IR 05000424/2019010
| ML19347C839 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 12/13/2019 |
| From: | James Baptist NRC/RGN-II/DRS/EB1 |
| To: | Gayheart C Southern Nuclear Operating Co |
| References | |
| IR 2019010 | |
| Download: ML19347C839 (19) | |
Text
December 13, 2019
SUBJECT:
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 & 2 - DESIGN BASIS ASSURANCE INSPECTION (TEAMS) INSPECTION REPORT 05000424/2019010 AND 05000425/2019010
Dear Ms. Gayheart:
On November 8, 2019, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at Vogtle Electric Generating Plant, Units 1 & 2 and 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.
One finding of very low safety significance (Green) is documented in this report. This finding involved a violation of NRC requirements. We are treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2 of the Enforcement Policy.
If you contest the violation or the significance or severity of the violation 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 Vogtle Electric Generating Plant, Units 1 & 2.
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 Vogtle Electric Generating Plant, Units 1 & 2.
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,
/RA/
James B. Baptist, Chief Engineering Branch 1 Division of Reactor Safety
Docket Nos. 05000424 and 05000425 License Nos. NPF-68 and NPF-81
Enclosure:
As stated
Inspection Report
Enclosure
Docket Numbers:
05000424 and 05000425
License Numbers:
Report Numbers:
05000424/2019010 and 05000425/2019010
Enterprise Identifier: I-2019-010-0030
Licensee:
Southern Nuclear Operating Co., Inc.
Facility:
Vogtle Electric Generating Plant, Units 1 & 2
Location:
Waynesboro, GA
Inspection Dates:
October 21, 2019 to November 08, 2019
Inspectors:
P. Braxton, Reactor Inspector
P. Carman, Senior Reactor Inspector
T. Fanelli, Senior Reactor Inspector
N. Floyd, Senior Reactor Inspector
Approved By:
James B. Baptist, Chief
Engineering Branch 1
Division of Reactor Safety
SUMMARY
The U.S. Nuclear Regulatory Commission (NRC) continued monitoring the licensees performance by conducting a design basis assurance inspection (teams) inspection at Vogtle Electric Generating Plant, Units 1 & 2, 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 corrective action to implement time critical operator action Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity Green NCV 05000425/2019010-01 Open/Closed
[P.2] -
Evaluation 71111.21M NRC inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B,
Criterion XVI, Corrective Action, because SNC did not assure that a deficiency associated with the steam generator atmospheric relief isolation valve was promptly identified and corrected, as required by procedure NMP-GM-002-001, Corrective Action Program Instructions. Specifically, the valve was difficult to operate and would have resulted in SNC staff not being able meet the completion time for a time critical operator action during an analyzed accident.
Additional Tracking Items
Type Issue Number Title Report Section Status URI 05000424,05000425/20 19010-02 Corrective Actions for Single Failure Vulnerabilities Identified in Condition Report, CR10606308 71111.21M Open
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
===71111.21M - Design Bases Assurance Inspection (Teams)
The inspectors evaluated the following components and listed applicable attributes, permanent modifications, and operating experience:
Design Review - Risk-Significant/Low Design Margin Components (IP Section 02.02) (6 Samples)
- (1) Emergency Diesel Generator 2A (Electrical)
- Material condition and configuration (e.g., visual inspection during a walkdown)
- Operating environment
- Consistency between station documentation (e.g. procedures) and vendor specifications
- Maintenance effectiveness
- Corrective maintenance records, and corrective action history
- Generator loading
- Voltage at key components
- Surveillance and maintenance testing
- (2) Unit 2 Sequencer
- Maintenance effectiveness
- Compliance with UFSAR, TS, and TS Bases
- Material condition and configuration (i.e. visual inspection during walkdown)
- Adequacy of corrective action activities
- Verify assessment under Cyber Security Program
- Verify translation of vendor software specifications into drawings and operational procedures
- Degraded Grid/ Undervoltage Relay Settings
- (3) Unit 2 NSCW transfer pump (2-1202-P4-007)
- Material condition and configuration review performed during a visual non-intrusive inspection to assess material condition, the presence of hazards, and consistency of installed equipment with design documentation and analyses
- Normal operating procedures
- Calculations: (pump head, capacity)
- Alternate transfer provisions for pump out of service
- (4) NSCW cooling tower basin 2A and 2B
- Calculations: (ultimate heat sink analysis, capacity, chemical treatment)
- Normal and emergency operating procedures
- Maximum basin post-accident temperature
- Adequacy of capacity for required mission time
- (5) Unit 2 Emergency Diesel Generator Fuel Oil Storage Tank (2-2403-T4-001)
- Adequacy of tank capacity
- Capability to refill tank to accomplish mission time
- Availability of fuel for offsite sources
- (6) Unit 2 Emergency Diesel Generator Jacket Water Heat Exchanger (2-2403-G4-001-E03)
- Material condition and configuration review performed during a visual non-intrusive inspection to assess material condition, the presence of hazards, and consistency of installed equipment with design documentation and analyses
- Thermal performance of heat exchanger with maximum NSCW temperature
- Performance testing of heat exchanger
Design Review - Large Early Release Frequency (LERFs) (IP Section 02.02)===
(1)125VDC Switchgear 2AD1
- Material condition and configuration (e.g., visual inspection during a walkdown)
- Operating environment
- Consistency between station documentation (e.g. procedures) and vendor specifications
- Maintenance effectiveness
- Corrective maintenance records, and corrective action history
- Breaker short circuit capacity
- Bus loading
- Overcurrent protection and coordination (2)480V System
- Compliance with UFSAR, TS, and TS Bases
- Visual inspection during walkdown of various components
- Adequacy of corrective action activities
- Coordination and overcurrent protection
- Design requirements (Voltage and Current, etc)
- Verify components conformance with manufacturer instructions for installation, maintenance, testing and operation
Modification Review - Permanent Mods (IP Section 02.03) (2 Samples)
- (1) NO08/2R14 480V Breaker & 4160/480V Transformer Replacement
- (2) Unit 2 Safety Features Sequencer Control Logic Upgrade
Review of Operating Experience Issues (IP Section 02.06) (2 Samples)
- (1) NRC Information Notice 2010-23, Malfunctions of Emergency Diesel Generator Speed Switch Circuits
- (2) NRC Information Notice 2016-09, Recent Issues Identified When Using Reverse Engineering Techniques in the Procurement of Safety-Related Components
INSPECTION RESULTS
Inadequate corrective action to implement time critical operator action Cornerstone Significance Cross-Cutting Aspect Report Section Barrier Integrity
Green NCV 05000425/2019010-01 Open/Closed
[P.2] -
Evaluation 71111.21M NRC inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because SNC did not assure that a deficiency associated with the steam generator atmospheric relief isolation valve was promptly identified and corrected, as required by procedure NMP-GM-002-001, Corrective Action Program Instructions. Specifically, the valve was difficult to operate and would have resulted in SNC staff not being able meet the completion time for a time critical operator action during an analyzed accident.
Description:
On April 27, 2019, the manual inlet isolation valve (component # 2-1301-U4-137) to the steam generator atmospheric relief valve was extremely difficult to close and required the force of more than two staff. SNC discovered this degraded condition during the quarterly in-service test of the atmospheric relief valve, which also included stroking the inlet isolation valve closed. SNC staff generated CR 10605455 and documented that the isolation valve was operable because it was still able to be manipulated and because there was a second isolation valve on the steam line that could be used. SNC did not perform any corrective actions or compensatory measures at that time.
The inlet isolation valve has a safety function to be manually closed by an operator as part of a time critical action to mitigate a steam generator tube rupture with a stuck open atmospheric relief valve. The required completion time for this action is 16 minutes as documented in NMP-OS-014-003, VNP Time Critical Operator Action Program, Version
4. The inspectors determined that SNC did not have a reasonable basis of operability
because the time critical action would not have been able to be completed within the required time frame. The inspectors noted that the excessive force would have significantly challenged a single operator. The inspectors further noted that the second isolation valve (component # 2-1301-U4-002) was not credited in the accident analysis nor part of the in-service testing program and that there was no procedural guidance for an operator to grab the associated key to unlock and operate this valve.
NMP-GM-002-001, Corrective Action Program Instructions, Version 38.1, describes the process for initiating condition reports, screening issues and assigning actions for resolution. Attachment 1, Severity Level Descriptions, Criteria, and Examples, states that conditions affecting nuclear safety, affecting safety related equipment, or quality as defined in 10 CFR 50 Appendix B, criterion I - XVIII shall be screened as Severity Level 2 and resolved by either: a SL2 corrective action report, SL2 work order, or by direct disposition of the condition report. The inspectors determined that SNC did not perform any of these required actions.
The inspectors discussed the impact of the degraded isolation valve with the plant operations staff. Subsequently, SNC generated CR 10611294 to classify the valve as operable, but degraded and non-conforming, and implemented corrective actions to ensure that the time critical action could be accomplished. In response to further discussions with the inspectors, SNC generated CR 10659467 to review other components like this isolation valve where degradation or impairments could impact completion times for critical operator actions.
Corrective Actions: SNC staff completed corrective actions, which included hanging a caution tag on valve 2-1301-U4-137 on May 20, 2019 and adding a compensatory action in the operations shift log on May 22, 2019 that provided direction to operate the second isolation valve.
Corrective Action References: CR 10605455 and CR 10611294
Performance Assessment:
Performance Deficiency: SNC staff failed to identify and assign corrective actions for a condition adverse to quality in accordance with NMP-GM-002-001, Corrective Action Program Instructions. Specifically, SNC incorrectly evaluated the impact of the degraded valve on the performance of a time critical action and did not perform any corrective actions to ensure the completion time would be met.
Screening: The inspectors determined the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. This issue was similar to Example 4.d in NRC Inspection Manual Chapter 0612, Appendix E, such that the degraded condition significantly impacted the operators ability to do the task.
Significance: The inspectors assessed the significance of the finding using Appendix H, Containment Integrity SDP. The issue screened to green because the degraded SSC had no direct impact on the likelihood of core damage and was not important to LERF.
Cross-Cutting Aspect: P.2 - Evaluation: The organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. SNC staff did not thoroughly evaluate the impact of the degraded condition on the safety-related function of the valve, which required timely action to mitigate the consequences of an accident.
Enforcement:
Violation: 10 CFR Part 50, Appendix B, Criterion XVI, states in part that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected.
Contrary to the above, from April 27, 2019 to May 22, 2019, the licensee did not assure that a deficiency associated with the steam generator atmospheric relief isolation valve was promptly identified and corrected. Specifically, the isolation valve was difficult to operate and would have resulted in the licensee not being able meet the completion time for a time critical action during an analyzed accident (i.e. steam generator tube rupture).
Enforcement Action: This violation is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.
Unresolved Item (Open)
Corrective Actions for Single Failure Vulnerabilities Identified in Condition Report, CR10606308 URI 05000424,05000425/2019010-02 71111.21 M
Description:
The team identified an unresolved item (URI) for the licensees corrective actions associated with a single-failure vulnerability that is revealed when implementing emergency operating procedures (EOPs).
The EOP 19010-1, E - 1 Loss of Reactor or Secondary Coolant, version 2, Step 17, directs the operators to shut down a train of engineered safety features (ESF) to preserve their long-term Nuclear Service Cooling Water (NSCW) inventory. The licensee identified a single failure hazard when implementing this procedure, and documented it in condition report, CR10606308 (CR). The hazard resulted from previously unidentified design flaws in the emergency diesel generators (EDGs) starting system. The procedure stopped one of the operating EDGs during an accident. A short time after this manual stop of the EDG, the EDG becomes incapacitated because the starting air receiver and batteries drain. The EDGs are part of the ESF. The NSCW is the site ultimate heat sink. Neither the procedures nor the updated final safety analysis reports (UFSAR) discussed the ESF train becoming incapacitated from these design flaws during an accident. The licensees immediate determination of operability (IDO) for the condition report stated, This CR is identifying scenarios that are beyond the current license basis or represent departures from recommended accident response (as stated in the FSAR). In summary of the licensees position, the scenarios were judged, beyond the current license basis, because the UFSAR Section 9.2.5, "Ultimate Heat Sink," specified, in part, that the two tower basins of the NSCW meet the combined storage capacity requirements without makeup if operated in conformance with Nuclear Regulatory Commission (NRC) Regulatory Guide 1.27. Section 9.2.5.1, "Design Bases," stated, in part, the design basis meets RG 1.27.
The UFSAR specified several accidents that required the NSCW. The design requirements postulated an accident along with a loss of offsite power, and an electrical or piping system single failure. The requirements for postulating the Single Failure Criterion for electrical and piping system cases were discussed in multiple chapters of the UFSAR. The two cases were postulated differently. For electrical single failures, any single failure within the protection system shall not prevent proper protective action at the system level when required, and there were no differences for active and passive failures. However, for piping systems there were differences between active and passive failures. Only active failures are postulated in the short-term period of an accident, while active or passive failures could be postulated in the long-term period of an accident. The UFSAR acknowledged the Single Failure Criterion in chapters one, two, three, four, five, six, seven, eight, nine, ten, eleven, fifteen, and eighteen and did not specify any exceptions to it. Some of these instances are as stated below.
The UFSAR for electrical systems:
The UFSAR Section 3.1.2, Protection by Multiple Fission Product Barriers, for electrical power systems, stated, in part, a failure of a single component will not prevent the safety-related systems from performing their function. The diesel generators are arranged so that a failure of a single component will not prevent the safe shutdown of the reactor.
The UFSAR Section 3.1.3, Protection and Reactivity Control Systems, stated, in part, the design basis for all protection systems is in accordance with the guidelines of Institute of Electrical and Electronic Engineers (IEEE) Standards 279-1971 and 379-1972.
- The IEEE 279-1971 standard specified two relevant criteria described by the NRC positions for electrical single failures described in the SECY paper; 4.2, any single failure within the protection system shall not prevent proper protective action at the system level when required, and, that 4.16, the protection system shall be so designed that, once initiated, a protective action at the system level shall go to completion. Return to operation shall require subsequent deliberate operator action.
- The relevant criteria from the IEEE 379-1972 standard was, "The system shall be capable of performing the protective actions required to accomplish a protective function in the presence of any single detectable failure within the system (this is the "single failure") concurrent with all identifiable, but non-detectable failures, all failures occurring as a result of the single failure, and all failures which would be caused by the design basis event requiring the protective function."
The UFSAR for fluid systems:
The UFSAR Section 6.3.2.5, System Reliability, for the emergency core cooling system (ECCS), stated, in part,
- 6.3.2.5.1, Active Failure Criteria, an active failure is the failure of a powered component such as a piece of mechanical equipment, a component of the electrical supply system, or instrumentation and control equipment to act on command to perform its design function. Examples include the failure of a motor-operated valve to move to its correct position, the failure of an electrical breaker or relay to respond, the failure of a pump, fan, or diesel generator to start, etc. the ECCS can sustain an active failure in either the short or long term and still meet the required level of performance for core cooling.
- 6.3.2.5.2, Passive Failure Criteria, A passive failure is the structural failure of a static component which limits the component's effectiveness in carrying out its design function. Examples include cracks in pipes, sprung flanges, valve packing leaks, or pump seal failures. the ECCS can sustain a single passive failure during the long-term phase and still retain an intact flow path to the core to supply sufficient flow to keep the core covered and to effect the removal of decay heat.
In 1977, the staff submitted a commission paper (SECY) to describe how the staff used the Single Failure Criterion as a tool in the review of the design of nuclear power plants (SECY 77-439). The SECY discussed the acceptable interpretations of IEEE 279-1972, IEEE 379-1972, and the Single Failure Criterion as applied to piping systems. A summary of the NRC positions from SECY 77-439 were as follows:
The General Design Criteria make it clear that for electrical, instrumentation and control systems, application of the Single Failure Criterion to systems evaluation depends not only on the initiating event that invokes safety action of these systems, together with consequential failures, but also on active or passive electrical failures which can occur independent of the event. Thus, evaluation proceeds on the proposition that single failures can occur at any time. In contrast, for various fluid systems, the most limiting single active failure is considered in evaluating systems performance capability within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Twenty-four hours or greater after the event, a single passive failure equal to the leakage that would occur from a valve or pump seal failure, is assumed.
In addition, the UFSAR chapter 8 identified that the standard for the power systems was IEEE 308-1974, IEEE Standard Criteria for Class 1E Power Systems for Nuclear Power Generating Stations. Section 4.8 of the standard stated, in part, an analysis of the failure modes of Class 1E power systems and the effect of these failures on the electric power available to Class 1E loads shall be performed to demonstrate that a single component failure does not prevent satisfactory performance of the minimum Class 1E loads required for safe shutdown and maintenance of post-shutdown or post-accident station security.
Planned Closure Actions: The licensee indicated that information could be provided to demonstrate that the CLB was being met. The team is awaiting information from the licensee to determine if a violation exists.
Licensee Actions: The licensee has not performed any corrective actions for this URI
EXIT MEETINGS AND DEBRIEFS
The inspectors verified no proprietary information was retained or documented in this report.
- On November 8, 2019, the inspectors presented the design basis assurance inspection (teams) inspection results to B. Keith Taber and other members of the licensee staff.
- On December 11, 2019, the inspectors presented the amended inspection results to B.
Keith Taber and other members of the licensee staff.
DOCUMENTS REVIEWED
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
71111.21M Calculations
GP-16989
SGTR Reanalysis Results for Revised Operator Action
Times
07/20/1999
GP-19499
SGTR Margin to Overfill Report to Address NSAL-07-11
2/27/2016
X3CA22
Unit 2 Load Study
Rev. 10
X3CA26
Protective Relaying Calculation
Rev. 15
X3CA27
Degraded Grid/Undervoltage Relay Setting For Unit 1 and
Unit 2
Rev. 7
X3CC02
480V Breaker Short Circuit Sizing
Rev. 10
X3CE01
EDG Steady State Loading Study
Rev.10
X3CF07
D.C. Breaker Sizing
Rev.19
X3CF12
Rev. 10
X3CK08A
Class 1E DC Power Cable Sizing
Rev. 10
X3CT08
Fire Event Safe Shutdown Circuit Analysis
Rev. 29
X4C1202S26
Ultimate Heat Sink Analysis
Rev. 5.0
X4C1202S27
NSCW Temperature to Diesel
Rev. 2
X4C1202V03
Verification of NSCW Constant Heat Loads and Flows and
Cooldown Heat Loads
Rev. 9
X4C1202V05
Nuclear Service Cooling Water (NSCW) Transfer Pump
Verification
Rev. 4
X4C1202V08
NSCW/COPATTA-11 Input Data/LOCA Design Case
Rev. 9.0
X4C1202V54
Maximum Ultimate Heat Sink Temperature (Post LOCA)
Rev. 2.0
X4C2403E01
EDG Jacket Water Heat Exchanger Fouling Factor
Rev. 4
X4C2403V08
Standby Diesel Generator Fuel Oil Consumption and
Storage Tank Capacity
Rev. 3
X4C2403V11
Emergency Diesel Generator Lube Oil Inventory Technical
Specification Values
Rev. 2
X4C2403V11
Determination of Leak Rate Acceptance Criteria for Plant
Vogtle Units 1 & 2 EDG Air Start Receiver Inlet Check
Valves
Rev. 2
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Corrective Action
Documents
10484659,
10606308,
10611186,
10648912
Corrective Action
Documents
Resulting from
Inspection
10657679
NMP-OS-014-003-F02 form incomplete
10/21/2019
CR10657888
EDG Housekeeping
10/22/2019
CR10657972
2019 NRC Design Bases Assurance Inspection (DBAI),
EDG walkdown observations
10/22/2019
CR10657979
Thermostat face cover has fallen off in 2A EDG building
10/22/2019
CR10658005
2019 NRC DBAI U2 ARV walkdown condition found for
housekeeping
10/22/2019
CR10658006
2019 NRC DBAI ARV walkdown condition found for valve
chains
10/22/2019
CR10658381
NRC DBAI - EDG drawing clarification
10/23/2019
CR10658410
10/23/2019
CR10658417
10/23/2019
CR10658726
Request for Valve Manipulation (Handwheel Count)
10/24/2019
CR10659440
2019 NRC DBAI - NSCW Peak Temperature
10/28/2019
CR10659467
2019 NRC DBAI - TCOA Recommendations
10/28/2019
CR10661822
11/5/2019
CR10662024
NRC DBAI Identified - Time Critical JPMs
11/6/2019
Drawings
Elementary Diagram Diesel Engine Control Diesel Engine-
Generator DG1A
Rev.10
2X3AE13-00005-
Diesel Generator SFS Load Sequencer Logic Diagram
Sheet 1
Rev. 5
2X3AE13-00005-
Diesel Generator SFS Load Sequencer Logic Diagram
Sheet 2
Rev. 6
2X3AE13-00005-
Diesel Generator SFS Load Sequencer Logic Diagram
Sheet 3
Rev.4
2X3AE13-00005-
Diesel Generator SFS Load Sequencer Logic Diagram
Sheet 4
Rev.3
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2X3AE13-00005-
Diesel Generator SFS Load Sequencer Logic Diagram
Sheet 5
Rev.3
2X3AE13-00005-
Diesel Generator SFS Load Sequencer Logic Diagram
Sheet 6
Rev. 3
Main One Line Unit 2
Rev. 19
One Line Diagram 125V DC Class 1E Distr Train A 2-1806-
S3-DSA, 2-1806-S3-DCA
Rev. 14
One Line Diagram Diesel - Generators 2A &
2B
Relays & Meters
Rev. 11
Elementary Diagram Electrical System 4160V Incm. Brkr
2-2AA0219 from Emergency Diesel Gen. 2A
Rev.8
Elementary Diagram Diesel Engine Control Diesel Engine -
Generator Data
Rev. 9
2X4DB133-1
P&ID Nuclear Service Cooling Water System
Rev. 54
2X4DB133-2
P&ID Nuclear Service Cooling Water System
Rev. 55
2X4DB134
P&ID Nuclear Service Cooling Water System
Rev. 31
2X4DB135-1
P&ID Nuclear Service Cooling Water System
Rev. 28
2X4DB135-2
P&ID Nuclear Service Cooling Water System
Rev. 29
2X4DB159-2
P& I Diagram Main Steam System
Rev. 33
2X4DB170-1
P&ID Diesel Generator System Train A
Rev. 42
2X4DB170-2
P&ID Diesel Generator System Train B
Rev. 44
D-72-12500-710
Schematic - Regulator Chassis
Rev. F
Engineering
Evaluations
965178
7/28/2016
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
Miscellaneous
SNC Response to IER L1-17-5 Line of Sight to the Reactor
Core
Rev. 2
2
License Amendment
09/19/17
2X4AK01-00590
Equipment Data
Rev. 3
AX3AC02B-
20016
ABB 480V Breakers Installation and Maintenance Manual
Rev. 5
AX3AE13-00001
Vogtle Unit 1/2 Diesel Generator SFS-Functional
Requirements
Rev. 4
AX3AE13-00011
Diesel Generator Safety Features Sequencer Software
Design Description
Rev. 6
AX3AE13-00113
SAFETY FEATURES SEQUENCER FAILURE MODES
AND EFFECTS ANALYSIS
Rev. 1
AX4AK01-00037
Standby Diesel Gen. - Generator Data
2/20/1979
AX4AK01-00564
Associated Publications Manual Book 2 Standby Diesel
Generator, Volume III
Rev.46
CDA-VNP-21805-
004
Vogtle Unit 2 Critical Digital Asset Functional Group
Assessment-EMAX Breakers SR
Rev. 2
CDA-VNP-21821-
001
Vogtle Unit 2 Critical Digital Asset Baseline Assessment-
2182
Rev. 4
DC-1000-E
General Design Criteria-Electrical
Rev. 20
DC-1009
Plant Single-Failure Criteria Design Bases
Rev. 9
DC-1202
Nuclear Service Cooling Water System Design Bases
Rev. 13
DC-1202-A
Nuclear Service Cooling Towers
Rev. 11
DC-1301
Main Steam System Design Bases
Rev. 11
DC-1804
4160V AC System
Rev. 10
DC-1805
480V AC System
Rev.13
DC-1821
Standby Power System
Rev. 12
DC-2403
Emergency Diesel Generator Systems
Rev.10
E220 (in plant)
Operator Response Time Validation Sheet
2/28/2017
E220 (simulator)
Operator Response Time Validation Sheet
2/26/2017
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
E221 (in plant)
Operator Response Time Validation Sheet
2/22/2016
GN-857
Procedures for Ordering Diesel Fuel Oil
04/03/1986
Supplement 2
Safety Evaluation Report
5/86
SQTS-01-MS-
Seismic Operability Data Sheet for Motor Starter or
Rev.1
Standing Order 2-
2019-02
Alternate ARV Isolation Valve
Rev. 1
TE76709
Engineering to Address NRC IN2010-23 As Requested by
CR 2011101525
06/02/2011
X4AK01
Specification for Standby Diesel Generators
Rev. 8
XFAK01
Specification for Standby Diesel Generator
Rev.8
Operability
Evaluations
OD 1-15-001
Rev. 1.0
Procedures
13145A-1
Diesel Generator Train A
Rev. 11
13150A-2
Train A Nuclear Service Cooling Water System
Rev. 13.1
13601-2
Steam Generator and Main Steam System Operation
Rev. 58
18021-C
Loss of Nuclear Service Cooling Water System
Rev. 20
19010-1
E-1 Loss of Reactor or Secondary Coolant
Rev. 2
19010-2
E-1 Loss of Reactor or Secondary Coolant
Rev. 2
19020-2
E-2 Faulted Steam Generator Isolation
Rev. 1.1
19030-2
E-3 Steam Generator Tube Rupture
Rev. 4
19100-2
ECA - 0.0 Loss of all AC Power
Rev. 5.5
27710-C
25 VDC Circuit Breaker Inspection and Testing
Rev. 46
27710A-C
25 VDC Breaker Preinstallation Check
Rev 7.1
27731-C
480 V Switchgear Cubicle/Transformer Maintenance
Rev. 41.1
27766-C
25 VDC Switchgear Cubicle Maintenance Inspect, Clean,
Rev. 18
28480-C
480V Emax Breaker Maintenance
Rev. 33.1
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
35363-C
Chemistry Control of the NSCW System
Rev. 9.3
83310-C
Emergency Diesel Generator Jacket Water Heat Exchanger
Testing
Rev. 8.2
DC-1009
Plant Single-Failure Criteria
Rev. 9
DC-1018
Pipe Break Criteria - Interdiscipline
Rev. 7
NMP-AD-012-
GL03
Immediate Determination of Operability Guideline
Rev. 3
NMP-AD-012-
GL04
Rev. 1.0
NMP-AP-003
Procedure and Work Instruction Use and Adherence
Rev. 6
NMP-ES-084-004
Equivalent Change Process
Rev. 1.4
NMP-ES-098
Interface Procedure for NISP-EN-02 Standard Item
Equivalency Process
Rev. 1.2
NMP-OS-014
Time Critical Operator Action Program
Rev. 2
Work Orders
SNC590686,
SNC395432,
SNC589750,
SNC594120,
SNC589725,
SNC589714,
SNC589717
,SNC366902,
SNC394772,
2091850401,
SNC831325,
SNC973848
,SNC135851,
SNC715030,
SNC371643,
SNC1015398,
SNC589713,
SNC968932,
SNC1020342,
Inspection
Procedure
Type
Designation
Description or Title
Revision or
Date
2060337004,
2060337025,
2060337026,
2060337021,
2060337028,
SNC130555,
2060337002,
2060337023,
2060337014,
SNC130544,
2060337015,
SNC139673,
SNC132183,
SNC132184,
SNC132185,
SNC132186,
SNC132192,
SNC132193,
SNC132194,
SNC132197,
SNC132275