IR 05000280/2017007
| ML17268A190 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 09/22/2017 |
| From: | Bartley J H NRC/RGN-II/DRS/EB1 |
| To: | Stoddard D G Virginia Electric & Power Co (VEPCO) |
| References | |
| IR 2017007 | |
| Download: ML17268A190 (18) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION REGION II 245 PEACHTREE CENTER AVENUE NE, SUITE 1200 ATLANTA, GEORGIA 30303-1257 September 22, 2017
Mr. Daniel Senior Vice President and Chief Nuclear Officer Innsbrook Technical Center 5000 Dominion Blvd.
Glen Allen, VA 23060-6711
SUBJECT: SURRY POWER STATION - NRC DESIGN BASES ASSURANCE INSPECTION (TEAM) REPORT NUMBER 05000280/2017007 AND 05000281/2017007
Dear Mr. Stoddard:
On September 1, 2017, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Surry Power Station, Units 1 and 2, and the NRC inspectors discussed the results of this inspection with Mr. Roy Simmons and other members of your staff. The results of this inspection are documented in the enclosed report.
NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding involved a violation of NRC requirem ents. The NRC is treating this violation as a non-cited violation (NCV) consistent with Section 2.3.2.a of the Enforcement Policy.
If you contest the violation or significance of the NCV, 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 the Surry Power 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 10 CFR 2.390, "Public Inspections, Exemptions, Requests for Withholding.
"
Sincerely,/RA/
Jonathan H. Bartley, Chief Engineering Branch 1
Division of Reactor Safety
Docket Nos.: 50-280, 50-281 License Nos.: DPR-32, DPR-37
Enclosure:
Inspection Report 05000280/2017007 and 05000281/2017007,
w/Attachment:
Supplementary Information
cc: Distribution via ListServ
_ _____________ SUNSI REVIEW COMPLETE FORM 665 ATTACHED OFFICE RI:DRS RII:DRS RII:DRS RII:DRS RII:DRS CONTRACTOR CONTRACTOR SIGNATURE Via Email Via Email Via Email Via Email Via Email Via Email Via Email NAME S.DOWNEY M.SCHWIEG R.PATTERSON G.OTTENBERGE.STAMM S.GARDNER W.SHERBIN DATE 9/19/17 9/18/17 9/21/17 9/18/17 9/21/17 9/18/17 9/20/17 E-MAIL COPY? YES NO YES NO YES NO YES NO YES NO YES NO YES NO OFFICE RII:DRP RII:DRS SIGNATURE Via Email JHB1 NAME A.MASTERS J.BARTLEY DATE 9/2/2017 9/22/2017 E-MAIL COPY? YES NO YES NO Enclosure U. S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket Nos.: 050000280, 05000281
Report Nos.: 05000280/2017007, 05000281/2017007
Licensee: Virginia Electric and Power Company (VEPCO)
Facility: Surry Power Station, Units 1 and 2
Location: 5850 Hog Island Road Surry, VA 23883
Dates: August 14 - September 1, 2017
Inspectors: E. Stamm, Senior Reactor Inspector (Lead)
G. Ottenberg, Senior Reactor Inspector S. Downey, Senior Reactor Inspector (Trainee) R. Patterson, Reactor Inspector M. Schwieg, Resident Inspector S. Gardner, Contractor
W. Sherbin, Contractor
Approved by: Jonathan H. Bartley, Chief Engineering Branch 1 Division of Reactor Safety
SUMMARY
IR 05000280/2017-007, 05000281/2017-007; 8/14/2017 - 9/1/2017; Surry Power Station, Units 1 and 2; Design Bases Assurance Inspection (Team).
The inspection activities described in this report were performed between August 14, 2017, and September 1, 2017, by five Nuclear Regulatory Commission (NRC) inspectors from Region II and two NRC contractors. The team identified one non-cited violation (NCV) of very low safety significance (Green). The significance of inspection findings are indicated by their color (i.e., greater than Green, or Green, White, Yellow, or Red) and determined using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," (SDP) dated April 29, 2015. Cross-cutting aspects are determined using IMC 0310, "Aspects Within the Cross-Cutting Areas," dated December 4, 2014. All violations of NRC requirements were dispositioned in accordance with the NRC's Enforcement Policy dated November 1, 2016. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 6.
NRC-Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green: The NRC identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to correctly evaluate the heat-up of the Main Steam Valve House (MSVH), which contains the auxilliary feedwater pumps as well as other safety-related mitigating systems. The violation was entered into the licensee's corrective action program as Condition Reports 1077007 and 1077684 and the licensee conducted a preliminary calculation and evaluation to determine the actual temperature increase and determined that the equipment located in MSVH remained operable.
The performance deficiency was determined to be 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 the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to evaluate worst-case design conditions resulted in a decreased margin for reliability and capability of mitigating systems contained in the MSVH. The inspectors determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance.. (Section 1R21.2.b.1)
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity
1R21 Design Bases Assurance Inspection (Team)
.1 Inspection Sample Selection Process
The team selected risk-significant samples and related operator actions for review using information contained in the licensee's probabilistic risk assessment. In general, this included risk significant structures, syst ems, and components (SSCs) that had a risk achievement worth factor greater than 1.3 or Birnbaum value greater than 1E-6. The sample included six components selected based on risk significance, one component associated with containment large early release frequency (LERF), five modifications to mitigatiing SSCs, and one operating experience (OE) item.
The team performed a margin assessment and a detailed review of the selected risk-significant components and associated operator actions to verify that the design bases had been correctly implemented and maintained. Where possible, this margin was determined by the review of the design basis and Updated Final Safety Analysis Report (UFSAR). This margin assessment also considered original design issues, margin reductions due to modifications, or margin reductions identified as a result of material condition issues. Equipment reliability issues were also considered in the selection of components for a detailed review. These reliability issues included items related to failed performance test results, significant corrective action, repeated maintenance, maintenance rule status, Inspection Manual Chapter 0326 conditions, NRC Resident Inspector input regarding problem equipment, system health reports, industry OE, and licensee problem equipment lists. Consideration was also given to the uniqueness and complexity of the design, OE, and the available defense-in-depth margins. An overall summary of the reviews performed and the specific inspection findings identified is included in the following sections of the report.
.2 Component and Modification Reviews
a. Inspection Scope
Components Selected Based on Risk Significance
- emergency switchgear room air handling units - 1/2-VS-AC-6/7
- turbine-driven auxiliary feedwater pumps - 1/2-FW-P-2
- emergency diesel generator number 3 fuel oil transfer system
- pressurizer power-operated relief valves - 1/2-RC-PCV-1/2455C, 1/2-RC-PCV-1/2456
- 4160V / 480V transformer - 2-EP-TRAN-2J-1
- emergency diesel generator number 2 output breaker to 2H emergency bus - 2-EP-BKR-25H3 Components with LERF Implications
- reactor trip breakers - 1/2-RP-BKR-52-RTA/RTB Modifications to Mitigation SSCs
- DCP 04-076 - Oversized Breakers and Undersized Cables Replacement
- DCP SU-15-01014 - Thermal Overload Repl acement for Safety Related Motor-Operated Valves
- DCP SU-15-01075 - Reconfiguring Reactor Protection Trip Breaker Ladder Logic Wiring
- DCP SU-16-00107 - Reconfigure Pressurizer Heaters For the seven components listed above, the team reviewed the plant technical specifications (TS), UFSAR, design bases documents, and drawings to establish an overall understanding of the design bases of the components. Design calculations and procedures were reviewed to verify that the design and licensing bases had been appropriately translated into these documents and that the most limiting parameters and equipment line-ups were used. Logic and wiring diagrams were also reviewed to verify that operation of electrical components conformed to design requirements. Test procedures and recent test results were reviewed against design bases documents to verify the adequacy of test methods and that acceptance criteria for tested parameters were supported by calculations or other engineering documents, and that individual tests and analyses served to validate component operation under accident conditions.
Maintenance procedures were reviewed to ensure components were appropriately included in the licensee's preventive maintenance program, that components or sub-components were being replaced before the end of their intended service life, and that the licensee has appropriate controls in place for components that are beyond vendor recommended life. Vendor documentation, system health reports, preventive and corrective maintenance history, and corrective action program documents were reviewed (as applicable) in order to verify that the performance capability of the component was not negatively impacted, and that potential degradation was monitored or prevented. Maintenance Rule information was reviewed to verify that the component was properly scoped, and that appropriate preventive maintenance was being performed to justify current Maintenance Rule status. Component walk downs (when accessible) and interviews were conducted to verify that the installed configurations would support their design and licensing bases functions under accident conditions, and had been maintained to be consistent with design assumptions.
For the five modifications listed above, the team reviewed design bases, licensing bases, and performance capability of components to ensure they had not been degraded through modifications. In addition, post-modification testing was reviewed to ensure operability was established by verifying unintended system interactions will not occur, SSC performance characteristics continue to meet the design bases, modification design assumptions are appropriate, and modification test acceptance criteria have been met. The team also verified design basis documentation was updated consistent with
the design change, verified other design basis features were not adversely impacted, verified procedures and training plans affected by the modification were updated, and verified that affected test documentation was updated or initiated as required by applicable test programs. Walk downs (when accessible) and interviews were conducted as necessary to verify that the modifications were adequately implemented. Documents reviewed are listed in the Attachment.
Additionally, the team performed the following specific reviews:
- The team reviewed completed time critical operator actions associated with tripping the reactor coolant pumps and aligning the auxiliary feedwater unit cross-tie for simulator scenarios involving loss of normal feedwater events to evaluate whether the required action times were supported by event analyses.
- The team observed a simulator scenario involving a loss of secondary side cooling requiring the alignment for bleed-and-feed core cooling in accordance with emergency operating procedures (EOPs) to evaluate the capabilities of the pressurizer power operated relief valves. The steps in the station EOPs were compared against the Westinghouse Owners Group Emergency Response Guidelines to evaluate the appropriateness of the expected actions relative to the pressurizer power operated relief valve capacity.
- The team walked down the actions required to manually trip the reactor trip breakers to evaluate the feasibility of accomplishing the action.
- The team observed various scenarios in the simulator involving the throttling of auxiliary feedwater flow to evaluate the challenge to the motor operated valves' motor duty cycle rating.
b. Findings
.1 Failure to Evaluate Design Maximum Ambient Temperature Effect on Main Steam Valve
House
Introduction:
The NRC identified a Green non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensee's failure to correctly evaluate the heat-up of the Main Steam Valve House (MSVH), which contains the auxilliary feedwater (AFW) pumps as well as other safety-related mitigating systems.
Description:
The MSVH is ventilated by natural circulation with air entering a motor-operated, normally-open, thermostatically-controlled damper at ground level and exiting through roof vents. There is also a non-safety related room exhaust fan, which is not credited for removing any heat generated from within. The team determined that since the outside air inlet damper was not safety-related or seismically qualified, nor were the thermostat or damper motor, it cannot be credited in a design basis accident, or seismic event. Failure of the damper to stay open would result in significantly less cooling air entering the MSVH.
MSVH ventilation calculation ME-0800, "MSVH Loss of Ventilation," Revision 0, Addendum A, had previously been performed to predict the room temperature with the damper closed, but was done with an outside ambient temperature of 70 degrees Fahrenheit (°F). It was noted that the room temperature near the AFW pumps approached 120°F in this analysis, which is the maximum allowed temperature described in the plant's Environmental Zone Description for this area. The team asked the licensee if the MSVH room temperature was evaluated with the summer outdoor design temperature of 95°F, with the damper closed, and was told there was no evaluation of this design basis condition. The licensee agreed that an evaluation was necessary at the higher temperature outside with the inlet damper failed closed.
Additionally, the team determined that calucation ME-0800 contained a non-conservative loss coefficient for the modeling of the roof vent. The roof vent contains missile barriers which cause a torturous path for air flow through the vent. The pressure loss coefficient in components depends primarily on the construction of the component and the impact the construction has on the fluid flow due to change in velocity and direction of flow. The higher the loss coefficient, the higher the pressure drop through the component, thus a lower flow rate. The inspectors determined that the loss coefficient value of 2.0 used in the calculation was non-conservative.
To evaluate these conditions, a preliminary calculation was performed to include a constant outside air temperature of 95°F, inlet damper closure, and loss of the forced ventilation at the start of the AFW pump 8-hour mission time. A conservative value of 5.0 was used for the loss coefficient of the roof vents. The preliminary results indicated a maximum temperature of 140°F on the 27' elevation, near the AFW pumps, and 160°F in the upper levels of the MSVH.
Dominion Design Engineering performed a preliminary review of the qualification documents for the components located in the MSVH, and concluded that the components would be expected to perform their function at 140°F on the 27' elevation, and 160°F in upper levels, for the duration of the calculated MSVH room temperature evaluation. The team reviewed the results of the evaluation and qualification documents and agreed with the licensee that the components would be expected to perform their
functions.
Analysis:
The team determined that the licensee's failure to evaluate design maximum outside ambient temperature when predicting room temperatures in the MSVH was a performance deficiency. The performance deficiency was determined to be 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 the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to evaluate worst-case design conditions resulted in a decreased margin which affected the reliability and capability of mitigating systems contained in the MSVH. The inspectors used IMC 0609, Att. 4, "Initial Characterization of Findings," issued December 7, 2016, for Mitigating Systems, and IMC 0609, App. A, "The Significance Determination Process (SDP) for Findings At-Power," issued June 19, 2012, and determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating SSC, and the SSC maintained its operability or functionality.
This finding was not assigned a cross-cutting aspect because the underlying cause was a legacy issue and not indicative of current performance.
Enforcement:
Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Calculation ME-0800, MSVH Loss of Ventilation, determines the maximum design basis temperatures experienced by mitigating systems contained in the MSVH. Contrary to the above, since original construction, the licensee failed to correctly translate the maximum design basis temperature into calculation ME-0800. Specifically, the licensee failed to evaluate the maximum MSVH room temperature and its potential effect on equipment operability at the higher temperature.
The licensee conducted a preliminary calculation and evaluation to determine the actual temperature increase and determined that the equipment located in MSVH remained operable. The violation was entered into the licensee's corrective action program as Condition Reports 1077007 and 1077684. This violation is being treated as an NCV consistent with section 2.3.2.a of the Enforcement Policy. (NCV 05000280, 281/2017007-01, Failure to Evaluate Design Maximum Ambient Temperature Effect on
Main Steam Valve House)
.3 Operating Experience
a. Inspection Scope
The team reviewed two operating experience items for applicability at the Surry Power Station. The team performed an independent review for these issues and, where applicable, assessed the licensee's evaluation and disposition of each item. The issues that received a detailed review by the team included:
- 10 CFR Part 21 Report Log Number 2013-09-01 - Wedge Pin Failure in Anchor Darling Motor Operated Double Disc Gate Valve with Threaded Stem to Upper Wedge Connections, dated July 11, 2017 (ADAMS Accession No. ML17194A825)
- OE20140 - Non-conservative Unverified Assumptions Used as Design Inputs for Calculating Thermal Overload (TOL) Sizes
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA6 Meetings, Including Exit
On September 1, 2017, the team presented the inspection results to Mr. Roy Simmons and other members of the licensee's staff. Proprietary information that was reviewed during the inspection was returned to the licensee or destroyed in accordance with prescribed controls.
ATTACHMENT:
SUPPLEMENTARY INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
- M. Antol, System Engineer
- C. Bruce, Supervisor, Engineering Coordination
- N. Dodenhoff, Supervisor, Engineering Design
- B. Garber, Manager, Station Licensing
- P. Hargrave, MOV Engineer
- L. Helstosky, Licensing Engineer
- J. Henderson, Director, Engineering
- R. Herbert, Manager, Design Engineering
- G. Hill, Unit Supervisor
- J. Holloway, Manager, Site Engineering
- R. Johnson, Manager, Operations
- J. LaFlam, System Engineer
- J. Lansing, System Engineer
- F. Mladen, Site Vice President
- J. Pelletier, System Engineer
- J. Rosenberger, Director, Nuclear Safety and Licensing
- J. Sears, Electrical Design Engineer
- R. Simmons, Plant Manager
- J. Stauffer, System Engineer
NRC personnel
- P. McKenna, Senior Resident Inspector, Surry
- C. Jones, Resident Inspector, Surry
- G. MacDonald, Senior Risk Analyst, Division of Reactor Projects
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened &
Closed
- 05000280, 281/2017007-01 NCV Failure to Evaluate Design Maximum Ambient Temperature Effect on Main Steam Valve
- House (Section 1R21.b.1)
LIST OF DOCUMENTS REVIEWED
Corrective Action Documents Written as a Result of the Inspection
- CR 1075852, Breaker Lifter Not Seismically Secured
- CR 1075974, Insulation Resistance Acceptance Criteria is Non-Conservative
- CR 1076279, Basis for Assumption in Ventilation Calculation is Not Stated
- CR 1076716, U2 MSVH Roof Block Ventilation Covers Eye Bolts Nuts Not Tight
- CR 1077007, Additional Calculation Results to Determine MSVH Temperatures
- SU-15-01014 Did Not Document Reduced TOL
- Margin
- CR 1077040, Evaluate Need for Continuing Training on Throttle MOVs
- CR 1077684, The Flow Coefficients for the MSVH Roof Vents in Two Calculations Are Different
Procedures
- 0-AP-13.02, Loss of ESGR Cooling, Rev. 9 0-AP-50.00, Opposite Unit Emergency (With 8 Attachments), Rev. 47 0-DRP-010, Air Operated Valve and Instrument Air Regulator Setpoints, Rev. 45
- 0-ECM-1201-02, Pressurizer Heater Maintenance, Rev. 18
- 0-EPM-2406-01, 4160/480 Volt Transformer Preventive Maintenance, Rev. 10
- 0-FCA-11.00, Remote Monitoring (with 7 attachments), Rev. 6
- 0-MCM-0701-11, Recirculation and Filtration of Emergency Diesel Underground Fuel Oil Tanks, Rev. 1 0-MCM-1935-02, Cleaning and Inspection of Fuel Oil Tanks, Rev. 2
- 0-MPM-0700-01, Emergency Diesel Generator Engine Eighteen Month Service and Inspection, Rev. 45 0-MPM-1900-02, Backflow Preventer, Rev. 17 0-NSP-CW-006, Canal Level Probe Removal, Inspection and Cleaning, performed 5/5/17 0-NSP-CW-006, Canal Level Probe Removal, Inspection and Cleaning, performed 6/6/17
- 0-NSP-CW-006, Canal Level Probe Removal, Inspection and Cleaning, performed 7/13/17
- 0-NSP-CW-006, Canal Level Probe Removal, Inspection and Cleaning, performed 7/28/17
- 0-NSP-VS-006, CR Envelope Air Conditioning Flow Measurement, performed 6/17/15
- 0-NSP-VS-006, CR Envelope Air Conditioning Flow Measurement, performed 8/18/15 0-OP-HS-001, Fuel Oil Storage Tanks, Rev. 25 0-OP-VS-006, Control Room and Relay Room Ventilation System, Rev. 72
- 0-OP-VS-006A, Control Room and Relay Room Ventilation System Alignment, Rev. 13 0-OPT-EG-005, 28 Day Freq. PT: No. 3 EDG Monthly FO Sys. Test -
- OC-22A, Rev. 20
- 0-OSP-HS-002, Static Test of Underground Fuel Oil Fill Piping, Rev. 8 0-OSP-TCA-001, Time Critical Action Validation and Verification, Revs. 12 and 13 0-VSP-M4, Flood Control Panel Trouble Alarm, Rev. 5
- 1-E-0, Reactor Trip or Safety Injection (with 10 attachments), Rev. 72
- 1-ECA_0.0, Loss of All AC Power (with 12 attachments), Rev. 42
- 1-ES-0.1, Reactor Trip Response (with 8 attachments), Rev. 53
- 1-FR-H.1, Response to Loss of Secondary Heat Sink (With 8 Attachments), Rev. 38 1-FR-S.1, Response to Nuclear Power Generation/ATWS (with 3 attachments), Rev. 27 1-IMP-C-IA-97,
- PCV-1455C b/u Bottled Air Low Pressure Check, performed 10/25/161-PT-8.1, Reactor Protection Logic (for normal operations), Rev. 42 1-OP-FW-001A, Auxiliary Feedwater System Valve Alignment, Rev. 7
- 1-OP-FW-002, Turbine Driven AFW Pump Startup and Shutdown, Rev. 23 1-OPT-EG-001, Emergency Diesel Generator Monthly Start Exercise Test, Rev. 69
- 1-OPT-FW-003, Operations Periodic Test, Turbine Driven AFW Pump 1-FW-P-2, Rev. 52 1-OPT-FW-003, 84 Day Frequency PT: Turbine Driven Aux. Feedwater Pump, performed
- 4/5/17 1-OPT-FW-021, 84 Day Frequency PT: Stroke Exercise Test of the AFW Crossover MOVs, performed 6/16/17 1-OPT-RC-001, Pressurizer PORV RFO Test, performed 11/2/16
- 1-PT-8.5, Consequence Limiting Safeguards Logic (Hi-Hi Train), Rev. 28 2-0PT-FW-006, Auxiliary Feedwater MOV Test, Rev. 11 2-DRP-007, Motor Operated Valve Operating Bands, Rev. 41
- 2-PT-8.2, Reactor Protection Logic, Rev. 26
- AD-AA-102, Procedure Use and Adherence, Rev. 11
- CM-AA-DDC-201, Design Changes, Rev. 20
- CM-AA-TCA-101, Operator Time Critical Actions, Rev. 1
- CY-AA-AUX-310, Diesel Fuel Oil Sampling and Testing, Rev. 9
- DNES-VA-EEN-0011, Standard for Protective Device Settings, Rev. 1
- ECM-0306-Electrical Corrective Maintenance, Rev. 24
- ECM-0309-01, Control Panel Maintenance, Rev. 8
- ER-AA-BPM-101, Underground Piping and Tank Integrity Program, Rev. 10
- OP-AP-104, Emergency and Abnormal Operating Procedures, Rev. 4
- NUS-2030, Specification for Electrical Installation, Revision 21
- VPAP-0505, Writers Guide for Dual-Column Procedures, Rev. 9
Drawings
- 113E244, Reactor Protection System Surry Power Station-Unit 1, Rev. 21
- 11448-FB-4A, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 19
- 11448-FB-4C, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 13
- 11448-FB-4D, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 2
- 11448-FB-4E, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 2 11448-FB-4F, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 1 11448-FB-4G, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 2
- 11448-FB-4H, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 1
- 11448-FB-4J, Yard Fuel oil Lines, Surry Power Station - Unit 1, Rev. 2
- 11448-FB-25D, Ventilation and Air Conditioning, Rev. 16
- 11448-FB-25E, Ventilation and Air Conditioning, Rev. 22 11448-FB-038A, Flow/Valve Operating Numbers Diagram, Fuel Oil Lines, Surry Power Station Unit 1, Sheet 1,
- Rev. 27 11448-FB-038A, Flow/Valve Operating Numbers Diagram, Fuel Oil Lines, Surry Power Station Unit 1, Sheet 2,
- Rev. 49 11448-FB-038A, Flow/Valve Operating Numbers Diagram, Fuel Oil Lines, Surry Power Station Unit 1, Sheet 3,
- Rev. 24 11448-FB-038A, Flow/Valve Operating Numbers Diagram, Fuel Oil Lines, Surry Power Station Unit 1, Sheet 4,
- Rev. 3 11448-FB-041A, Chill Water System, Rev. 72
- 11448-FB-041B, Chill Water System, Rev. 38
- 11448-FM-068A, Flow and Valve Operating Numbers Diagram, Feedwater System, Sheet 3, Rev. 69 11448-FM-075C, Flow and Valve Operating Numbers Diagram, Compressed Air System, Sheet
- 2, Rev. 40
- 11448-FM-086B, Flow and Valve Operating Number
s Diagram, Reactor Coolant System, Sheet
- 1, Rev. 33 11448-FV-36A, Underground Fuel Oil Storage Tanks, Rev. 4 11448-RE-25H, Vertical Control Boards, Rev. 22 11448-RE-25J, Vertical Control Boards, Rev. 9
- 11548-ESK-11AC, Bus H Degraded and Undervoltage Protection, Rev. 5
- 11548-FE-1A, Main One Line Diagram Surry Power Station, Rev. 25
- 11548-FE-1D, 4160V One Line Surry Power Station Unit 2, Rev. 19
- 11548-FE-1Q, 480V One Line Diagram Emergency Switchgear 2H1 & 2J1, Rev. 15 11548-FE-19AG, Elementary Diagram EDG Engine Auxiliaries, Rev. 21 11548-FE-21Q, DC Elementary Diagram 4160 Bus Bkr 25H3 & 25H8, Rev. 13
- 1501075-113E244A-B, Reactor Protection System Surry Power Station-Unit 2, Rev. 0
- 1501075-E-801-A, Reactor Protection Trip Breaker Ladder Logic, Rev. 0
- 1600107-11448-FE-1E, 480 V One Line Diagram, Surry Power Station Unit 1, Rev 0 1600107-11448-FE-1F, 480 V One Line Diagram, Surry Power Station Unit 1, Rev 0 D-338840, Pressurizer PORV Drawing, Series D Valve Assy. with Model 1000 Actuator, Rev.3
- S-95003-3-C-001, Site Plan & Details, Fuel Oil Line Replacement, Surry Power Station, Rev. 9
- S-95003-3-C-002, Base Line Yard Plan & Profile, Fuel Oil Lines Replacement, Surry Power Station, Rev. 5 S-95003-3-C-003, Base Line Yard Plan & Profile, Fuel Oil Lines Replacement, Surry Power Station, Rev. 6 S-95003-3-S-101, Plan & Sections - Fuel Oil Line Missile Shields, Fuel Oil Line Replacement, Surry Power Station Units 1 & 2, Rev. 3
Calculations
- 85-046B, Evaluation of Installing Noise Dampening Capacitor in
- FT-1433 Loop, dated 2/20/85
- 7797.04-E-001, Effect of Maximum Fault Current on Cable from Safety Related 4KV, Rev. 0
- 010139.3410-M-006, Fuel Oil Pumping Requirements from the Underground Storage Tanks to the Day Tanks, Rev. 0 14257.88-S-001, Design of MSVH Roof Plugs, Rev. 2 52308.04-C-013, Reevaluation of the Over-turning Capacity of the Surry ECSTs, Rev. 0
- CE-1210, Pipe Stress Analysis Of 2.5 Inch Fuel Oil Lines Buried in the Yard Between Fuel Oil Pump House and Service Building, Rev. 0
- CE-1228, Evaluation of Buried EDG Fuel Oil Lines Structural Response Under the Cask Transporter Loading, Rev. 0, Addendum 00B
- DEO-0096, PORV Redundant Bottle Air System Capacity, Rev. 0
- DEO-0598, Underground Fuel Oil Storage Tank Pressure Requirements, Rev. 1
- EE-0029, Station Electrical Load List, Rev. 2
- EE-0034, Surry Voltage Profiles, Rev. 3
- EE-0306, Evaluation of MOV Thermal Overload Settings, Rev. 3
- EE-0306, Evaluation of MOV Thermal Overload Settings, Rev. 3, Add. A
- EE-0336, Relay Setting of Feeder Breakers on Buses 2H and 2J, Rev. 0
- EE-0340, Relay Setting Calculations for the Protection of Bus 2J, Rev. 0
- EE-0497,
- SR 480V Load Center Coordination, Rev 2
- EE-0497,
- SR 480V Load Center Coordination, Rev 2, Add. 2
- EE-0499, DC Vital Bus Short Circuit Current, Rev. 2 Add B
- EE-0501, Motor Terminal Voltage for Motor-Operated Valves, Rev. 1
- EE-0501, Motor Terminal Voltage for Motor-Operated Valves, Rev. 1, Add. AE
- EE-0879,
- SPS 4160V and 480V Short-Circuit Analysis, Rev. 0
- ETE-CEE-2014-1008, Scaling and Recommended Setpoints for the Model FLT93S Canal Level Probes for Surry Power Station, Rev. 1
- ETE-SU-2015-0083, Bio-fouling Contingency Factor in the canal Level Probe Response Time Calculations, Rev. 0 M-4, Emergency CST Volume and Design Basis Check and Process Instrument Setpoint Location, Rev. 0 M-045, Back-up Nitrogen Gas Supply for Turbine Driven AFW Pump AOVs, Rev. 0
- ME-0318, Canal Level Probe Response Time, Revs. 0-2
- ME-0561, Emergency Diesel Day Tank 1 Hour Fuel Oil Requirements, Rev. 0
- ME-0597, Minimum Delivered AFW Flow and Acceptance Criteria for AFW Pump Operability Verification Testing, Rev. 4 and Associated Addenda
- ME-0602, Maximum AFW Flow to the Steam Generator s during a MSLB Event and AFW Pump
- ME-0603, EDG 7 Day Fuel Oil Requirements, Rev. 0
- ME-0631, System Level Calculation for Air Operated Valves, Rev. 0
- ME-0800, GOTHIC MSVH Loss of Ventilation, Rev. 0 and Add. A
- ME-0871, Loss of Ventilation in the Fuel Oil Pump House, Rev. 0
- ME-0914, Vacuum Pressure Margin for the Fuel Oil Transfer Pump, Rev. 0
- ME-0931, Minimum Air Flowrates for MCR and ESGR Air Handlers, Rev. 0 NAI- 1232-03, GOTHIC Model for Surry MSVH, Rev. 0
- SEO-1298, Diesel Fuel Oil Tanks Volume Table Calculation, Rev. 0
- SM-945, Revised Heat-up and Cooldown Curves and LTOPS Setpoint for Surry Units 1 and 2, Rev. 0
- SM-1002, Surry PORV Cycling Study, Rev. 1
- SM-1123, Complete Loss of ESGR Ventilation, Rev. 0
- SM-1123C,
- Loss of Chillers Event, Rev. 0
- SM-1149, Surry Loss of Feedwater Due to High Energy Line Break- Reduced SAL for SG Level, Addendum B, Rev. 1
- SM-1431, ATWS Critical Power Trajectories for Surry Power Station, Rev. 0
- SM-1612, Surry ECST Volume Sizing for Design Basis Analysis, Rev. 0
- ZZZ-01039. 0810-M-4, Distribution Control Room Envelope, Rev. 0
- Corrective Action Documents
- CA 305101
- CA 3048346
- CA 3048347
- CA 3052804
- CR 37581
- CR 119541
- CR 247650
- CR 398628
- CR 541177
- CR 556254
- CR 558445
- CR 559872
- CR 559875
- CR 560488
- CR 560598
- CR 561817
- CR 561820
- CR 561995
- CR 562327
- CR 564698
- CR 565104
- CR 565188
- CR 569764
- CR 580467
- CR 580471
- CR 583168
- CR 1000557
- CR 1010756
- CR 1018275
- CR 1042126
- CR 1042133
- CR 1045354
- CR 1049951
- CR 1052133
- CR 1052211
- CR 1052268
- CR 1058209
- CR 1058210
- CR 1068858
- CR 1069087
- CR 1071016
- CR 1075726
- CR 1075739
- CR 1075742
- CR 1075744
- CR 1075747
- CR 1075748
- CR 1075749
- CR 1075750
- CR 1075751
- CR 1075752
- CR 1075753
- CR 1075754
- CR 1075755
- CR 1075757
- CR 1075758
- CR 1075761
- CR 1075762
- CR 1075763
- CR 1075764
- CR 1075766
- CR 1075767
- CR 1075768
- CR 1075769
- CR 1075771
- S-2006-0814
Work Orders
- 0077654101
- 38013766808
- 38047961501
- 38047968201
- 38073631012
- 38077654101
- 38102326076
- 38102624398
- 38102755794
- 38102764999
- 38102765015
- 38102765019
- 38102777836
- 38102778506
- 38102778507
- 38102948149
- 38103030140
- 38103262896
- 38103294407
- 38103368197
- 38103368198
- 38103368201
- 38103390153
- 38103403942
- 38103454875
- 38103533934
- 38103559165
- 38103580674
- 38103584844
- 38103601786
- 38103601789
- 38103629816
- 38103634886
- 38103639754
- 38103639759
- 38103672278
- 38103673668
- 38103685915
- 38103690856
- 38103693684
- 38103702119
- 38103723788
- 38103739680
- 38103740349
- 38103755804
- 38103778492
- 38103784535
- 38103789566
- 38103789865
- 38103790200
- 38103790989
- 38103797590
- 38103797591
Miscellaneous Documents
- 11448-CG-15, Seismic Pump List, Rev. 0 38-B851-00008, Installation and Maintenance instruction indoor Dry-Type Transformer, Rev. 3
- 38-L553-00001, Limitorque Tyoe SMB Instruction and Maintenance Manual, Rev. 19
- ACE 018398, 2-CS-MR-1B Fire Damage
- ACN 7306920612, Effects of Piping System Breaks Outside Containment, dated 6/22/73
- Component Health Report, Breakers, Q2 2017
- DC 89-14-3, Replacement of MSVH Roof Plug Covers Surry Units 1 &2, Rev. 1
- DC 95-003, Fuel Oil Line Replacement
- DCP 04-076, Oversized Breakers and Undersized Cables Replacement, Rev. 0
- SU-15-01014, Thermal Overload Replacement SR MOVs, Rev. 0
- DNES-AA-GN-1002, Document Impact Summary, Rev. 15
- DNES-AA-GN-1003, Design Effects and Considerations, Rev. 18
- DNES-VA-EEN-0002, Design Standard for Cable, Rev. 0
- DNV-NP-1999, Emergency Communications Review, dated 4/2/90
- Engineering Information Bulletin, Failure to Perform Periodic Testing for Safety Related Breakers, dated 3/10/15
- ESGR Air Handler Maintenance Rule Scoping Document
- ET-N-07-0049, Revision to No Anna TS to Eliminate Periodic Pressure Sensor Response TimeTests, Rev. 1
- ET-NAF-09-0012, Surry Measurement Uncertainty Recapture (MUR) Project: Evaluation of
- ETE-NAF-2014-0026,
- NE-1200, Rev. 12 Impact to Surry and North Anna Time Critical Operator Action Program, Rev. 0 ET S-01-0027, Determine Acceptance Criteria for the Six AFW Pumps to Ensure that Both the IST Program and the Design Basis Requirements are Satisfied, dated 3/8/01 ET S-03-0160, IST Acceptance Criteria for 1-FW-P-2 Following Pump Replacement, Rev. 0
- EWR 89-268, Diesel Fuel Oil Biofouling Support IEEE 43-2000, Recommended Practice for Testing Insulation Resistance
- IEEE 279, Nuclear Power Plant Protection Systems, 1968
- IEEE 308, Class 1E Electric Systems for Nuclear Power Generating Stations, 1970
- IEEE 323, Qualifying Class 1E Equipment for Nuclear Power Generating Stations, 1974 IEEE 338, Periodic Testing of Nuclear Power Generating Station Protection Systems, 1971
- IEEE 344, Seismic Qualification of Class 1E E
quipment for Nuclear G
enerating Stations, 1975
- K0932-09, ESGR AHU Vendor Manual, Rev. 0 Lawrence Livermore Labs, Technical Evaluation of Susceptible Safety Related Systems to Flooding caused by the Failure of Non-Category I Systems, December 1980 Letter from R. V. Green to R. H. Blount, Status of the Evaluation of 2-VS-D-41, dated 3/19/93
- Memorandum, Nuclear Safety Analysis Input on Surry EOP Engineering Issues, dated 10/25/89
- Memorandum, Plant Specific Technical Review- EOP's Surry Power Station- PES No.
- NRC Letter, VEPCO Serial Number 93-382, Closeout of NRC Bulletin 88-04, Safety Related Pump Loss, AFW Pumps, dated 6/8/93
- NE-1050, Technical and Operational Basis for Pressurizer PORV Air Bottles, May 1996
- NE-1200, Key Operator Actions Assumed in the Safety Analysis, Rev. 15
- NS-TMA-2182, ATWS Submittal, dated 12/30/79
- NUS-0196, Specification for Fuel Oil Pumps for Surry Power Station, Rev. 2
- QDR-N-3.1/QDR-S-3.1, Qualification Document, MOVs Outside Containment, Rev. 27
- QDR-N-9.1/QDR-S-9.1, Qualification Document, Limit Switches, Rev. 28 Quick Cause
- 3040587, Inadvertent opening of the Unit 1 A-train Reactor Trip Bypass Breaker during the performance of surveillance test 1-PT-8.1 on 8/23/2016.
- RCE 001128, Surry Unit 2 Trip, October 13, 2014
- Report of Analysis 2016-NFLK-000368-003, Diesel Fuel representing 1-EE-TK-2B
- Report of Analysis 2016-NFLK-000615-003, Diesel Fuel representing 1-EE-TK-2B
- Report of Analysis 2016-NFLK-000807-006, Diesel Fuel representing 1-EE-TK-2B Report of Analysis 2017-NFLK-000235-003, Diesel Fuel representing 1-EE-TK-2B Report of Analysis 2017-NFLK-000430-001, Diesel Fuel representing 1-EE-TK-2A
- Report of Analysis 2017-NFLK-000430-002, Diesel Fuel representing 1-EE-TK-2B
- RTE-342361, Unit 1 Recurring Task Evaluation for testing of 125VDC MCCB, dated 5/7/15
- RTE-342362, Unit 2 Recurring Task Evaluation for testing of 125VDC MCCB, dated 5/7/15
- SDBD-SPS-EP, SDD for Emergency Power System, Rev. 18
- SDBD-SPS-FW, System Design Documents for Feedwater System, Rev. 16
- SDBD-SPS-HC, CR Ventilation System, Rev. 15
- SDBD-SPS-RPS, Design Basis Reactor Protection System, Rev. 17
- Simulator Training, Course #:
- RQ-17.5-ST-2, performed 8/31/17
- Simulator Training, Course #:
- RQ-17.5-ST-3, performed 8/16/17 SPS
- FR-H.1, Response to Loss of Secondary Heat Sink, dated 9/3/10
- SU-09-00078, Startup on AFW & Carbohydrazide Usage Evaluation / Unit 2, dated 11/18/09
- SUI-0018, Specification for Excavation, Fill Placement, Compaction, and Testing Fuel Oil Lines Replacement Surry Power Station Units 1 and 2 Surry Environmental Zone Descriptions, Units 1 and 2, Rev. 28 Surry Power Station Unit 1 Inservice Testing Program Plan for Pumps and Valves, Fifth Testing Interval, May 10, 2014 to May 09, 2024, Rev. 1 System Health Reports, 480VAC, Various System Health Reports, 4kV Transformer, Various System Health Report, >4160VAC, Q4 2016
- System Health Report, Feedwater, Q1 2017 System Health Report, Reactor Coolant, Q2 2017 System Health Report, Reactor Protection, Q2 2017
- TP16-1-112, Recommendations to Resolve Flowserve 10CFR Part 21 Notification Affecting Anchor Darling Double Disc Gate Valve Wedge Pin Failures, Rev. 3
- VEP-FRD-41-P-A, VEPCO Reactor System Transient Analyses Using the RETRAN Computer Code, Rev. 0.2 Vendor Tech Manual 38-C-432-00029, PORV Model D-1000 Upgrade Instructions, Rev. 1 Vendor Tech Manual 38-E035-00001, Operating Manual 999 System Generating Plant, Rev. 51 Vendor Tech Manual 38-F-264-00005, Type 667 Diaphragm Actuator, Rev. 15
- Vendor Tech Manual 38-S902-00001, Installation and Operating Instructions for Sier-Bath Hydrex II Pump, Rev. 10 Vendor Tech Manual 38-W893-00033, Instructions for Pressurizer for Virginia Electric and Power Company Surry power Station Unit No.1, Rev. 4
- Vendor Tech Manual
- SU-VTM-000-38-T291-00001, AFW Turbine Manual, Rev. 14 Vendor Tech Manual
- SU-VTM-38-A-499-00011, High Flow 4-Way Solenoid Valves, Rev. 4
- VTM, Westinghouse Maintenance Program Manual for Safety Related Type DB Low Voltage Metal Enclosed Switchgear, Rev. 4
- WCAP-8373, Qualification of Westinghouse Seismic Testing Procedure for Electrical Equipment Tested Prior to May 1974, August 1974
- WCAP-14036-P-A, Elimination of Periodic Protection Channel Response Time Tests, Rev. 1
- Westinghouse Owners Group Emergency Response Guidelines, Background Volume E-0,
- ECA-0, High Pressure Version, Rev. 2 Westinghouse Owners Group Emergency Response Guidelines, Background Volume
- FR-S,
- FR-C,
- FR-H, High Pressure Version, Rev. 2 Westinghouse
- TB-14-2, Aging Issues and Subsequent Operating Issues for Molded Case Circuit Breakers That Have Reached 20 Year Design/Qualified Lives, Rev. 0 WOG Report, Maintenance Program for
- DB-50 Reactor Trip Switchgear, Rev. 0