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#REDIRECT [[RC-16-0035, (VCSNS) Unit 1 - Exigent License Amendment Request - LAR (16-00848) Technical Specification Change Request for the Emergency Feedwater System Limiting Condition for Operation 3.7.1.2 Action B]]
| number = ML16062A368
| issue date = 03/01/2016
| title = Virgil C. Summer Nuclear Station (VCSNS) Unit 1 - Exigent License Amendment Request - LAR (16-00848) Technical Specification Change Request for the Emergency Feedwater System Limiting Condition for Operation 3.7.1.2 Action B
| author name = Lippard G A
| author affiliation = South Carolina Electric & Gas Co
| addressee name =
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000395
| license number = NPF-012
| contact person =
| case reference number = CR-16-00848, RC-16-0035
| document type = Letter, License-Operating (New/Renewal/Amendments) DKT 50, Technical Specification, Bases Change
| page count = 21
}}
 
=Text=
{{#Wiki_filter:George A. LippardVice President, Nuclear Operations803.345.48101, 2016A SCANA COMPANY RC-1 6-0035Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555
 
==Dear Sir I Madam:==
 
==Subject:==
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1DOCKET NO. 50-395OPERATING LICENSE NO. NPF-12EXIGENT LICENSE AMENDMENT REQUEST -LAR (1 6-00848)TECHNICAL SPECIFICATION CHANGE REQUEST FOR THE EMERGENCYFEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2ACTION bSouth Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for SouthCarolina Public Service Authority pursuant to 10 CFR 50.90 and 10 CFR 50.91, hereby submitsa request for an exigent amendment to Technical Specifications (TS). The proposedamendment would modify the action statement for two inoperable pumps or flow paths withinSection 3.7.1.2, "Plant Systems -Emergency Feedwater System."Attachment I provides an evaluation of the proposed change to the action statement to amendthe six hour action to be in at least HOT STANDBY to 24 hours to allow for maintenance andretesting. This amendment request was evaluated and found to have no significant hazards forconsideration. An exigent TS change is justified in that compliance with TS could involve anunnecessary plant action to shutdown the reactor to COLD SHUTDOWN and potential relianceon the turbine driven emergency feedwater pump for plant cooldown without a correspondinghealth and safety benefit. The station proposes that the action statement be amended to 24hours to allow for maintenance and retesting. Attachment 2 contains the marked-up version ofthe affected TS page. Attachment 3 contains the reprinted versions of the affected TS page.In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being providedto the designated South Carolina Official. This proposed change has been reviewed andapproved by both the VCSNS Plant Safety Review Committee and the VCSNS Nuclear SafetyReview Committee.SCE&G requests approval of the proposed amendment by March 10, 2016. Once approved,the amendment shall be implemented immediately.The proposed change does introduce one new commitment. If you have any questions orrequire additional information, please contact Bruce Thompson at (803) 931-5042.V. C. Summer Nuclear Station .P. O. Box 88
* Jenkinsville, SC. 29065.* F (803) 941-9776 Document Control DeskRC-1 6-0 035CR-I16-00848Page 2 of 2I certify under penalty of perjury that the information contained herein is true and correct.Executed onG ieo e A i~rWLT/GAL/Attachments:1. Analysis of Proposed Technical Specification Change2. Proposed Changes -Marked Up TS Page3. Proposed TS Pages -Retyped4. Commitment Pagec: K. B. MarshS. A. ByrneJ. B. ArchieN. S. CamnsJ. H. HamiltonJ. W. WilliamsW. M. CherryC. HaneyS. A. WilliamsNRC Resident InspectorK. M. SuttonP. LedbetterS. E. JenkinsNSRCRTS (CR-I16-00848)File (813.20)PRSF (RC-16-0035)
Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 1 of 14VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)DOCKET NO. 50-395OPERATING LICENSE NO. NPF-12Attachment 1*Analysis of Proposed Technical Specification Change
 
==Subject:==
This evaluation supports a request to amend South Carolina Electric & GasCompany (SCE&G), Technical Specifications (TS) to modify the action statementof 3.7.1.2, Emergency Feedwater System, Limiting Conditions For Operation(LCO) for two inoperable motor driven pumps.1.0 SUMMARY DESCRIPTIONIn accordance with the provisions of 10 CFR 50.90, South Carolina Electric & Gas Company,acting for itself and as agent for South Carolina Public Service Authority, requests NuclearRegulatory Commission (NRC) review and approval to amend Operating License NPF-12 forVirgil C. Summer Nuclear Station (VCSNS) Unit 1.VCSNS is proposing an exigent TS change. It is the station's position that compliance with TScould involve an unnecessary plant shutdown and the potential reliance on the turbine drivenemergency feedwater pump (TDEFP) for plant shutdown without a corresponding health andsafety benefit. Due to an oversight, the station missed a surveillance test during the fall 2015startup from refueling outage 22 (RF-22) associated with the emergency feedwater (EF) controlvalves in accordance within VCSNS Technical Specification 4.7.1.2.c.2. This surveillancerequires verifying the flow control valves can be closed and held closed for three hours whennormal instrument air is not available. The surveillance is normally conducted in Mode 4 orbelow when the Steam Generators are not relied on for heat removal. Due to the designconfiguration of the EF system, the six hour action statement b for two inoperable emergencyfeedwater pumps is entered anytime a motor driven emergency feedwater pump (MDEFP) flowcontrol valve is closed in modes 1, 2 or 3. The station proposes to modify limiting conditions foroperation 3.7.1.2 action statement b which currently requires: for two inoperable emergencyfeedwater pumps, be in at least HOT STANDBY within six hours and be in HOT SHUTDOWNwithin the following six hours. The station proposes that the action statement be amended to bein at least HOT STANDBY within 24 hours to allow for timely completion of any requiredmaintenance and surveillance retest.
Document Control DeskAttachment 1RC-I16-0035CR-I16-00848Page 2 of 142.0 DETAILED DESCRIPTIONDue to an oversight the station has missed performing a surveillance associated with the EFcontrol valves as reflected within TS 4.7.1 .2.c.2 during the startup from refueling outage 22.The surveillance requirement is for at least once per 18 months during shutdown and is typicallycompleted in HOT SHUTDOWN or below when the steam generators are not relied on for heatremoval. This test requires the MDEFP flow control valves be held closed for 3 hours with airfrom the accumulators. Due to the design configuration of the EF system, the six hour actionstatement b for two inoperable EF pumps is entered anytime a MDEFP flow control valve isclosed in modes 1, 2 or 3. With the test time period of three hours, no time is available toconduct remedial corrective maintenance and repeat the surveillance. This could result in anunnecessary plant shutdown.The end date for this test is March 17, 2016, based on an 18 month surveillance interval plus25% per TS 4.0.2. During performance of the General Operating Procedure (GOP-2) for PlantStartup and Heatup (MODE 5 to MODE 3), the surveillance was thought to be complete basedon completion of surveillance for the TDEFP flow control valves. This error was not detecteduntil the plant was in HOT SHUTDOWN at approximately 345 degrees Fahrenheit and wasrelying on MDEFP flow for heat removal as is normal for the start-up process.The station has prepared to conduct the test during Model by entering the TS 3.7.1.2 actionstatement b to conduct the test. However, conducting the 3 hour surveillance test at powerwhile in a 6 hour shutdown action statement leaves no time to make repairs and conduct aretest. Changing the action statement to 24 hours will allow for unforeseen correctivemaintenance and subsequent retest would prevent the station from an unnecessary plantshutdown without a corresponding health and safety benefit.2.1 Possible Repairs TimelineThe following potential component failures could be required following the surveillance. Theestimated repair times are based on repairing each item identified below and include tagging outthe appropriate isolation devices. The time reflected also accounts for retesting of the EFcontrol valves to ensure the capability to hold the valve closed for three hours as required by TS4.7.1 .2.c.2.Air accumulator check valve replacement -18 hours.Air pressure regulator rebuild and calibration -10 hours.Air actuator diaphragm casing bolts torque adjustment -8 hours.Air actuator diaphragm replacement -12 hours.Air solenoid valve replacement -14 hours.Air relief valve replacement and setup -8 hours.
Document Control DeskAttachment 1RC-1 6-0035CR-i16-00848Page 3 of 142.2 Bases for Exigent ChangeSurveillance Test Procedure (STP)-1 20.006, "Emergency Feedwater Valves Backup Air SupplyTest," was not performed during the fall 2015 outage for the MDEFP flow control valves. Theend date for this test is March 17, 2016, based on an 18 month surveillance interval plus 25%per TS 4.0.2.During performance of GOP-2, "Plant Startup And Heatup (MODE 5 to MODE 3)," thissurveillance was signed off as being complete based on completion of the A-train portion of thetest, which tests the TDEFP flow control valves, done under a separate task sheet from the onewritten for the B-train valves (MDEFP flow control valves). This error was not detected untilperformance of General Testing Procedure (GTP-702), "Surveillance Activity Tracking andTriggering," for Mode 3 entry, which lists the A-train and B-train tasks as separate line items. Bythat time, the plant was relying on MDEFP flow and the steam generators for heat removal as isnormal during start-up. To perform the testing at this point in the outage would require the plantto cool down to less than 183 degrees Fahrenheit and reinitiate Residual Heat Removal (RHR)cooling. The precautions in STP-120.006 showed the procedure allows the subject testing inModes 1, 2, and 3 as long as both emergency diesel generators are operable with nomaintenance or testing in progress on either emergency diesel generator.Because the procedure allowed testing in Mode 1, 2, or 3 the decision was made to not cool theplant back down to less than 183 degrees Fahrenheit and reinitiate RHR cooling, but instead toschedule the performance of the required testing in Mode 1 once the plant reached a 100%power.The surveillance test was placed in the plant online work week schedule to be performed onFebruary 26, 2016. During a normal process schedule review on January 30, 2016, it wasdiscussed that this test would need additional focus to be performed online due to the shortduration six hour action to HOT STANDBY required by TS 3.7.1.2 action b. Station personnelthen began to apply additional planning considerations and focus to the testing includingdesignating the test as an Infrequently Performed Test or Evolution (IPTE) and developingcontingencies for repairs should valve repairs be required. After input by several plant groupsthe contingency matrix was finalized late on February 23, 2016. Based on the estimated timesfor repairs and retesting in the matrix, it was determined that a reasonable repair could not beapplied within the actions specified in TS 3.7.1.2 action b. The test was rescheduled to beperformed on March 11, 2016, to allow additional planning time. While the actual end date forsurveillance is March 17, 2016, major maintenance has been scheduled on one of theemergency diesel generators for the week of March 13, 2016.
Document Control DeskAttachment IRC-1 6-0035CR-16-00848Page 4 of 14While the station has prepared to conduct the test during Mode 1 by entering the TS 3.7.1.2action statement b to conduct the test, conducting the 3 hour surveillance test at power while ina 6 hour shutdown action statement leaves no time to make repairs and conduct a retest.Changing the action statement to allow 24 hours to allow for unforeseen corrective maintenanceand subsequent retest would prevent the station from an unnecessary plant shutdown without acorresponding health and safety benefit.
 
==3.0 TECHNICAL EVALUATION==
The EF system consists of three pumps, two motor driven and one steam turbine driven. TheEF System is used to supply feedwater to the steam generators during startup, shutdown, andlayup operations. A simplified system drawing is shown as Figure 1 where the full version canbe viewed within the FSAR Figure 10.4-16 or VCSNS drawing 302-085.The EF flow control valves fail open which is the "safe" position for most accidents but are alsorequired to be closed due to a faulted Steam Generator. The flow control valves are held closedfor three hours by their associated air accumulators following an EF high-flow signal to a faultedSteam Generator. The three hour time permits automatic valve closure following a secondarysystem break when local valve operation cannot be accomplished because local conditions areunsuitable for personnel access. The valves are supported by safety class air accumulatorswith sufficient capacity to permit remote valve closure for at least three hours during a loss ofinstrument air system. The air accumulators provide a regulated air supply as needed to closethe valves against spring force. The accumulators are supported by a non-safety instrument airsystem. Additionally, each valve has a handwheel to provide manual control.3.1 System OperationThe EF system "is required to deliver sufficient feedwater to the Steam Generators for cooldownupon loss of the normal feedwater supply and during an Anticipated Transient Without Scram(ATVVS) event. The EF system is used to supply feedwater to the Steam Generators duringstartup, shutdown, and iayup operations. The EF system operates in conjunction with theturbine bypass system, if available, or the main steam power relief valves and safety valves, toremove thermal energy from the Steam Generators.The system is designed to automatically deliver feedwater, at a minimum total flow of 380 gpm,to at least two Steam Generators pressurized to 1211 psig. There is sufficient redundancy toestablish this flow while sustaining a single active failure in the system in the short term or asingle active or passive failure in the long term. The EF system operates until the RHR Systemcan be placed in operation.
Document Control DeskAttachment 1RC-l16-0035CR-I16-00848Page 5 of 14When forced circulation from the reactor coolant pumps is not available, EF operation isrequired down to a main steam pressure of 100 psia. This corresponds to a reactor coolant coldleg temperature of 325 degrees Fahrenheit and a hot leg temperature of 350 degreesFahrenheit. The primary coolant temperature differential is required in order to maintain adensity gradient to drive natural circulation of primary coolant, in the absence of reactor coolantpump operation.Sufficient feedwater is available under emergency conditions to bring the plant to a safeshutdown condition. Assuming prior plant operation at engineered safety design rating (ESDR)of 2900 MWt in the core, the minimum required usable volume for the condensate storage tankis 158,570 gallons based on maintaining the plant at hot standby conditions for 11 hours.This volume also satisfies the minimum required volume to cool down the plant to HOTSHUTDOWN conditions assuming the plant is maintained at HOT STANDBY for 2 hours andthen cooled down to HOT SHUTDOWN in 4 hours.The system consists of three pumps, two motor driven and one steam turbine driven. The twomotor driven pumps share a common discharge header that splits off into three branches. Eachbranch has a pneumatic flow control valve, which controls flow to its respective steamgenerator. The one steam turbine driven pump has a separate header from the motor drivenpumps: This header splits off into three branches, which controls flow to its respective SteamGenerator.The three flow control valves for the turbine driven pump have control elements fed from A-trainpower. The three flow control valves for the motor driven pumps have control elements fed fromB-train power.During the performance of testing of the B-train flow control valves, the flow path from both ofthe motor driven pumps is disrupted. This is due to both pumps sharing a common dischargeheader prior to branching off to the three flow control valves.The MDEFPs are powered from separate and independent safety related emergency dieselgenerator backed buses.EF is a dual purpose system. During normal operation, the motor driven pumps are used duringheatup and cooldown to supply feedwater to the steam generators for reactor coolant systemtemperature control. During emergency operation, all three pumps can provide feedwater tosupport reactor coolant system heat sink capabilities via the Steam Generators.
Document Control DeskAttachment 1RC-1 6-0035CR-i16-00848Page 6 of 14EMERGENCY FEEDWATER SYSTEMFigure 1 Simplified System Drawing3.2 Component DesignThe EF System design consists of two redundant trains: i.e., the motor driven pump train andthe turbine driven pump train. The motor driven pump train is designed for use during normalplant conditions (i.e. startup, hot standby, and cooldown) and for emergency shutdown of thereactor. The turbine driven pump train is designed for use for emergency shutdown of thereactor. EF initiation arises from any of several types of signals, which may be generated inresponse to a variety of plant conditions. These initiation signals are generated in response tolow Steam Generator levels, loss of main feedwater, low voltage on the essential electric powerbuses, a Safety Injection signal, and an ATWVS mitigation signal.The control systems for the turbine driven and motor driven pump flow control valves areidentical. Automatic valve opening signals are generated by the reactor protection and logicsystem and depend upon the given plant condition which will determine whether only the motordriven pump flow control valves open or the turbine driven pump valves open also. Thesevalves will receive an open signal whenever their respective pumps receive an auto-start. Theexception is the MDEFP flow control valves which do not receive an open signal when all threemain feed pumps trip. If the flow control valves are in MANUAL control, the valves fully open inresponse to an automatic open signal.
Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 7 of 14The EF system is designed with three flow control valves at the discharge of the TDEFP andthree flow control valves at the discharges of the MDEFPs. The valves are required to controlEF flow to the Steam Generators to maintain program level and to produce sufficient mainsteam to permit main feedwater pump turbine operation and for plant cooldown after mainsteam is no longer able to drive the main feedwater pump turbines. All six valves are identical,3-inch Fisher ET, normally open, air-operated, globe valves. The valves are safety relateddevices which meet the requirements of ASME B&PV Code Section III, class 2, 1974 edition,Summer, 1975 addenda.The valves fail open which is the "safe" position for most accidents. The valves are supportedby safety-class air accumulators with sufficient capacity to permit remote valve closure on ahigh-flow signal and maintain the valve closed for at least three hours during a loss ofinstrument air system. The air accumulators provide a regulated air supply as needed to closethe valves against spring force. The accumulators are supported by a non-safety instrument airsystem. Additionally each valve has a handwheel to allow manual control.The primary safety function of the air accumulator is to assure a source of safety related air isavailable to isolate the flow control valve to a faulted Steam Generator. The three hour supplypermits automatic or remote manual valve closure following a secondary system break whenlocal valve operation cannot be accomplished due to unsuitable conditions for personnel accessin the Intermediate Building. Figure 2 is provided as a simplified sketch of the control airsystem.b .......Figure 2 Simplified Control Air System (VCSNS Drawing 817-056-001)
Document Control DeskAttachment IRC-1 6-0035CR-I16-00848Page 8 of 143.3 Component HistoryThe MDEFP flow control valves are subject to periodic stroke time testing to the open andclosed positions under TS 4.0.5. Portions of the test circuit in the pressure drop tested underTS Surveillance Requirement 4.7.1 .2.c.2 are also monitored on a quarterly test frequency underTS Surveillance Requirement 4.7.1.2.b. The EF flow control valve is not closed during the TSsurveillance requirement 4.7.1 .2.b quarterly test. However this test does provide assurance asto the leak tightness of a large portion of the circuit with the exception of the actuator andassociated solenoid. No Condition Reports (CR) were found in recent history associated withthe solenoid for each EF flow control valve which is part of the three hour drop test boundary.In-service test history for the quarterly TS Surveillance Requirement 4.7.1 .2.b surveillanceindicates reliable performance. Problems with unacceptable leakage were encountered in2010. CR-10-01427 documents pressure regulator relief for IFV03541-PR2-EF needed to bereset to restore acceptable leakage. CR-I10-03793 documents high but acceptable leakage for1FV03551-CVI-EF, which was corrected by tightening fittings in the tested boundary.The actuator diaphragms are replaced on an every third refueling (R03) frequency underpreventative maintenance tasks. Preventative maintenance history shows the actuatordiaphragms were last replaced in RF-20.MWR 1110289, (IFV03531-O-EF), task completed on 10/23/12.MWR 1110296, (I FV03541-O-EF), task completed on 10/23/1 2.MWR 1110307, (IFV03551-O-EF), task completed on 10/27/1 2.Recent performance history is documented in the following task sheets for the three hour droptest for the MDEFP flow control valves has been satisfactory.STTS 0800070, RF-17 (5/29/08).STTS 0812592, RF-18 (1 1/26/09).STTS 1004151, RF-19 (5/21/1 1).STTS 1112452, RF-20 (11/12/12).STTS 1307841, RF-21 (5/5/14).The MDEFP valve TS 4.0.5 stroke time history was also reviewed to assess reliability. CRswritten against the flow control valve operator were reviewed. The most recent three CRs werewritten against IFV03551-EF: CR-12-05596, CR-15-02675, and CR-15-04294. Review of CR-12-05596 indicates 1FV03551-EF was successfully retested as allowed by the ASME OM Code.CR-i15-02675 and CR-I15-04294 were related to test conditions rather than structure, system, orcomponent (SSC) degradation.
Document Control DeskAttachment IRC-1 6-0035CR-I16-00848Page 9 of 14Recent maintenance history for the MDEFP flow control valves indicates all were calibratedduring the fall 2015 outage using their associated instrumentation and control (l&C) procedure:ICP-1 95.01 0, ICP-195.011, or ICP-195.012. Leak checks of fittings associated with the valvesfollowing this maintenance are documented under the following Work Orders: 1410934,1410939, and 1410944. The seat, plug/stem of IFV03531-EF was also replaced under WorkOrder 1513005. Step 3 of this Work Order documents leak testing fittings associated with thevalve after this maintenance.3.4 PRA InsightsThe VCSNS PRA (Version 7B4) is the current model of record for internal events. The VCSNSPRA modeling is highly detailed, including a wide variety of initiating events, modeled systems,operator actions, and common cause events. The PRA model quantification process used forthe VCSNS PRA is based on the event tree/fault tree methodology.The initial version of the VCSNS PRA model (March 1993) was used to support the IndividualPlant Examination (IPE) process. Since this model was finalized, there have been more than 30updates, including minor modeling convention changes and data updates, as well as changes toincorporate significant plant modifications, including Chilled Water and Component CoolingWater system modifications, crediting an alternate AC power source, alternate cooling forcharging pumps, and alternate seal injection.SCE&G employs a multi-faceted approach for establishing and maintaining the technicaladequacy and plant fidelity of the VCSNS PRA model. This approach includes both aproceduralized PRA maintenance and update process and the use of independent peerreviews. The findings and observations (F&Os) from the initial peer review and theirresolutions, along with additional F&Os from the 2007 assessment, have been fully addressedand closed.PRA model updates have been performed to address all the identified gaps, and the VCSNSPRA has been independently verified to conform to capability category II of ASME RA-Sb-2005,ASME/ANS Standard for Probabilistic Risk Assessment of Nuclear Power Plant Applications asendorsed by Regulatory Guide (RG) 1.200 Revision 1.RG 1.174 provides guidance on determining acceptable risk increases. The total VCSNS.baseline core damage frequency (CDF) and large early release frequency (LERF) are less than1 .0E-04/yr and I1.0E-05/yr respectively. Limiting the increase to an incremental core damageprobability (ICDP) of 5.0E-07 and incremental large early release probability (ILERP) to 5.0E-08provides margin to RG 1.174 Figures 3 and 4 limits to allow for uncertainties.The EF flow control valve function that is in question is remaining closed for the 24 hour missiontime when required to isolate a faulted or ruptured steam generator. These valves do notcontribute to any PRA modeled initiating event. Therefore their failure to remain closed has noimpact on the probability of occurrence of any initiator, including flooding, fires, or seismic Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 10 of 14events. Flooding and fire events cannot cause a faulted or ruptured Steam Generator so thereis no need for the isolation function of these valves in mitigating those events and no increase inCDF or LERF. Large seismic events (greater than safe shutdown earthquake (SSE)) couldpossibly induce a Steam Generator fault or rupture but its probability in a 24 hour window issmall enough (<2.0E-07) that it has little impact on the risk insights. Therefore the impact of theEF flow control valves not remaining closed for the mission time on mitigation of core damageand large early release sequences can be conservatively estimated by taking all three of thesubject valves (IFV-3531-EF, IFV-3541-EF, and IFV-3551-EF) out of service in the EOOSsoftware using the at power internal events model. This is conservative because itsimultaneously fails both their function to open and throttle EF flow to provide a heat sink as wellas their function to isolate a faulted or ruptured steam generator. With these three valves takenout of service in EOOS, ODE increases by a factor of 67 to 2.08E-04/yr and LERF increases bya factor of 163 to 8.6E-06/yr. Conservatively ignoring the relatively small baseline CDF, thesevalves could be out of service for 5E-07*8760/2.08E-04=21 .05hrs. Similarly for LERF, thesevalves could be out of service for 5.0E-08*8760/8.6E-06=50.93hrs. The ICDP is limiting andresults in an acceptable LCO time of 6+21 =27hrs.The delta risk of shutting down the plant is qualitatively equivalent to that of the increased LCOtime. A shutdown driven by a short TS action statement is more likely to result in a reactor tripthan a controlled shutdown. The conditional core damage probability (CCDP) of a reactor trip is5.3E-07. Therefore the ODE due to an accelerated shutdown is estimated to be on the order of1 .0E-07 to I1.0E-08. Since the valve testing and any needed repairs are expected to becompleted in less than the requested 24 hour LCO time, the outage time estimated risk and theshutdown risk are of the same magnitude. There is little difference in the two scenarios(shutdown vs. staying at power to test and repair the valves).Defense in depth for the heat sink function is provided by the redundant EF flow control valvesassociated with the steam driven emergency feedwater pump and by use of the charging/safetyinjection pumps in the feed and bleed mode. Defense in depth for the function to isolate afaulted or ruptured steam generator is provided by a manually operated stop check valve(XVKI01019A/B/C-EF) in series with each of the flow control valves. Defense in depth for apostulated loss of EF flow associated with a steam line break outside containment (SLBO) in thesupply lines for the TDEFP is by the use of charging/safety injection pumps in the feed andbleed mode.Conservatively meeting the RG 1.174 (Reference 6.10) limits for risk increases and addingcompensatory measures ensures sufficient safety margin to account for analysis and datauncertainties.The following compensatory measures will be taken: Both emergency diesel generators will beverified available (not in Removal and Restoration Log), the TDEFP will be placarded and itsroom locked, a dedicated operator will be stationed locally to manually operate the flow controlvalves as required, the weather forecast will be reviewed for sever conditions (hurricane ortornado), and no other planned maintenance or testing will be in progress prior to entering theaction statement.
Document Control DeskAttachment IRC-1 6-0035CR-i16-00848Page 11 of 1
 
==44.0 REGULATORY EVALUATION==
The EF system automatically supplies feedwater to the Steam Generators to remove decayheat from the reactor coolant system upon the loss of normal feedwater supply. The SteamGenerators function as a heat sink for core decay heat. The heat load is dissipated by releasingsteam to the atmosphere from the Steam Generators via the main steam safety valves. The EFsystem consists of two motor driven pumps and one steam turbine driven pump configured intothree flow paths to supply three Steam Generators by common headers. The EF system isconsidered OPERABLE when the components and flow paths required to provide redundant EFflow to the steam generators are OPERABLE. This requires that the two MDEFP beOPERABLE with two diverse paths, each supplying EF to separate Steam Generators. TheTDEFP is required to be OPERABLE with redundant steam supplies from each of two mainsteam lines upstream of the main steam isolation valves and shall be capable of supplying EFto any of the three Steam Generators. The piping, valves, instrumentation, and controls in therequired flow paths also are required to be OPERABLE.4.1 Applicable Regulatory Requirements I Criteria4.1.1 GDC 34General Design Criteria (GDC) 34 establish the requirements to assure the capability to transferheat from the reactor to a heat sink under normal and accident conditions with sufficientredundancy and isolation capability to accomplish the safety function with a single failure of anactive component with or without a coincident loss of offsite power.The safe shutdown design basis of the Virgil C. Summer Nuclear Station is HOT STANDBY, asit is for all other Westinghouse designed pressurized water reactors. HOT STANDBY is a safeand stable plant condition which can be maintained for an extended period of time following anyCondition II, Ill, or IV event. In the HOT STANDBY condition, residual heat removal, incompliance with GDC 34 (10OCFR5O, Appendix A), is provided by the EF system in conjunctionwith the Steam Generator safety valves. Cross connections from the service water system tothe EF system provide a long term (i.e., greater than 7 days) source of EF. (ESAR, Section5.5.7.3.1)4.1.2 10 CFR 50.6210 CFR 50.62 requires that pressurized water reactors have equipment diverse from the reactorprotection system to initiate the EF system under conditions indicative of an ATWS. The EFsystem is required to assure adequate removal of heat from the reactor coolant system duringan ATVVS.
Document Control DeskAttachment 1RC-1 6-0035CR-i16-00848Page 12 of 14The worst common mode failure which is postulated to occur is the failure to scram the reactorafter an anticipated transient has occurred. The effects of ATWVS events are not considered aspart of the design basis for transients analyzed in Chapter 15. The final NRC ATWS rulerequires that Westinghouse designed plants install ATWVS Mitigation System Actuation Circuitry(AMSAC) to initiate a turbine trip and actuate EF flow independent of the Reactor ProtectionSystem. The V. C. Summer AMSAC design is described in ESAR Section 7.8.4.2 PrecedentNone.4.3 No Significant Hazards Consideration1. Does the proposed amendment involve a significant increase in the probability orconsequences of an accident previously evaluated?No. A onetime change to the action statement of TS 3.7.1.2, ACTION b, does notincrease the probability or consequences of any analyzed accident addressed withinESAR Chapter 15. The EF system is not an initiator of any Chapter 15 accidents, andthe one-time change does not make it an initiator. Therefore, there cannot be anincrease in the probability of an accident previously evaluated.The relevant consequences stem from the ability to maintain core cooling. The changeis not detrimental to the ability to remove core heat, because while the maintenance isbeing performed affects the two MDEFPs, the TDEFP remains available for SteamGenerator cooling. A review of the Chapter 15 analyses shows that for single failureconsiderations, only one safety train is credited for accident mitigation. Events creditingEF flow assume one EF pump is able to deliver flow to the Steam Generators. This ispreserved by maintaining the availability and operability of the TDEFP. The only specificcircumstance in which TDEFP operation could be potentially affected is the occurrenceof a break of the Main Steam 4" branch line that supplies steam to the TDEFP. Sincethe activity does not involve a change to the main steam system, or otherwise affects theability of the main steam system to supply the TDEFWP, there cannot be an increase inthe probability of such a break. Nonetheless, in the unlikely event that the MDEFPflowpaths cannot be restored quickly because of that break, and with the area potentiallyinaccessible, core cooling can still be assured by initiating safety injection to establishfeed and bleed cooling, as di'rected by the Emergency Operating Procedures (EOPs).Failure to automatically isolate EF to the affected Steam Generator is an importantconsideration within two secondary pipe break analyses. For secondary side pipebreaks inside containment (FSAR Section 6.2), operator action at 30 minutes is creditedto isolate EF to the affected Steam Generator. Local or remote operator action within 30minutes is required to prevent overpressurizing the containment. Secondly, forsecondary side pipe breaks outside containment (FSAR Section 3.11.2.2.2.2 and Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 13 of 1410.4.9.3), credit is taken for operator action at 10 minutes to isolate EF to the affectedSteam Generator. Since the harsh environment will limit local access and manualactions, operator action from the control room is required for secondary pipe breaksoutside containment to preserve environment conditions for equipment qualification.Extending the action statement from six hours to 24 hours does not increase theprobability or consequences of an accident previously evaluated.2. Does the proposed amendment create the possibility of a new or different kind ofaccident from any accident previously evaluated?No. Extension of the action statement does not create the possibility of a new ordifferent kind of accident from any accident previously evaluated. In the case ofsecondary breaks outside the reactor building, which would make the flow control valvesinaccessible for local operation, procedural guidance outside of the EOPs directs theoperators to take alternative action (secure the MDEFPs) if the flow control valveassociated with the faulted Steam Generator cannot be closed from the control board.Increasing the duration of the allowed action from six hours to 24 hours does not resultin a new or different kind of accident.3. Does the proposed amendment involve a significant reduction in a margin ofsafety?No. The relevant margin of safety stems from the ability to maintain core cooling usingthe Steam Generators. As described previously, the continued operability of the TDEFPpreserves the core cooling function in the event of an emergency. The postulation of asingle failure is not required while in the LCO Action Time. Nonetheless, because of theavailability of safety injection and the ability to perform feed and bleed cooling, corecooling will be assured. Therefore, there will not be a significant reduction in the marginof safety.4.4 ConclusionIn conclusion, based on the considerations discussed above, (1) there is reasonable assurancethat the health and safety of the public will not be endangered by operation in the proposedmanner, (2) such activities will be conducted in compliance with the Commission's regulations,and (3) the issuance of the amendment will not be inimical to the common defense and securityor to the health and safety of the public.
Document Control DeskAttachment IRC-1 6-0035CR-I16-00848Page 14 of 145.0 Environmental considerationA review has determined that the proposed amendment would change a requirement withrespect to installation or use of a facility component located within the restricted area, as definedin 10 CFR 20, or would change an inspection or surveillance requirement. However, theproposed amendment does not involve (i) a significant hazards consideration, (ii) a significantchange in the types or a significant increase in the amounts of any effluents that may bereleased offsite, or (iii) a significant increase in individual or cumulative occupational radiationexposure. Accordingly, the proposed amendment meets the eligibility criterion for categoricalexclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), noenvironmental impact statement or environmental assessment need be prepared in connectionwith the proposed amendment.
 
==6.0 REFERENCES==
6.1 10CFR50, Appendix A6.2 FSAR Section 3.11.2.2.2.26.3 FSAR Section 5.5.7.3.16.4 FSAR Section 6.26.5 FSAR Section 7.86.6 FSAR Section 10.4.96.7 ESAR Chapter 156.8 VCSNS Drawing6.9 VCSNS Drawing6.10 RG 1.174General Design Criteria for Nuclear Power PlantsMain Steam Line Break Outside Containment EquipmentQualificationRESIDUAL HEAT REMOVAL SYSTEM, System Availabilityand ReliabilityContainment SystemsATWS MITIGATION SYSTEM ACTUATION CIRCUITRY(AMSAC)EMERGENCY FEEDWATER SYSTEMACCIDENT ANALYSESI MS-5O-I181, 1 MS-SO-I182 Fisher 657-ET Diaphragm ControlValve302-085 (ESAR Figure 10.4-16) Emergency Feedwater FlowDiagram,An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis Document Control DeskAttachment 2RC-I16-0035CR-I16-00848Page 1 of 2VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1ATTACHMENT 2PROPOSED TECHNICAL SPECIFICATION CHANGE (MARK-UP)Proposed Technical Specification Changes SummaryPae Affected Bar # Description of Change -Reason for ChangeSection3/4 7-4 3.7.1.2. Add note for one time EXIGENTAction b action requirementsallowing 24 hours to be inat least HOT STANDBYdue to two inoperablepumps or flow paths Document Control DeskAttachment 2RC-1 6-0035CR-i16-00848Page 2 of 2PLANT SYSTEMSEMERGENCY FEEDWATER SYSTEMLIMITNG CONDITION FOR OPERATION3.7.1.2 At least three independent steam generator emergency feedwater pumps and flowpaths shall be OPERABLE with:a. Two motor-driven emergency feedwater pumps, each capable of being poweredfrom separate emergency busses, andb. One steam turbine driven emergency feedwater pump capable of being poweredfrom an OPERABLE steam supply system.APPLICABILITY: MODES 1, 2 and 3.ACTION:Ada Wihone eegnyfdatrpumpinerberstethreuemergency feedwater pumps to OPERABLE status withi or bein at least HOT STANDBY within the next 6 ho HOT SHUTDOWNwithin the following 6 b. ithtwoemeg _ate pumps inoperable, be in at least HOT STANDBYwithin 6 hours a in HOT SHUTDOWN within the following 6 hours.c. With three emergency feedwater pumps inoperable, immediately initiatecorrective action to restore at least one emergency feedwater pump toOPERABLE status as soon as possible.SURVEILLANCE REQUIREMENTS4.7.1.2 Each emergency feedwater pump shall be demonstrated OPERABLE:a. At least once per 31 days by:1. Verifying that each motor driven pump develops a total head of greater jthan or equal to 3800 feet at greater than or equal to 90 gpm flow.2. Verifying that the steam turbine driven pump develops a total head of Jgreater than or equal to 3140 feet at a flow of greater than or equal to 97gpm when the secondary steam supply pressure is greater than 865 psig.The provisions of Specification 4.0.4 are not applicable.3. Verifying that each non-automatic valve in the flow path that is not locked,sealed, or otherwise secured in position, is in its correct position.*The ACTION to be in at least HOT STANDBY in 6 hours is extended to 24 hours to test (and performremedial maintenance on) the motor driven emergency feedwater pump flow control valves per surveillancerequirement 4.7.1.2.c.2. This extension expires on March 18, 2016.SUMMER- UNIT 13/4 7-4Amendment No.
Document Control DeskAttachment 3RC-I16-0035CR-I16-00848Page 1 of 2VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1ATTACHMENT 3PROPOSED TECHNICAL SPECIFICATION CHANGE (RETYPED)Replace the following pages of the Technical Specifications with the attached revised pages.The revised pages are identified by amendment number and contain marginal lines indicatingthe areas of change.Remove Paqes3/4 7-4Insert Pages3/4 7-4 PLANT SYSTEMSEMERGENCY FEEDWATER SYSTEMLIMITING CONDITION FOR OPERATION3.7.1.2 At least three independent steam generator emergency feedwater pumps and flowpaths shall be OPERABLE with:a. Two motor-driven emergency feedwater pumps, each capable of being poweredfrom separate emergency busses, andb. One steam turbine driven emergency feedwater pump capable of being poweredfrom an OPERABLE steam supply system.APPLICABILITY: MODES 1, 2 and 3.ACTION:a. With one emergency feedwater pump inoperable, restore the requiredemergency feedwater pumps to OPERABLE status within 72 hours or bein at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWNwithin the following 6 hours.b. With two emergency feedwater pumps inoperable, be in at least HOT STANDBYwithin 6 hours* and in HOT SHUTDOWN within the following 6 hours.c. With three emergency feedwater pumps inoperable, immediately initiatecorrective action to restore at least one emergency feedwater pump toOPERABLE status as soon as possible.SURVEILLANCE REQUIREMENTS4.7.1.2 Each emergency feedwater pump shall be demonstrated OPERABLE:a. At least once per 31 days by:1. Verifying that each motor driven pump develops a total head of greaterthan or equal to 3800 feet at greater than or equal to 90 gpm flow.2. Verifying that the steam turbine driven pump develops a total head ofgreater than or equal to 3140 feet at a flow of greater than or equal to 97gpm when the secondary steam supply pressure is greater than 865 psig.The provisions of Specification 4.0.4 are not applicable.3. Verifying that each non-automatic valve in the flow path that is not locked,sealed, or otherwise secured in position, is in its correct position.* The ACTION to be in at least HOT STANDBY in 6 hours is extended to 24 hours to test (andperform remedial maintenance on) the motor driven emergency feedwater pump flow controlvalves per surveillance requirement 4.7.1.2.c.2. This extension expires on March 18, 2016.SUMMER- UNIT 13/4 7-4SUMMR-UIT I3/47-4Amendment No. 112, 111, 173, Document Control DeskAttachment 4RC-1 6-0035CR-I16-00848Page 1 of 1VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1ATTACHMENT 4LIST OF REGULATORY COMMITMENTSThere is one regulatory commitment created due to this License Amendment Request. Anyother statements in this submittal are provided for information purposes and are not consideredto be regulatory commitments. Please direct questions regarding these commitments toMr. Bruce L. Thompson at (803) 931-5042.Commitment Due DateLicense Amendment Request submitted to remove the note June 30, 2017permitting 24 hours to HOT STANDBY.
George A. LippardVice President, Nuclear Operations803.345.48101, 2016A SCANA COMPANY RC-1 6-0035Document Control DeskU. S. Nuclear Regulatory CommissionWashington, DC 20555
 
==Dear Sir I Madam:==
 
==Subject:==
VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1DOCKET NO. 50-395OPERATING LICENSE NO. NPF-12EXIGENT LICENSE AMENDMENT REQUEST -LAR (1 6-00848)TECHNICAL SPECIFICATION CHANGE REQUEST FOR THE EMERGENCYFEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2ACTION bSouth Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for SouthCarolina Public Service Authority pursuant to 10 CFR 50.90 and 10 CFR 50.91, hereby submitsa request for an exigent amendment to Technical Specifications (TS). The proposedamendment would modify the action statement for two inoperable pumps or flow paths withinSection 3.7.1.2, "Plant Systems -Emergency Feedwater System."Attachment I provides an evaluation of the proposed change to the action statement to amendthe six hour action to be in at least HOT STANDBY to 24 hours to allow for maintenance andretesting. This amendment request was evaluated and found to have no significant hazards forconsideration. An exigent TS change is justified in that compliance with TS could involve anunnecessary plant action to shutdown the reactor to COLD SHUTDOWN and potential relianceon the turbine driven emergency feedwater pump for plant cooldown without a correspondinghealth and safety benefit. The station proposes that the action statement be amended to 24hours to allow for maintenance and retesting. Attachment 2 contains the marked-up version ofthe affected TS page. Attachment 3 contains the reprinted versions of the affected TS page.In accordance with 10 CFR 50.91, a copy of this application, with attachments, is being providedto the designated South Carolina Official. This proposed change has been reviewed andapproved by both the VCSNS Plant Safety Review Committee and the VCSNS Nuclear SafetyReview Committee.SCE&G requests approval of the proposed amendment by March 10, 2016. Once approved,the amendment shall be implemented immediately.The proposed change does introduce one new commitment. If you have any questions orrequire additional information, please contact Bruce Thompson at (803) 931-5042.V. C. Summer Nuclear Station .P. O. Box 88
* Jenkinsville, SC. 29065.* F (803) 941-9776 Document Control DeskRC-1 6-0 035CR-I16-00848Page 2 of 2I certify under penalty of perjury that the information contained herein is true and correct.Executed onG ieo e A i~rWLT/GAL/Attachments:1. Analysis of Proposed Technical Specification Change2. Proposed Changes -Marked Up TS Page3. Proposed TS Pages -Retyped4. Commitment Pagec: K. B. MarshS. A. ByrneJ. B. ArchieN. S. CamnsJ. H. HamiltonJ. W. WilliamsW. M. CherryC. HaneyS. A. WilliamsNRC Resident InspectorK. M. SuttonP. LedbetterS. E. JenkinsNSRCRTS (CR-I16-00848)File (813.20)PRSF (RC-16-0035)
Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 1 of 14VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)DOCKET NO. 50-395OPERATING LICENSE NO. NPF-12Attachment 1*Analysis of Proposed Technical Specification Change
 
==Subject:==
This evaluation supports a request to amend South Carolina Electric & GasCompany (SCE&G), Technical Specifications (TS) to modify the action statementof 3.7.1.2, Emergency Feedwater System, Limiting Conditions For Operation(LCO) for two inoperable motor driven pumps.1.0 SUMMARY DESCRIPTIONIn accordance with the provisions of 10 CFR 50.90, South Carolina Electric & Gas Company,acting for itself and as agent for South Carolina Public Service Authority, requests NuclearRegulatory Commission (NRC) review and approval to amend Operating License NPF-12 forVirgil C. Summer Nuclear Station (VCSNS) Unit 1.VCSNS is proposing an exigent TS change. It is the station's position that compliance with TScould involve an unnecessary plant shutdown and the potential reliance on the turbine drivenemergency feedwater pump (TDEFP) for plant shutdown without a corresponding health andsafety benefit. Due to an oversight, the station missed a surveillance test during the fall 2015startup from refueling outage 22 (RF-22) associated with the emergency feedwater (EF) controlvalves in accordance within VCSNS Technical Specification 4.7.1.2.c.2. This surveillancerequires verifying the flow control valves can be closed and held closed for three hours whennormal instrument air is not available. The surveillance is normally conducted in Mode 4 orbelow when the Steam Generators are not relied on for heat removal. Due to the designconfiguration of the EF system, the six hour action statement b for two inoperable emergencyfeedwater pumps is entered anytime a motor driven emergency feedwater pump (MDEFP) flowcontrol valve is closed in modes 1, 2 or 3. The station proposes to modify limiting conditions foroperation 3.7.1.2 action statement b which currently requires: for two inoperable emergencyfeedwater pumps, be in at least HOT STANDBY within six hours and be in HOT SHUTDOWNwithin the following six hours. The station proposes that the action statement be amended to bein at least HOT STANDBY within 24 hours to allow for timely completion of any requiredmaintenance and surveillance retest.
Document Control DeskAttachment 1RC-I16-0035CR-I16-00848Page 2 of 142.0 DETAILED DESCRIPTIONDue to an oversight the station has missed performing a surveillance associated with the EFcontrol valves as reflected within TS 4.7.1 .2.c.2 during the startup from refueling outage 22.The surveillance requirement is for at least once per 18 months during shutdown and is typicallycompleted in HOT SHUTDOWN or below when the steam generators are not relied on for heatremoval. This test requires the MDEFP flow control valves be held closed for 3 hours with airfrom the accumulators. Due to the design configuration of the EF system, the six hour actionstatement b for two inoperable EF pumps is entered anytime a MDEFP flow control valve isclosed in modes 1, 2 or 3. With the test time period of three hours, no time is available toconduct remedial corrective maintenance and repeat the surveillance. This could result in anunnecessary plant shutdown.The end date for this test is March 17, 2016, based on an 18 month surveillance interval plus25% per TS 4.0.2. During performance of the General Operating Procedure (GOP-2) for PlantStartup and Heatup (MODE 5 to MODE 3), the surveillance was thought to be complete basedon completion of surveillance for the TDEFP flow control valves. This error was not detecteduntil the plant was in HOT SHUTDOWN at approximately 345 degrees Fahrenheit and wasrelying on MDEFP flow for heat removal as is normal for the start-up process.The station has prepared to conduct the test during Model by entering the TS 3.7.1.2 actionstatement b to conduct the test. However, conducting the 3 hour surveillance test at powerwhile in a 6 hour shutdown action statement leaves no time to make repairs and conduct aretest. Changing the action statement to 24 hours will allow for unforeseen correctivemaintenance and subsequent retest would prevent the station from an unnecessary plantshutdown without a corresponding health and safety benefit.2.1 Possible Repairs TimelineThe following potential component failures could be required following the surveillance. Theestimated repair times are based on repairing each item identified below and include tagging outthe appropriate isolation devices. The time reflected also accounts for retesting of the EFcontrol valves to ensure the capability to hold the valve closed for three hours as required by TS4.7.1 .2.c.2.Air accumulator check valve replacement -18 hours.Air pressure regulator rebuild and calibration -10 hours.Air actuator diaphragm casing bolts torque adjustment -8 hours.Air actuator diaphragm replacement -12 hours.Air solenoid valve replacement -14 hours.Air relief valve replacement and setup -8 hours.
Document Control DeskAttachment 1RC-1 6-0035CR-i16-00848Page 3 of 142.2 Bases for Exigent ChangeSurveillance Test Procedure (STP)-1 20.006, "Emergency Feedwater Valves Backup Air SupplyTest," was not performed during the fall 2015 outage for the MDEFP flow control valves. Theend date for this test is March 17, 2016, based on an 18 month surveillance interval plus 25%per TS 4.0.2.During performance of GOP-2, "Plant Startup And Heatup (MODE 5 to MODE 3)," thissurveillance was signed off as being complete based on completion of the A-train portion of thetest, which tests the TDEFP flow control valves, done under a separate task sheet from the onewritten for the B-train valves (MDEFP flow control valves). This error was not detected untilperformance of General Testing Procedure (GTP-702), "Surveillance Activity Tracking andTriggering," for Mode 3 entry, which lists the A-train and B-train tasks as separate line items. Bythat time, the plant was relying on MDEFP flow and the steam generators for heat removal as isnormal during start-up. To perform the testing at this point in the outage would require the plantto cool down to less than 183 degrees Fahrenheit and reinitiate Residual Heat Removal (RHR)cooling. The precautions in STP-120.006 showed the procedure allows the subject testing inModes 1, 2, and 3 as long as both emergency diesel generators are operable with nomaintenance or testing in progress on either emergency diesel generator.Because the procedure allowed testing in Mode 1, 2, or 3 the decision was made to not cool theplant back down to less than 183 degrees Fahrenheit and reinitiate RHR cooling, but instead toschedule the performance of the required testing in Mode 1 once the plant reached a 100%power.The surveillance test was placed in the plant online work week schedule to be performed onFebruary 26, 2016. During a normal process schedule review on January 30, 2016, it wasdiscussed that this test would need additional focus to be performed online due to the shortduration six hour action to HOT STANDBY required by TS 3.7.1.2 action b. Station personnelthen began to apply additional planning considerations and focus to the testing includingdesignating the test as an Infrequently Performed Test or Evolution (IPTE) and developingcontingencies for repairs should valve repairs be required. After input by several plant groupsthe contingency matrix was finalized late on February 23, 2016. Based on the estimated timesfor repairs and retesting in the matrix, it was determined that a reasonable repair could not beapplied within the actions specified in TS 3.7.1.2 action b. The test was rescheduled to beperformed on March 11, 2016, to allow additional planning time. While the actual end date forsurveillance is March 17, 2016, major maintenance has been scheduled on one of theemergency diesel generators for the week of March 13, 2016.
Document Control DeskAttachment IRC-1 6-0035CR-16-00848Page 4 of 14While the station has prepared to conduct the test during Mode 1 by entering the TS 3.7.1.2action statement b to conduct the test, conducting the 3 hour surveillance test at power while ina 6 hour shutdown action statement leaves no time to make repairs and conduct a retest.Changing the action statement to allow 24 hours to allow for unforeseen corrective maintenanceand subsequent retest would prevent the station from an unnecessary plant shutdown without acorresponding health and safety benefit.
 
==3.0 TECHNICAL EVALUATION==
The EF system consists of three pumps, two motor driven and one steam turbine driven. TheEF System is used to supply feedwater to the steam generators during startup, shutdown, andlayup operations. A simplified system drawing is shown as Figure 1 where the full version canbe viewed within the FSAR Figure 10.4-16 or VCSNS drawing 302-085.The EF flow control valves fail open which is the "safe" position for most accidents but are alsorequired to be closed due to a faulted Steam Generator. The flow control valves are held closedfor three hours by their associated air accumulators following an EF high-flow signal to a faultedSteam Generator. The three hour time permits automatic valve closure following a secondarysystem break when local valve operation cannot be accomplished because local conditions areunsuitable for personnel access. The valves are supported by safety class air accumulatorswith sufficient capacity to permit remote valve closure for at least three hours during a loss ofinstrument air system. The air accumulators provide a regulated air supply as needed to closethe valves against spring force. The accumulators are supported by a non-safety instrument airsystem. Additionally, each valve has a handwheel to provide manual control.3.1 System OperationThe EF system "is required to deliver sufficient feedwater to the Steam Generators for cooldownupon loss of the normal feedwater supply and during an Anticipated Transient Without Scram(ATVVS) event. The EF system is used to supply feedwater to the Steam Generators duringstartup, shutdown, and iayup operations. The EF system operates in conjunction with theturbine bypass system, if available, or the main steam power relief valves and safety valves, toremove thermal energy from the Steam Generators.The system is designed to automatically deliver feedwater, at a minimum total flow of 380 gpm,to at least two Steam Generators pressurized to 1211 psig. There is sufficient redundancy toestablish this flow while sustaining a single active failure in the system in the short term or asingle active or passive failure in the long term. The EF system operates until the RHR Systemcan be placed in operation.
Document Control DeskAttachment 1RC-l16-0035CR-I16-00848Page 5 of 14When forced circulation from the reactor coolant pumps is not available, EF operation isrequired down to a main steam pressure of 100 psia. This corresponds to a reactor coolant coldleg temperature of 325 degrees Fahrenheit and a hot leg temperature of 350 degreesFahrenheit. The primary coolant temperature differential is required in order to maintain adensity gradient to drive natural circulation of primary coolant, in the absence of reactor coolantpump operation.Sufficient feedwater is available under emergency conditions to bring the plant to a safeshutdown condition. Assuming prior plant operation at engineered safety design rating (ESDR)of 2900 MWt in the core, the minimum required usable volume for the condensate storage tankis 158,570 gallons based on maintaining the plant at hot standby conditions for 11 hours.This volume also satisfies the minimum required volume to cool down the plant to HOTSHUTDOWN conditions assuming the plant is maintained at HOT STANDBY for 2 hours andthen cooled down to HOT SHUTDOWN in 4 hours.The system consists of three pumps, two motor driven and one steam turbine driven. The twomotor driven pumps share a common discharge header that splits off into three branches. Eachbranch has a pneumatic flow control valve, which controls flow to its respective steamgenerator. The one steam turbine driven pump has a separate header from the motor drivenpumps: This header splits off into three branches, which controls flow to its respective SteamGenerator.The three flow control valves for the turbine driven pump have control elements fed from A-trainpower. The three flow control valves for the motor driven pumps have control elements fed fromB-train power.During the performance of testing of the B-train flow control valves, the flow path from both ofthe motor driven pumps is disrupted. This is due to both pumps sharing a common dischargeheader prior to branching off to the three flow control valves.The MDEFPs are powered from separate and independent safety related emergency dieselgenerator backed buses.EF is a dual purpose system. During normal operation, the motor driven pumps are used duringheatup and cooldown to supply feedwater to the steam generators for reactor coolant systemtemperature control. During emergency operation, all three pumps can provide feedwater tosupport reactor coolant system heat sink capabilities via the Steam Generators.
Document Control DeskAttachment 1RC-1 6-0035CR-i16-00848Page 6 of 14EMERGENCY FEEDWATER SYSTEMFigure 1 Simplified System Drawing3.2 Component DesignThe EF System design consists of two redundant trains: i.e., the motor driven pump train andthe turbine driven pump train. The motor driven pump train is designed for use during normalplant conditions (i.e. startup, hot standby, and cooldown) and for emergency shutdown of thereactor. The turbine driven pump train is designed for use for emergency shutdown of thereactor. EF initiation arises from any of several types of signals, which may be generated inresponse to a variety of plant conditions. These initiation signals are generated in response tolow Steam Generator levels, loss of main feedwater, low voltage on the essential electric powerbuses, a Safety Injection signal, and an ATWVS mitigation signal.The control systems for the turbine driven and motor driven pump flow control valves areidentical. Automatic valve opening signals are generated by the reactor protection and logicsystem and depend upon the given plant condition which will determine whether only the motordriven pump flow control valves open or the turbine driven pump valves open also. Thesevalves will receive an open signal whenever their respective pumps receive an auto-start. Theexception is the MDEFP flow control valves which do not receive an open signal when all threemain feed pumps trip. If the flow control valves are in MANUAL control, the valves fully open inresponse to an automatic open signal.
Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 7 of 14The EF system is designed with three flow control valves at the discharge of the TDEFP andthree flow control valves at the discharges of the MDEFPs. The valves are required to controlEF flow to the Steam Generators to maintain program level and to produce sufficient mainsteam to permit main feedwater pump turbine operation and for plant cooldown after mainsteam is no longer able to drive the main feedwater pump turbines. All six valves are identical,3-inch Fisher ET, normally open, air-operated, globe valves. The valves are safety relateddevices which meet the requirements of ASME B&PV Code Section III, class 2, 1974 edition,Summer, 1975 addenda.The valves fail open which is the "safe" position for most accidents. The valves are supportedby safety-class air accumulators with sufficient capacity to permit remote valve closure on ahigh-flow signal and maintain the valve closed for at least three hours during a loss ofinstrument air system. The air accumulators provide a regulated air supply as needed to closethe valves against spring force. The accumulators are supported by a non-safety instrument airsystem. Additionally each valve has a handwheel to allow manual control.The primary safety function of the air accumulator is to assure a source of safety related air isavailable to isolate the flow control valve to a faulted Steam Generator. The three hour supplypermits automatic or remote manual valve closure following a secondary system break whenlocal valve operation cannot be accomplished due to unsuitable conditions for personnel accessin the Intermediate Building. Figure 2 is provided as a simplified sketch of the control airsystem.b .......Figure 2 Simplified Control Air System (VCSNS Drawing 817-056-001)
Document Control DeskAttachment IRC-1 6-0035CR-I16-00848Page 8 of 143.3 Component HistoryThe MDEFP flow control valves are subject to periodic stroke time testing to the open andclosed positions under TS 4.0.5. Portions of the test circuit in the pressure drop tested underTS Surveillance Requirement 4.7.1 .2.c.2 are also monitored on a quarterly test frequency underTS Surveillance Requirement 4.7.1.2.b. The EF flow control valve is not closed during the TSsurveillance requirement 4.7.1 .2.b quarterly test. However this test does provide assurance asto the leak tightness of a large portion of the circuit with the exception of the actuator andassociated solenoid. No Condition Reports (CR) were found in recent history associated withthe solenoid for each EF flow control valve which is part of the three hour drop test boundary.In-service test history for the quarterly TS Surveillance Requirement 4.7.1 .2.b surveillanceindicates reliable performance. Problems with unacceptable leakage were encountered in2010. CR-10-01427 documents pressure regulator relief for IFV03541-PR2-EF needed to bereset to restore acceptable leakage. CR-I10-03793 documents high but acceptable leakage for1FV03551-CVI-EF, which was corrected by tightening fittings in the tested boundary.The actuator diaphragms are replaced on an every third refueling (R03) frequency underpreventative maintenance tasks. Preventative maintenance history shows the actuatordiaphragms were last replaced in RF-20.MWR 1110289, (IFV03531-O-EF), task completed on 10/23/12.MWR 1110296, (I FV03541-O-EF), task completed on 10/23/1 2.MWR 1110307, (IFV03551-O-EF), task completed on 10/27/1 2.Recent performance history is documented in the following task sheets for the three hour droptest for the MDEFP flow control valves has been satisfactory.STTS 0800070, RF-17 (5/29/08).STTS 0812592, RF-18 (1 1/26/09).STTS 1004151, RF-19 (5/21/1 1).STTS 1112452, RF-20 (11/12/12).STTS 1307841, RF-21 (5/5/14).The MDEFP valve TS 4.0.5 stroke time history was also reviewed to assess reliability. CRswritten against the flow control valve operator were reviewed. The most recent three CRs werewritten against IFV03551-EF: CR-12-05596, CR-15-02675, and CR-15-04294. Review of CR-12-05596 indicates 1FV03551-EF was successfully retested as allowed by the ASME OM Code.CR-i15-02675 and CR-I15-04294 were related to test conditions rather than structure, system, orcomponent (SSC) degradation.
Document Control DeskAttachment IRC-1 6-0035CR-I16-00848Page 9 of 14Recent maintenance history for the MDEFP flow control valves indicates all were calibratedduring the fall 2015 outage using their associated instrumentation and control (l&C) procedure:ICP-1 95.01 0, ICP-195.011, or ICP-195.012. Leak checks of fittings associated with the valvesfollowing this maintenance are documented under the following Work Orders: 1410934,1410939, and 1410944. The seat, plug/stem of IFV03531-EF was also replaced under WorkOrder 1513005. Step 3 of this Work Order documents leak testing fittings associated with thevalve after this maintenance.3.4 PRA InsightsThe VCSNS PRA (Version 7B4) is the current model of record for internal events. The VCSNSPRA modeling is highly detailed, including a wide variety of initiating events, modeled systems,operator actions, and common cause events. The PRA model quantification process used forthe VCSNS PRA is based on the event tree/fault tree methodology.The initial version of the VCSNS PRA model (March 1993) was used to support the IndividualPlant Examination (IPE) process. Since this model was finalized, there have been more than 30updates, including minor modeling convention changes and data updates, as well as changes toincorporate significant plant modifications, including Chilled Water and Component CoolingWater system modifications, crediting an alternate AC power source, alternate cooling forcharging pumps, and alternate seal injection.SCE&G employs a multi-faceted approach for establishing and maintaining the technicaladequacy and plant fidelity of the VCSNS PRA model. This approach includes both aproceduralized PRA maintenance and update process and the use of independent peerreviews. The findings and observations (F&Os) from the initial peer review and theirresolutions, along with additional F&Os from the 2007 assessment, have been fully addressedand closed.PRA model updates have been performed to address all the identified gaps, and the VCSNSPRA has been independently verified to conform to capability category II of ASME RA-Sb-2005,ASME/ANS Standard for Probabilistic Risk Assessment of Nuclear Power Plant Applications asendorsed by Regulatory Guide (RG) 1.200 Revision 1.RG 1.174 provides guidance on determining acceptable risk increases. The total VCSNS.baseline core damage frequency (CDF) and large early release frequency (LERF) are less than1 .0E-04/yr and I1.0E-05/yr respectively. Limiting the increase to an incremental core damageprobability (ICDP) of 5.0E-07 and incremental large early release probability (ILERP) to 5.0E-08provides margin to RG 1.174 Figures 3 and 4 limits to allow for uncertainties.The EF flow control valve function that is in question is remaining closed for the 24 hour missiontime when required to isolate a faulted or ruptured steam generator. These valves do notcontribute to any PRA modeled initiating event. Therefore their failure to remain closed has noimpact on the probability of occurrence of any initiator, including flooding, fires, or seismic Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 10 of 14events. Flooding and fire events cannot cause a faulted or ruptured Steam Generator so thereis no need for the isolation function of these valves in mitigating those events and no increase inCDF or LERF. Large seismic events (greater than safe shutdown earthquake (SSE)) couldpossibly induce a Steam Generator fault or rupture but its probability in a 24 hour window issmall enough (<2.0E-07) that it has little impact on the risk insights. Therefore the impact of theEF flow control valves not remaining closed for the mission time on mitigation of core damageand large early release sequences can be conservatively estimated by taking all three of thesubject valves (IFV-3531-EF, IFV-3541-EF, and IFV-3551-EF) out of service in the EOOSsoftware using the at power internal events model. This is conservative because itsimultaneously fails both their function to open and throttle EF flow to provide a heat sink as wellas their function to isolate a faulted or ruptured steam generator. With these three valves takenout of service in EOOS, ODE increases by a factor of 67 to 2.08E-04/yr and LERF increases bya factor of 163 to 8.6E-06/yr. Conservatively ignoring the relatively small baseline CDF, thesevalves could be out of service for 5E-07*8760/2.08E-04=21 .05hrs. Similarly for LERF, thesevalves could be out of service for 5.0E-08*8760/8.6E-06=50.93hrs. The ICDP is limiting andresults in an acceptable LCO time of 6+21 =27hrs.The delta risk of shutting down the plant is qualitatively equivalent to that of the increased LCOtime. A shutdown driven by a short TS action statement is more likely to result in a reactor tripthan a controlled shutdown. The conditional core damage probability (CCDP) of a reactor trip is5.3E-07. Therefore the ODE due to an accelerated shutdown is estimated to be on the order of1 .0E-07 to I1.0E-08. Since the valve testing and any needed repairs are expected to becompleted in less than the requested 24 hour LCO time, the outage time estimated risk and theshutdown risk are of the same magnitude. There is little difference in the two scenarios(shutdown vs. staying at power to test and repair the valves).Defense in depth for the heat sink function is provided by the redundant EF flow control valvesassociated with the steam driven emergency feedwater pump and by use of the charging/safetyinjection pumps in the feed and bleed mode. Defense in depth for the function to isolate afaulted or ruptured steam generator is provided by a manually operated stop check valve(XVKI01019A/B/C-EF) in series with each of the flow control valves. Defense in depth for apostulated loss of EF flow associated with a steam line break outside containment (SLBO) in thesupply lines for the TDEFP is by the use of charging/safety injection pumps in the feed andbleed mode.Conservatively meeting the RG 1.174 (Reference 6.10) limits for risk increases and addingcompensatory measures ensures sufficient safety margin to account for analysis and datauncertainties.The following compensatory measures will be taken: Both emergency diesel generators will beverified available (not in Removal and Restoration Log), the TDEFP will be placarded and itsroom locked, a dedicated operator will be stationed locally to manually operate the flow controlvalves as required, the weather forecast will be reviewed for sever conditions (hurricane ortornado), and no other planned maintenance or testing will be in progress prior to entering theaction statement.
Document Control DeskAttachment IRC-1 6-0035CR-i16-00848Page 11 of 1
 
==44.0 REGULATORY EVALUATION==
The EF system automatically supplies feedwater to the Steam Generators to remove decayheat from the reactor coolant system upon the loss of normal feedwater supply. The SteamGenerators function as a heat sink for core decay heat. The heat load is dissipated by releasingsteam to the atmosphere from the Steam Generators via the main steam safety valves. The EFsystem consists of two motor driven pumps and one steam turbine driven pump configured intothree flow paths to supply three Steam Generators by common headers. The EF system isconsidered OPERABLE when the components and flow paths required to provide redundant EFflow to the steam generators are OPERABLE. This requires that the two MDEFP beOPERABLE with two diverse paths, each supplying EF to separate Steam Generators. TheTDEFP is required to be OPERABLE with redundant steam supplies from each of two mainsteam lines upstream of the main steam isolation valves and shall be capable of supplying EFto any of the three Steam Generators. The piping, valves, instrumentation, and controls in therequired flow paths also are required to be OPERABLE.4.1 Applicable Regulatory Requirements I Criteria4.1.1 GDC 34General Design Criteria (GDC) 34 establish the requirements to assure the capability to transferheat from the reactor to a heat sink under normal and accident conditions with sufficientredundancy and isolation capability to accomplish the safety function with a single failure of anactive component with or without a coincident loss of offsite power.The safe shutdown design basis of the Virgil C. Summer Nuclear Station is HOT STANDBY, asit is for all other Westinghouse designed pressurized water reactors. HOT STANDBY is a safeand stable plant condition which can be maintained for an extended period of time following anyCondition II, Ill, or IV event. In the HOT STANDBY condition, residual heat removal, incompliance with GDC 34 (10OCFR5O, Appendix A), is provided by the EF system in conjunctionwith the Steam Generator safety valves. Cross connections from the service water system tothe EF system provide a long term (i.e., greater than 7 days) source of EF. (ESAR, Section5.5.7.3.1)4.1.2 10 CFR 50.6210 CFR 50.62 requires that pressurized water reactors have equipment diverse from the reactorprotection system to initiate the EF system under conditions indicative of an ATWS. The EFsystem is required to assure adequate removal of heat from the reactor coolant system duringan ATVVS.
Document Control DeskAttachment 1RC-1 6-0035CR-i16-00848Page 12 of 14The worst common mode failure which is postulated to occur is the failure to scram the reactorafter an anticipated transient has occurred. The effects of ATWVS events are not considered aspart of the design basis for transients analyzed in Chapter 15. The final NRC ATWS rulerequires that Westinghouse designed plants install ATWVS Mitigation System Actuation Circuitry(AMSAC) to initiate a turbine trip and actuate EF flow independent of the Reactor ProtectionSystem. The V. C. Summer AMSAC design is described in ESAR Section 7.8.4.2 PrecedentNone.4.3 No Significant Hazards Consideration1. Does the proposed amendment involve a significant increase in the probability orconsequences of an accident previously evaluated?No. A onetime change to the action statement of TS 3.7.1.2, ACTION b, does notincrease the probability or consequences of any analyzed accident addressed withinESAR Chapter 15. The EF system is not an initiator of any Chapter 15 accidents, andthe one-time change does not make it an initiator. Therefore, there cannot be anincrease in the probability of an accident previously evaluated.The relevant consequences stem from the ability to maintain core cooling. The changeis not detrimental to the ability to remove core heat, because while the maintenance isbeing performed affects the two MDEFPs, the TDEFP remains available for SteamGenerator cooling. A review of the Chapter 15 analyses shows that for single failureconsiderations, only one safety train is credited for accident mitigation. Events creditingEF flow assume one EF pump is able to deliver flow to the Steam Generators. This ispreserved by maintaining the availability and operability of the TDEFP. The only specificcircumstance in which TDEFP operation could be potentially affected is the occurrenceof a break of the Main Steam 4" branch line that supplies steam to the TDEFP. Sincethe activity does not involve a change to the main steam system, or otherwise affects theability of the main steam system to supply the TDEFWP, there cannot be an increase inthe probability of such a break. Nonetheless, in the unlikely event that the MDEFPflowpaths cannot be restored quickly because of that break, and with the area potentiallyinaccessible, core cooling can still be assured by initiating safety injection to establishfeed and bleed cooling, as di'rected by the Emergency Operating Procedures (EOPs).Failure to automatically isolate EF to the affected Steam Generator is an importantconsideration within two secondary pipe break analyses. For secondary side pipebreaks inside containment (FSAR Section 6.2), operator action at 30 minutes is creditedto isolate EF to the affected Steam Generator. Local or remote operator action within 30minutes is required to prevent overpressurizing the containment. Secondly, forsecondary side pipe breaks outside containment (FSAR Section 3.11.2.2.2.2 and Document Control DeskAttachment 1RC-1 6-0035CR-I16-00848Page 13 of 1410.4.9.3), credit is taken for operator action at 10 minutes to isolate EF to the affectedSteam Generator. Since the harsh environment will limit local access and manualactions, operator action from the control room is required for secondary pipe breaksoutside containment to preserve environment conditions for equipment qualification.Extending the action statement from six hours to 24 hours does not increase theprobability or consequences of an accident previously evaluated.2. Does the proposed amendment create the possibility of a new or different kind ofaccident from any accident previously evaluated?No. Extension of the action statement does not create the possibility of a new ordifferent kind of accident from any accident previously evaluated. In the case ofsecondary breaks outside the reactor building, which would make the flow control valvesinaccessible for local operation, procedural guidance outside of the EOPs directs theoperators to take alternative action (secure the MDEFPs) if the flow control valveassociated with the faulted Steam Generator cannot be closed from the control board.Increasing the duration of the allowed action from six hours to 24 hours does not resultin a new or different kind of accident.3. Does the proposed amendment involve a significant reduction in a margin ofsafety?No. The relevant margin of safety stems from the ability to maintain core cooling usingthe Steam Generators. As described previously, the continued operability of the TDEFPpreserves the core cooling function in the event of an emergency. The postulation of asingle failure is not required while in the LCO Action Time. Nonetheless, because of theavailability of safety injection and the ability to perform feed and bleed cooling, corecooling will be assured. Therefore, there will not be a significant reduction in the marginof safety.4.4 ConclusionIn conclusion, based on the considerations discussed above, (1) there is reasonable assurancethat the health and safety of the public will not be endangered by operation in the proposedmanner, (2) such activities will be conducted in compliance with the Commission's regulations,and (3) the issuance of the amendment will not be inimical to the common defense and securityor to the health and safety of the public.
Document Control DeskAttachment IRC-1 6-0035CR-I16-00848Page 14 of 145.0 Environmental considerationA review has determined that the proposed amendment would change a requirement withrespect to installation or use of a facility component located within the restricted area, as definedin 10 CFR 20, or would change an inspection or surveillance requirement. However, theproposed amendment does not involve (i) a significant hazards consideration, (ii) a significantchange in the types or a significant increase in the amounts of any effluents that may bereleased offsite, or (iii) a significant increase in individual or cumulative occupational radiationexposure. Accordingly, the proposed amendment meets the eligibility criterion for categoricalexclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), noenvironmental impact statement or environmental assessment need be prepared in connectionwith the proposed amendment.
 
==6.0 REFERENCES==
6.1 10CFR50, Appendix A6.2 FSAR Section 3.11.2.2.2.26.3 FSAR Section 5.5.7.3.16.4 FSAR Section 6.26.5 FSAR Section 7.86.6 FSAR Section 10.4.96.7 ESAR Chapter 156.8 VCSNS Drawing6.9 VCSNS Drawing6.10 RG 1.174General Design Criteria for Nuclear Power PlantsMain Steam Line Break Outside Containment EquipmentQualificationRESIDUAL HEAT REMOVAL SYSTEM, System Availabilityand ReliabilityContainment SystemsATWS MITIGATION SYSTEM ACTUATION CIRCUITRY(AMSAC)EMERGENCY FEEDWATER SYSTEMACCIDENT ANALYSESI MS-5O-I181, 1 MS-SO-I182 Fisher 657-ET Diaphragm ControlValve302-085 (ESAR Figure 10.4-16) Emergency Feedwater FlowDiagram,An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the LicensingBasis Document Control DeskAttachment 2RC-I16-0035CR-I16-00848Page 1 of 2VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1ATTACHMENT 2PROPOSED TECHNICAL SPECIFICATION CHANGE (MARK-UP)Proposed Technical Specification Changes SummaryPae Affected Bar # Description of Change -Reason for ChangeSection3/4 7-4 3.7.1.2. Add note for one time EXIGENTAction b action requirementsallowing 24 hours to be inat least HOT STANDBYdue to two inoperablepumps or flow paths Document Control DeskAttachment 2RC-1 6-0035CR-i16-00848Page 2 of 2PLANT SYSTEMSEMERGENCY FEEDWATER SYSTEMLIMITNG CONDITION FOR OPERATION3.7.1.2 At least three independent steam generator emergency feedwater pumps and flowpaths shall be OPERABLE with:a. Two motor-driven emergency feedwater pumps, each capable of being poweredfrom separate emergency busses, andb. One steam turbine driven emergency feedwater pump capable of being poweredfrom an OPERABLE steam supply system.APPLICABILITY: MODES 1, 2 and 3.ACTION:Ada Wihone eegnyfdatrpumpinerberstethreuemergency feedwater pumps to OPERABLE status withi or bein at least HOT STANDBY within the next 6 ho HOT SHUTDOWNwithin the following 6 b. ithtwoemeg _ate pumps inoperable, be in at least HOT STANDBYwithin 6 hours a in HOT SHUTDOWN within the following 6 hours.c. With three emergency feedwater pumps inoperable, immediately initiatecorrective action to restore at least one emergency feedwater pump toOPERABLE status as soon as possible.SURVEILLANCE REQUIREMENTS4.7.1.2 Each emergency feedwater pump shall be demonstrated OPERABLE:a. At least once per 31 days by:1. Verifying that each motor driven pump develops a total head of greater jthan or equal to 3800 feet at greater than or equal to 90 gpm flow.2. Verifying that the steam turbine driven pump develops a total head of Jgreater than or equal to 3140 feet at a flow of greater than or equal to 97gpm when the secondary steam supply pressure is greater than 865 psig.The provisions of Specification 4.0.4 are not applicable.3. Verifying that each non-automatic valve in the flow path that is not locked,sealed, or otherwise secured in position, is in its correct position.*The ACTION to be in at least HOT STANDBY in 6 hours is extended to 24 hours to test (and performremedial maintenance on) the motor driven emergency feedwater pump flow control valves per surveillancerequirement 4.7.1.2.c.2. This extension expires on March 18, 2016.SUMMER- UNIT 13/4 7-4Amendment No.
Document Control DeskAttachment 3RC-I16-0035CR-I16-00848Page 1 of 2VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1ATTACHMENT 3PROPOSED TECHNICAL SPECIFICATION CHANGE (RETYPED)Replace the following pages of the Technical Specifications with the attached revised pages.The revised pages are identified by amendment number and contain marginal lines indicatingthe areas of change.Remove Paqes3/4 7-4Insert Pages3/4 7-4 PLANT SYSTEMSEMERGENCY FEEDWATER SYSTEMLIMITING CONDITION FOR OPERATION3.7.1.2 At least three independent steam generator emergency feedwater pumps and flowpaths shall be OPERABLE with:a. Two motor-driven emergency feedwater pumps, each capable of being poweredfrom separate emergency busses, andb. One steam turbine driven emergency feedwater pump capable of being poweredfrom an OPERABLE steam supply system.APPLICABILITY: MODES 1, 2 and 3.ACTION:a. With one emergency feedwater pump inoperable, restore the requiredemergency feedwater pumps to OPERABLE status within 72 hours or bein at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWNwithin the following 6 hours.b. With two emergency feedwater pumps inoperable, be in at least HOT STANDBYwithin 6 hours* and in HOT SHUTDOWN within the following 6 hours.c. With three emergency feedwater pumps inoperable, immediately initiatecorrective action to restore at least one emergency feedwater pump toOPERABLE status as soon as possible.SURVEILLANCE REQUIREMENTS4.7.1.2 Each emergency feedwater pump shall be demonstrated OPERABLE:a. At least once per 31 days by:1. Verifying that each motor driven pump develops a total head of greaterthan or equal to 3800 feet at greater than or equal to 90 gpm flow.2. Verifying that the steam turbine driven pump develops a total head ofgreater than or equal to 3140 feet at a flow of greater than or equal to 97gpm when the secondary steam supply pressure is greater than 865 psig.The provisions of Specification 4.0.4 are not applicable.3. Verifying that each non-automatic valve in the flow path that is not locked,sealed, or otherwise secured in position, is in its correct position.* The ACTION to be in at least HOT STANDBY in 6 hours is extended to 24 hours to test (andperform remedial maintenance on) the motor driven emergency feedwater pump flow controlvalves per surveillance requirement 4.7.1.2.c.2. This extension expires on March 18, 2016.SUMMER- UNIT 13/4 7-4SUMMR-UIT I3/47-4Amendment No. 112, 111, 173, Document Control DeskAttachment 4RC-1 6-0035CR-I16-00848Page 1 of 1VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1ATTACHMENT 4LIST OF REGULATORY COMMITMENTSThere is one regulatory commitment created due to this License Amendment Request. Anyother statements in this submittal are provided for information purposes and are not consideredto be regulatory commitments. Please direct questions regarding these commitments toMr. Bruce L. Thompson at (803) 931-5042.Commitment Due DateLicense Amendment Request submitted to remove the note June 30, 2017permitting 24 hours to HOT STANDBY.}}

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