD-MN-ALL-0009 Nuclear Rigging, Lifting, and Material Handling 6
AD-MN-ALL-0010 Nuclear Rigging & Lifting Forms 6
AD-OP-ALL-0201 Protected Equipment 5
AD-WC-ALL-0240 On-line Risk Management Process 1
PD-MN-ALL-0009 Nuclear Rigging & Lifting Program 7
71111.15 Calculations OSC-11233 Standby Shutdown Facility (SSF) Thermal Environment 0
Evaluation Using GOTHIC (SSF, HVAC)
Corrective Action 02266198,
Documents 02271172,
272266,
271172,
271174
Drawings OFD-102A 3.3, Flow Diagram of Low Pressure Injection System (Core 24
Flood)
OFD-133A-2.5 Flow Diagram of Condenser Circulating Water System 60
(SSF Aux. Service)
OM 245.-0675.001 150 LB, C.S. Swing Check Valve
Miscellaneous Oconee Nuclear Station - Technical Specifications 11/28/2018
Oconee Nuclear Station - Updated Final Safety Analysis
Report, Effective Date of Contents: December 31, 2017
Operator Narrative Logs May 3 and 4,
2019
Oconee Nuclear Station - Selected Licensee Commitments 11/28/2018
OSS-0254.00-00- (MECH) Design Basis Document for the Low Pressure 58
28 Injection and Core Flood System (LPI)
Inspection Type Designation Description or Title Revision or
Procedure Date
OSS-0254.00-00- (MECH) Design Basis Spec for Keowee Governor Oil 14
1045 (OG)System
OSS-0254.00-00- (ELECT) Keowee Emergency Power Design Basis 31
2005 Document
Procedures AD-OP-ALL-0105 Operability Determinations and Functionality Assessments 4
AD-OP-ALL-1000 Conduct of Operations 15
AP/1-2/A/1700/036 Degraded Control Room Area Cooling 20
AP/3/A/1700/036 Degraded Control Room Area Cooling 9
MP/0/A/1200/078 Valve - Pacific - Swing Check - Disassembly, Repair, and 23
Reassembly
OP/3/A/1104/001 Core Flooding System 80
TT/0/A/2200/026 KHS MVAR Capability Verification 1
Work Orders 20138638,
20141674,
20327748
71111.18 Calculations OSC-1160 Postulated Cask Drop Accident-Poison Racks 6
Engineering EC 412504 Evaluation of SF-2 Valve Installation Sequence Over Unit 1/2 0
Changes and 3 Spent Fuel
Miscellaneous AR 2212375 Evaluation of SF-2 Installation
Procedures AD-EG-ALL-1132 Preparation and Control of Design Change Engineering 14
Changes
AD-EG-ALL-1134 Preparation and Control of Evaluation Engineering 2
Changes
AD-EG-ALL-1137 Engineering Change Product Selection 6
71111.19 Corrective Action 02269120,
Documents 02272448,
276624,
277750
Drawings OFD-101A-1.3 Flow Diagram of High Pressure Injection System (Charging 34
Section)
OFD-101A-1.4 Flow Diagram of High Pressure Injection System (Charging 49
Section)
OFD-133A-2.5 Condenser Circulating Water System (SSF Aux. Service) 60
Inspection Type Designation Description or Title Revision or
Procedure Date
Miscellaneous Duke Energy ASME IST Program 27, July
2012
2004
OSS-0254.00-00- (Mech) Design Basis Specification for the Standby 50
1008 Shutdown Facility Diesel Support System
Procedures AD-EG-ALL-1155 Post Modification Testing 4
AD-EG-ALL-1720 Inservice Testing (IST) Program Implementation 4
AD-PI-ALL-0100 Corrective Action Program 21
IP/0/A/0380/006 Standby Shutdown Facility (SSF) Diesel Load and Speed 35
Control
IP/1/A/0202/001D High Pressure Injection System Emergency HP Injection 5
Flow Instrument Calibration
OP/0/A/1104/052 SSW System 35
OP/3/A/1104/012 CCW System 108
PT/0/A/0400/004 SSF Diesel Engine Service Water Pump Test 41
PT/0/A/0600/021 Standby Shutdown Facility Diesel-Generator Operation 16
PT/1/A/0152/017 Reactor Coolant System Valve Stroke Test 33
PT/3/A/0261/010 Essential Siphon Vacuum System Test 25
Work Orders WO 20270234-02,
20324170,
225494,
20309525,
260071,
20301777,
295098
71111.22 Corrective Action 02207594
Documents
Drawings O-800-E One Line Diagram - AC Elementary Diagram - 3 Line 19
Connection Diagram - Transformer NO.CT5 & OCB NO.
101
OFD-121A-1.7 Flow Diagram of Condensate System (Upper Surge Tanks 45
1A & 1B, Upper Surge Tank Dome, & Condensate Storage
Tank)
Inspection Type Designation Description or Title Revision or
Procedure Date
OFD-121D-1.1 Flow Diagram of Emergency Feedwater System 38
OFD-124A-1.1 Flow Diagram of Low Pressure Service Water System 54
Turbine BLDG. (Low Pressure Service Water Pumps)
OFD-124A-1.3 Flow Diagram of Low Pressure Service Water System 24
(Turbine Bldg. Services)
OFD-129A-1.1 Flow Diagram of Siphon Seal Water System 10
OFD-129A-2.2 Flow Diagram of Siphon Seal Water System (CCW Pump 7
Cooling)
OFD-130A-2.1 Flow Diagram of Essential Siphon Vacuum (ESV) System 10
OFD-133A-1.1 Flow Diagram of Condenser Circulating Water System 32
(CCW Intake Pumps Discharge)
OFD-121A-1.8 Flow Diagram of Condensate System (Condensate Make- 25
up & Emergency FDW Pump Suction)
OSFD-133A-1.2 Summary Flow Diagram of Condenser Circulating Water 0
System
Procedures IP/0/B/0275/005 H Emergency Feedwater System Motor Driven Emergency 61
Feedwater Pump - Non-Safety Related Instrument
Calibration
IP/2/A/0315/014 A TXS RPS Interposing Relay Test and Control Rod Drive 2
Breaker Trip Timing Test
OP/0/A/1600/010 Operation of the SSF Diesel-Generator 95
PT/0/A/0600/021 Standby Shutdown Facility Diesel-Generator Operation 16
PT/1/A/0251/001 Low Pressure Service Water Pump Test 107
PT/1/A/0600/013 Motor Driven Emergency Feedwater Pump Test 75
PT/2/A/0261/010 Essential Siphon Vacuum System Test 30
Work Orders WO 20319437-03,
20109052,
257191
71114.06 Corrective Action 2273912,
Documents 2276286, 2185296
Miscellaneous EPDR-ONS-19-01 Drill Critique Report
OMP 4-02 Verification of Written Correctness for EP/1/A/1800/001 0B 5
Procedures EP/1/A/1800/001 EOP Unit 1 Blackout 4 and 5
0B
Inspection Type Designation Description or Title Revision or
Procedure Date
71151 Drawings 0-2702-A Unit 3 Elementary Diagram AC Circuits Transformer No. Rev. 22
CT3 & 3T
Miscellaneous Mitigating Systems Performance Index (MPSI) Basis 26
Document
MSPI Derivation Report for period ending March 2019,
Units 1, 2, 3 Cooling Water System
MSPI Indicator Margin Report for period ending March
2019, Units 1, 2, 3 High Pressure Injection System &
Cooling Water System
MSPI Derivation Report for period ending March 2019,
Units 1, 2, 3 High Pressure Injection System
MSPI Margin Report for period ending March 2019, Units 1,
2, 3 High Pressure Injection System
Nuclear Energy Institute 99-02, Regulatory Assessment
Performance Indicator Guideline
Unit 1, 2, and 3 PI Summary Document for RCS Leakage
Calculation
Equipment Reliability Suite (ERS)
OSFD-107A-1 Summary Flow Diagram of Coolant Storage System 6A
Quench Tank Portion
WCAP-16423-NP Pressurized Water Reactor Owners Group Standard 0
Process and Methods for Calculating RCS Leak Rate for
Pressurized Water Reactors
Procedures PT/1 2 Reactor Coolant Leakage
3/A/0600/010
71152 Procedures MP/0/A/1705/032 Fire Hose Stations - Common Areas - SLC Related - 43
Inspections
MP/0/A/1705/032 Fire Extinguishers - Auxiliary Building - Monthly Inspection 2
A
MP/0/A/1705/032 Fire Extinguisher Carts - Inspection 2
F
PT/0/B/0120/032 Field Equipment and Procedures Surveillance 53
PT/0/B/0250/030 Quarterly Fire Brigade Equipment Inspection 22
71153 Corrective Action NCR 02203967 CT-3 Lockout 05/17/2018
Documents NCR 02205188 Monitoring of the Vital DC ground bus voltages 05/09/2018
NCR 02205189 I/R Positive DC Ground on Vital DC System 05/09/2018
NCR 02205589 Summary of Loss of Power Events for U3 05/11/2018
Drawings 0-0702-A-002 Units 1-3 One Line Diagram 6900V & 4160V Station Rev. 21
Auxiliary System
0-2700 Unit 3 One Line Diagram Relays & Meters 19KV Rev. 29
0-2702-A Unit 3 One Line Diagram 6900V & 4160V Station Auxiliary Rev. 16
System
0-2705 Unit 3 One Line Diagram 120 VAC & 125 VDC Station Rev. 87
Auxiliary Circuits Instrumentation Vital Buses
Miscellaneous Oconee Unit 3 Operator Log 05/10/2018
Work Orders WO 20252409 CT3, Support Lockout Trip Investigation 05/10/2018
WO 20252422 3B FWPT Terminal Cabinet, Grounded Cable 3TP404 05/10/2018
24