Text

(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
	ML16062A368
Person / Time
Site:	Summer
Issue date:	03/01/2016
From:	Lippard G; South Carolina Electric & Gas Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CR-16-00848, RC-16-0035
	Download: ML16062A368 (21)
	v • d • e

George A. Lippard Vice President, Nuclear Operations 803.345.4810

11 m,* **'~f'GMarch 1, 2016 A SCANA COMPANY RC-1 6-0035 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Dear Sir I Madam:

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 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 South Carolina Electric & Gas Company (SCE&G), acting for itself and as an agent for South Carolina Public Service Authority pursuant to 10 CFR 50.90 and 10 CFR 50.91, hereby submits a request for an exigent amendment to Technical Specifications (TS). The proposed amendment would modify the action statement for two inoperable pumps or flow paths within Section 3.7.1.2, "Plant Systems - Emergency Feedwater System."

Attachment I provides an evaluation of the proposed change to the action statement to amend the six hour action to be in at least HOT STANDBY to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to allow for maintenance and retesting. This amendment request was evaluated and found to have no significant hazards for consideration. An exigent TS change is justified in that compliance with TS could involve an unnecessary plant action to shutdown the reactor to COLD SHUTDOWN and potential reliance on the turbine driven emergency feedwater pump for plant cooldown without a corresponding health and safety benefit. The station proposes that the action statement be amended to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to allow for maintenance and retesting. Attachment 2 contains the marked-up version of the 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 provided to the designated South Carolina Official. This proposed change has been reviewed and approved by both the VCSNS Plant Safety Review Committee and the VCSNS Nuclear Safety Review 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 or require 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 Desk RC-1 6-0 035 CR-I16-00848 Page 2 of 2 I certify under penalty of perjury that the information contained herein is true and correct.

Executed on G ieo e A i~r WLT/GAL/

Attachments:

1. Analysis of Proposed Technical Specification Change

2. Proposed Changes - Marked Up TS Page

3. Proposed TS Pages - Retyped

4. Commitment Page c: K. B. Marsh S. A. Williams S. A. Byrne NRC Resident Inspector J. B. Archie K. M. Sutton N. S. Camns P. Ledbetter J. H. Hamilton S. E. Jenkins J. W. Williams NSRC W. M. Cherry RTS (CR-I16-00848)

C. Haney File (813.20)

PRSF (RC-16-0035)

Document Control Desk RC-1 6-0035 CR-I16-00848 Page 1 of 14 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS)

DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 Attachment 1

Analysis of Proposed Technical Specification Change

Subject:

This evaluation supports a request to amend South Carolina Electric & Gas Company (SCE&G), Technical Specifications (TS) to modify the action statement of 3.7.1.2, Emergency Feedwater System, Limiting Conditions For Operation (LCO) for two inoperable motor driven pumps.

1.0

SUMMARY

DESCRIPTION In 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 Nuclear Regulatory Commission (NRC) review and approval to amend Operating License NPF-12 for Virgil C. Summer Nuclear Station (VCSNS) Unit 1.

VCSNS is proposing an exigent TS change. It is the station's position that compliance with TS could involve an unnecessary plant shutdown and the potential reliance on the turbine driven emergency feedwater pump (TDEFP) for plant shutdown without a corresponding health and safety benefit. Due to an oversight, the station missed a surveillance test during the fall 2015 startup from refueling outage 22 (RF-22) associated with the emergency feedwater (EF) control valves in accordance within VCSNS Technical Specification 4.7.1.2.c.2. This surveillance requires verifying the flow control valves can be closed and held closed for three hours when normal instrument air is not available. The surveillance is normally conducted in Mode 4 or below when the Steam Generators are not relied on for heat removal. Due to the design configuration of the EF system, the six hour action statement b for two inoperable emergency feedwater pumps is entered anytime a motor driven emergency feedwater pump (MDEFP) flow control valve is closed in modes 1, 2 or 3. The station proposes to modify limiting conditions for operation 3.7.1.2 action statement b which currently requires: for two inoperable emergency feedwater pumps, be in at least HOT STANDBY within six hours and be in HOT SHUTDOWN within the following six hours. The station proposes that the action statement be amended to be in at least HOT STANDBY within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to allow for timely completion of any required maintenance and surveillance retest.

Document Control Desk RC-I16-0035 CR-I16-00848 Page 2 of 14 2.0 DETAILED DESCRIPTION Due to an oversight the station has missed performing a surveillance associated with the EF control 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 typically completed in HOT SHUTDOWN or below when the steam generators are not relied on for heat removal. This test requires the MDEFP flow control valves be held closed for 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months with air from the accumulators. Due to the design configuration of the EF system, the six hour action statement b for two inoperable EF pumps is entered anytime a MDEFP flow control valve is closed in modes 1, 2 or 3. With the test time period of three hours, no time is available to conduct remedial corrective maintenance and repeat the surveillance. This could result in an unnecessary plant shutdown.

The end 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 the General Operating Procedure (GOP-2) for Plant Startup and Heatup (MODE 5 to MODE 3), the surveillance was thought to be complete based on completion of surveillance for the TDEFP flow control valves. This error was not detected until the plant was in HOT SHUTDOWN at approximately 345 degrees Fahrenheit and was relying 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 action statement b to conduct the test. However, conducting the 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months surveillance test at power while in a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months shutdown action statement leaves no time to make repairs and conduct a retest. Changing the action statement to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months will allow for unforeseen corrective maintenance and subsequent retest would prevent the station from an unnecessary plant shutdown without a corresponding health and safety benefit.

2.1 Possible Repairs Timeline The following potential component failures could be required following the surveillance. The estimated repair times are based on repairing each item identified below and include tagging out the appropriate isolation devices. The time reflected also accounts for retesting of the EF control valves to ensure the capability to hold the valve closed for three hours as required by TS 4.7.1 .2.c.2.

Air accumulator check valve replacement - 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

Air pressure regulator rebuild and calibration - 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months .

Air actuator diaphragm casing bolts torque adjustment - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Air actuator diaphragm replacement - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Air solenoid valve replacement - 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .

Air relief valve replacement and setup - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Document Control Desk RC-1 6-0035 CR-i16-00848 Page 3 of 14 2.2 Bases for Exigent Change Surveillance Test Procedure (STP)-1 20.006, "Emergency Feedwater Valves Backup Air Supply Test," was not performed during the fall 2015 outage for the MDEFP flow control valves. The end 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)," this surveillance was signed off as being complete based on completion of the A-train portion of the test, which tests the TDEFP flow control valves, done under a separate task sheet from the one written for the B-train valves (MDEFP flow control valves). This error was not detected until performance of General Testing Procedure (GTP-702), "Surveillance Activity Tracking and Triggering," for Mode 3 entry, which lists the A-train and B-train tasks as separate line items. By that time, the plant was relying on MDEFP flow and the steam generators for heat removal as is normal during start-up. To perform the testing at this point in the outage would require the plant to 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 in Modes 1, 2, and 3 as long as both emergency diesel generators are operable with no maintenance 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 the plant back down to less than 183 degrees Fahrenheit and reinitiate RHR cooling, but instead to schedule 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 on February 26, 2016. During a normal process schedule review on January 30, 2016, it was discussed that this test would need additional focus to be performed online due to the short duration six hour action to HOT STANDBY required by TS 3.7.1.2 action b. Station personnel then began to apply additional planning considerations and focus to the testing including designating the test as an Infrequently Performed Test or Evolution (IPTE) and developing contingencies for repairs should valve repairs be required. After input by several plant groups the contingency matrix was finalized late on February 23, 2016. Based on the estimated times for repairs and retesting in the matrix, it was determined that a reasonable repair could not be applied within the actions specified in TS 3.7.1.2 action b. The test was rescheduled to be performed on March 11, 2016, to allow additional planning time. While the actual end date for surveillance is March 17, 2016, major maintenance has been scheduled on one of the emergency diesel generators for the week of March 13, 2016.

Document Control Desk Attachment I RC-1 6-0035 CR-16-00848 Page 4 of 14 While the station has prepared to conduct the test during Mode 1 by entering the TS 3.7.1.2 action statement b to conduct the test, conducting the 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months surveillance test at power while in a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months shutdown action statement leaves no time to make repairs and conduct a retest.

Changing the action statement to allow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to allow for unforeseen corrective maintenance and subsequent retest would prevent the station from an unnecessary plant shutdown without a corresponding health and safety benefit.

3.0 TECHNICAL EVALUATION

The EF system consists of three pumps, two motor driven and one steam turbine driven. The EF System is used to supply feedwater to the steam generators during startup, shutdown, and layup operations. A simplified system drawing is shown as Figure 1 where the full version can be 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 also required to be closed due to a faulted Steam Generator. The flow control valves are held closed for three hours by their associated air accumulators following an EF high-flow signal to a faulted Steam Generator. The three hour time permits automatic valve closure following a secondary system break when local valve operation cannot be accomplished because local conditions are unsuitable for personnel access. The valves are supported by safety class air accumulators with sufficient capacity to permit remote valve closure for at least three hours during a loss of instrument air system. The air accumulators provide a regulated air supply as needed to close the valves against spring force. The accumulators are supported by a non-safety instrument air system. Additionally, each valve has a handwheel to provide manual control.

3.1 System Operation The EF system "is required to deliver sufficient feedwater to the Steam Generators for cooldown upon 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 during startup, shutdown, and iayup operations. The EF system operates in conjunction with the turbine bypass system, if available, or the main steam power relief valves and safety valves, to remove 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 to establish this flow while sustaining a single active failure in the system in the short term or a single active or passive failure in the long term. The EF system operates until the RHR System can be placed in operation.

Document Control Desk RC-l16-0035 CR-I16-00848 Page 5 of 14 When forced circulation from the reactor coolant pumps is not available, EF operation is required down to a main steam pressure of 100 psia. This corresponds to a reactor coolant cold leg temperature of 325 degrees Fahrenheit and a hot leg temperature of 350 degrees Fahrenheit. The primary coolant temperature differential is required in order to maintain a density gradient to drive natural circulation of primary coolant, in the absence of reactor coolant pump operation.

Sufficient feedwater is available under emergency conditions to bring the plant to a safe shutdown 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 tank is 158,570 gallons based on maintaining the plant at hot standby conditions for 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months .

This volume also satisfies the minimum required volume to cool down the plant to HOT SHUTDOWN conditions assuming the plant is maintained at HOT STANDBY for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and then cooled down to HOT SHUTDOWN in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The system consists of three pumps, two motor driven and one steam turbine driven. The two motor driven pumps share a common discharge header that splits off into three branches. Each branch has a pneumatic flow control valve, which controls flow to its respective steam generator. The one steam turbine driven pump has a separate header from the motor driven pumps: This header splits off into three branches, which controls flow to its respective Steam Generator.

The three flow control valves for the turbine driven pump have control elements fed from A-train power. The three flow control valves for the motor driven pumps have control elements fed from B-train power.

During the performance of testing of the B-train flow control valves, the flow path from both of the motor driven pumps is disrupted. This is due to both pumps sharing a common discharge header prior to branching off to the three flow control valves.

The MDEFPs are powered from separate and independent safety related emergency diesel generator backed buses.

EF is a dual purpose system. During normal operation, the motor driven pumps are used during heatup and cooldown to supply feedwater to the steam generators for reactor coolant system temperature control. During emergency operation, all three pumps can provide feedwater to support reactor coolant system heat sink capabilities via the Steam Generators.

Document Control Desk RC-1 6-0035 CR-i16-00848 Page 6 of 14 EMERGENCY FEEDWATER SYSTEM Figure 1 Simplified System Drawing 3.2 Component Design The EF System design consists of two redundant trains: i.e., the motor driven pump train and the turbine driven pump train. The motor driven pump train is designed for use during normal plant conditions (i.e. startup, hot standby, and cooldown) and for emergency shutdown of the reactor. The turbine driven pump train is designed for use for emergency shutdown of the reactor. EF initiation arises from any of several types of signals, which may be generated in response to a variety of plant conditions. These initiation signals are generated in response to low Steam Generator levels, loss of main feedwater, low voltage on the essential electric power buses, a Safety Injection signal, and an ATWVS mitigation signal.

The control systems for the turbine driven and motor driven pump flow control valves are identical. Automatic valve opening signals are generated by the reactor protection and logic system and depend upon the given plant condition which will determine whether only the motor driven pump flow control valves open or the turbine driven pump valves open also. These valves will receive an open signal whenever their respective pumps receive an auto-start. The exception is the MDEFP flow control valves which do not receive an open signal when all three main feed pumps trip. If the flow control valves are in MANUAL control, the valves fully open in response to an automatic open signal.

Document Control Desk RC-1 6-0035 CR-I16-00848 Page 7 of 14 The EF system is designed with three flow control valves at the discharge of the TDEFP and three flow control valves at the discharges of the MDEFPs. The valves are required to control EF flow to the Steam Generators to maintain program level and to produce sufficient main steam to permit main feedwater pump turbine operation and for plant cooldown after main steam 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 related devices 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 supported by safety-class air accumulators with sufficient capacity to permit remote valve closure on a high-flow signal and maintain the valve closed for at least three hours during a loss of instrument air system. The air accumulators provide a regulated air supply as needed to close the valves against spring force. The accumulators are supported by a non-safety instrument air system. 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 is available to isolate the flow control valve to a faulted Steam Generator. The three hour supply permits automatic or remote manual valve closure following a secondary system break when local valve operation cannot be accomplished due to unsuitable conditions for personnel access in the Intermediate Building. Figure 2 is provided as a simplified sketch of the control air system.

b .......

Figure 2 Simplified Control Air System (VCSNS Drawing 817-056-001)

Document Control Desk Attachment I RC-1 6-0035 CR-I16-00848 Page 8 of 14 3.3 Component History The MDEFP flow control valves are subject to periodic stroke time testing to the open and closed positions under TS 4.0.5. Portions of the test circuit in the pressure drop tested under TS Surveillance Requirement 4.7.1 .2.c.2 are also monitored on a quarterly test frequency under TS Surveillance Requirement 4.7.1.2.b. The EF flow control valve is not closed during the TS surveillance requirement 4.7.1 .2.b quarterly test. However this test does provide assurance as to the leak tightness of a large portion of the circuit with the exception of the actuator and associated solenoid. No Condition Reports (CR) were found in recent history associated with the 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 surveillance indicates reliable performance. Problems with unacceptable leakage were encountered in 2010. CR-10-01427 documents pressure regulator relief for IFV03541-PR2-EF needed to be reset to restore acceptable leakage. CR-I10-03793 documents high but acceptable leakage for 1FV03551-CVI-EF, which was corrected by tightening fittings in the tested boundary.

The actuator diaphragms are replaced on an every third refueling (R03) frequency under preventative maintenance tasks. Preventative maintenance history shows the actuator diaphragms 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 drop test 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. CRs written against the flow control valve operator were reviewed. The most recent three CRs were written 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, or component (SSC) degradation.

Document Control Desk Attachment I RC-1 6-0035 CR-I16-00848 Page 9 of 14 Recent maintenance history for the MDEFP flow control valves indicates all were calibrated during 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 valves following this maintenance are documented under the following Work Orders: 1410934, 1410939, and 1410944. The seat, plug/stem of IFV03531-EF was also replaced under Work Order 1513005. Step 3 of this Work Order documents leak testing fittings associated with the valve after this maintenance.

3.4 PRA Insights The VCSNS PRA (Version 7B4) is the current model of record for internal events. The VCSNS PRA 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 for the 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 Individual Plant Examination (IPE) process. Since this model was finalized, there have been more than 30 updates, including minor modeling convention changes and data updates, as well as changes to incorporate significant plant modifications, including Chilled Water and Component Cooling Water system modifications, crediting an alternate AC power source, alternate cooling for charging pumps, and alternate seal injection.

SCE&G employs a multi-faceted approach for establishing and maintaining the technical adequacy and plant fidelity of the VCSNS PRA model. This approach includes both a proceduralized PRA maintenance and update process and the use of independent peer reviews. The findings and observations (F&Os) from the initial peer review and their resolutions, along with additional F&Os from the 2007 assessment, have been fully addressed and closed.

PRA model updates have been performed to address all the identified gaps, and the VCSNS PRA 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 as endorsed 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 than 1 .0E-04/yr and I1.0E-05/yr respectively. Limiting the increase to an incremental core damage probability (ICDP) of 5.0E-07 and incremental large early release probability (ILERP) to 5.0E-08 provides 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 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mission time when required to isolate a faulted or ruptured steam generator. These valves do not contribute to any PRA modeled initiating event. Therefore their failure to remain closed has no impact on the probability of occurrence of any initiator, including flooding, fires, or seismic

Document Control Desk RC-1 6-0035 CR-I16-00848 Page 10 of 14 events. Flooding and fire events cannot cause a faulted or ruptured Steam Generator so there is no need for the isolation function of these valves in mitigating those events and no increase in CDF or LERF. Large seismic events (greater than safe shutdown earthquake (SSE)) could possibly induce a Steam Generator fault or rupture but its probability in a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months window is small enough (<2.0E-07) that it has little impact on the risk insights. Therefore the impact of the EF flow control valves not remaining closed for the mission time on mitigation of core damage and large early release sequences can be conservatively estimated by taking all three of the subject valves (IFV-3531-EF, IFV-3541-EF, and IFV-3551-EF) out of service in the EOOS software using the at power internal events model. This is conservative because it simultaneously fails both their function to open and throttle EF flow to provide a heat sink as well as their function to isolate a faulted or ruptured steam generator. With these three valves taken out of service in EOOS, ODE increases by a factor of 67 to 2.08E-04/yr and LERF increases by a factor of 163 to 8.6E-06/yr. Conservatively ignoring the relatively small baseline CDF, these valves could be out of service for 5E-07*8760/2.08E-04=21.05hrs. Similarly for LERF, these valves could be out of service for 5.0E-08*8760/8.6E-06=50.93hrs. The ICDP is limiting and results 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 LCO time. A shutdown driven by a short TS action statement is more likely to result in a reactor trip than a controlled shutdown. The conditional core damage probability (CCDP) of a reactor trip is 5.3E-07. Therefore the ODE due to an accelerated shutdown is estimated to be on the order of 1 .0E-07 to I1.0E-08. Since the valve testing and any needed repairs are expected to be completed in less than the requested 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months LCO time, the outage time estimated risk and the shutdown 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 valves associated with the steam driven emergency feedwater pump and by use of the charging/safety injection pumps in the feed and bleed mode. Defense in depth for the function to isolate a faulted 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 a postulated loss of EF flow associated with a steam line break outside containment (SLBO) in the supply lines for the TDEFP is by the use of charging/safety injection pumps in the feed and bleed mode.

Conservatively meeting the RG 1.174 (Reference 6.10) limits for risk increases and adding compensatory measures ensures sufficient safety margin to account for analysis and data uncertainties.

The following compensatory measures will be taken: Both emergency diesel generators will be verified available (not in Removal and Restoration Log), the TDEFP will be placarded and its room locked, a dedicated operator will be stationed locally to manually operate the flow control valves as required, the weather forecast will be reviewed for sever conditions (hurricane or tornado), and no other planned maintenance or testing will be in progress prior to entering the action statement.

Document Control Desk Attachment I RC-1 6-0035 CR-i16-00848 Page 11 of 14

4.0 REGULATORY EVALUATION

The EF system automatically supplies feedwater to the Steam Generators to remove decay heat from the reactor coolant system upon the loss of normal feedwater supply. The Steam Generators function as a heat sink for core decay heat. The heat load is dissipated by releasing steam to the atmosphere from the Steam Generators via the main steam safety valves. The EF system consists of two motor driven pumps and one steam turbine driven pump configured into three flow paths to supply three Steam Generators by common headers. The EF system is considered OPERABLE when the components and flow paths required to provide redundant EF flow to the steam generators are OPERABLE. This requires that the two MDEFP be OPERABLE with two diverse paths, each supplying EF to separate Steam Generators. The TDEFP is required to be OPERABLE with redundant steam supplies from each of two main steam lines upstream of the main steam isolation valves and shall be capable of supplying EF to any of the three Steam Generators. The piping, valves, instrumentation, and controls in the required flow paths also are required to be OPERABLE.

4.1 Applicable Regulatory Requirements I Criteria 4.1.1 GDC 34 General Design Criteria (GDC) 34 establish the requirements to assure the capability to transfer heat from the reactor to a heat sink under normal and accident conditions with sufficient redundancy and isolation capability to accomplish the safety function with a single failure of an active 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, as it is for all other Westinghouse designed pressurized water reactors. HOT STANDBY is a safe and stable plant condition which can be maintained for an extended period of time following any Condition II, Ill, or IV event. In the HOT STANDBY condition, residual heat removal, in compliance with GDC 34 (10OCFR5O, Appendix A), is provided by the EF system in conjunction with the Steam Generator safety valves. Cross connections from the service water system to the EF system provide a long term (i.e., greater than 7 days) source of EF. (ESAR, Section 5.5.7.3.1) 4.1.2 10 CFR 50.62 10 CFR 50.62 requires that pressurized water reactors have equipment diverse from the reactor protection system to initiate the EF system under conditions indicative of an ATWS. The EF system is required to assure adequate removal of heat from the reactor coolant system during an ATVVS.

Document Control Desk RC-1 6-0035 CR-i16-00848 Page 12 of 14 The worst common mode failure which is postulated to occur is the failure to scram the reactor after an anticipated transient has occurred. The effects of ATWVS events are not considered as part of the design basis for transients analyzed in Chapter 15. The final NRC ATWS rule requires that Westinghouse designed plants install ATWVS Mitigation System Actuation Circuitry (AMSAC) to initiate a turbine trip and actuate EF flow independent of the Reactor Protection System. The V. C. Summer AMSAC design is described in ESAR Section 7.8.

4.2 Precedent None.

4.3 No Significant Hazards Consideration

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

No. A onetime change to the action statement of TS 3.7.1.2, ACTION b, does not increase the probability or consequences of any analyzed accident addressed within ESAR Chapter 15. The EF system is not an initiator of any Chapter 15 accidents, and the one-time change does not make it an initiator. Therefore, there cannot be an increase in the probability of an accident previously evaluated.

The relevant consequences stem from the ability to maintain core cooling. The change is not detrimental to the ability to remove core heat, because while the maintenance is being performed affects the two MDEFPs, the TDEFP remains available for Steam Generator cooling. A review of the Chapter 15 analyses shows that for single failure considerations, only one safety train is credited for accident mitigation. Events crediting EF flow assume one EF pump is able to deliver flow to the Steam Generators. This is preserved by maintaining the availability and operability of the TDEFP. The only specific circumstance in which TDEFP operation could be potentially affected is the occurrence of a break of the Main Steam 4" branch line that supplies steam to the TDEFP. Since the activity does not involve a change to the main steam system, or otherwise affects the ability of the main steam system to supply the TDEFWP, there cannot be an increase in the probability of such a break. Nonetheless, in the unlikely event that the MDEFP flowpaths cannot be restored quickly because of that break, and with the area potentially inaccessible, core cooling can still be assured by initiating safety injection to establish feed and bleed cooling, as di'rected by the Emergency Operating Procedures (EOPs).

Failure to automatically isolate EF to the affected Steam Generator is an important consideration within two secondary pipe break analyses. For secondary side pipe breaks inside containment (FSAR Section 6.2), operator action at 30 minutes is credited to isolate EF to the affected Steam Generator. Local or remote operator action within 30 minutes is required to prevent overpressurizing the containment. Secondly, for secondary side pipe breaks outside containment (FSAR Section 3.11.2.2.2.2 and

Document Control Desk RC-1 6-0035 CR-I16-00848 Page 13 of 14 10.4.9.3), credit is taken for operator action at 10 minutes to isolate EF to the affected Steam Generator. Since the harsh environment will limit local access and manual actions, operator action from the control room is required for secondary pipe breaks outside containment to preserve environment conditions for equipment qualification.

Extending the action statement from six hours to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months does not increase the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

No. Extension of the action statement does not create the possibility of a new or different kind of accident from any accident previously evaluated. In the case of secondary breaks outside the reactor building, which would make the flow control valves inaccessible for local operation, procedural guidance outside of the EOPs directs the operators to take alternative action (secure the MDEFPs) if the flow control valve associated with the faulted Steam Generator cannot be closed from the control board.

Increasing the duration of the allowed action from six hours to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months does not result in a new or different kind of accident.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

No. The relevant margin of safety stems from the ability to maintain core cooling using the Steam Generators. As described previously, the continued operability of the TDEFP preserves the core cooling function in the event of an emergency. The postulation of a single failure is not required while in the LCO Action Time. Nonetheless, because of the availability of safety injection and the ability to perform feed and bleed cooling, core cooling will be assured. Therefore, there will not be a significant reduction in the margin of safety.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (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 security or to the health and safety of the public.

Document Control Desk Attachment I RC-1 6-0035 CR-I16-00848 Page 14 of 14 5.0 Environmental consideration A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 10CFR50, Appendix A General Design Criteria for Nuclear Power Plants 6.2 FSAR Section 3.11.2.2.2.2 Main Steam Line Break Outside Containment Equipment Qualification 6.3 FSAR Section 5.5.7.3.1 RESIDUAL HEAT REMOVAL SYSTEM, System Availability and Reliability 6.4 FSAR Section 6.2 Containment Systems 6.5 FSAR Section 7.8 ATWS MITIGATION SYSTEM ACTUATION CIRCUITRY (AMSAC) 6.6 FSAR Section 10.4.9 EMERGENCY FEEDWATER SYSTEM 6.7 ESAR Chapter 15 ACCIDENT ANALYSES 6.8 VCSNS Drawing I MS-5O-I181, 1 MS-SO-I182 Fisher 657-ET Diaphragm Control Valve 6.9 VCSNS Drawing 302-085 (ESAR Figure 10.4-16) Emergency Feedwater Flow Diagram, 6.10 RG 1.174 An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis

Document Control Desk RC-I16-0035 CR-I16-00848 Page 1 of 2 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 2 PROPOSED TECHNICAL SPECIFICATION CHANGE (MARK-UP)

Proposed Technical Specification Changes Summary Pae Affected Bar # Description of Change -Reason for Change Section 3/4 7-4 3.7.1.2. Add note for one time EXIGENT Action b action requirements allowing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to be in at least HOT STANDBY due to two inoperable pumps or flow paths

Document Control Desk Attachment 2 RC-1 6-0035 CR-i16-00848 Page 2 of 2 PLANT SYSTEMS EMERGENCY FEEDWATER SYSTEM LIMITNG CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator emergency feedwater pumps and flow paths shall be OPERABLE with:

a. Two motor-driven emergency feedwater pumps, each capable of being powered from separate emergency busses, and

b. One steam turbine driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:Ad a Wihone eegnyfdatrpumpinerberstethreu emergency feedwater pumps to OPERABLE status withi ,..* or be in at least HOT STANDBY within the next 6 ho HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .,,'*

b.ithtwoemeg _ate pumps inoperable, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months a in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three emergency feedwater pumps inoperable, immediately initiate corrective action to restore at least one emergency feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2 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 j than 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 J greater than or equal to 3140 feet at a flow of greater than or equal to 97 gpm 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 inthe flow path that is not locked, sealed, or otherwise secured inposition, is in its correct position.

The ACTION to be in at least HOT STANDBY in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is extended to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to test (and perform remedial maintenance on) the motor driven emergency feedwater pump flow control valves per surveillance requirement 4.7.1.2.c.2. This extension expires on March 18, 2016.

SUMMER- UNIT 1 3/4 7-4 Amendment No. *,*'*

Document Control Desk RC-I16-0035 CR-I16-00848 Page 1 of 2 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 3 PROPOSED 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 indicating the areas of change.

Remove Paqes Insert Pages 3/4 7-4 3/4 7-4

PLANT SYSTEMS EMERGENCY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator emergency feedwater pumps and flow paths shall be OPERABLE with:

a. Two motor-driven emergency feedwater pumps, each capable of being powered from separate emergency busses, and

b. One steam turbine driven emergency feedwater pump capable of being powered from an OPERABLE steam supply system.

APPLICABILITY: MODES 1, 2 and 3.

ACTION:

a. With one emergency feedwater pump inoperable, restore the required emergency feedwater pumps to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With two emergency feedwater pumps inoperable, be in at least HOT STANDBY within 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s* and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With three emergency feedwater pumps inoperable, immediately initiate corrective action to restore at least one emergency feedwater pump to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.7.1.2 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 than 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 greater than or equal to 3140 feet at a flow of greater than or equal to 97 gpm 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 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is extended to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to test (and perform remedial maintenance on) the motor driven emergency feedwater pump flow control valves per surveillance requirement 4.7.1.2.c.2. This extension expires on March 18, 2016.

SUMMER- UNIT 1 3/4 7-4 SUMMR-UIT I3/47-4Amendment No. 112, 111, 173,

Document Control Desk RC-1 6-0035 CR-I16-00848 Page 1 of 1 VIRGIL C. SUMMER NUCLEAR STATION (VCSNS) UNIT 1 ATTACHMENT 4 LIST OF REGULATORY COMMITMENTS There is one regulatory commitment created due to this License Amendment Request. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please direct questions regarding these commitments to Mr. Bruce L. Thompson at (803) 931-5042.

Commitment Due Date License Amendment Request submitted to remove the note June 30, 2017 permitting 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to HOT STANDBY.

George A. Lippard Vice President, Nuclear Operations 803.345.4810

11 m,* **'~f'GMarch 1, 2016 A SCANA COMPANY RC-1 6-0035 Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555