ML20095A689

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Proposed Tech Specs Table 3.3-5, ESFs Response Times
ML20095A689
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 04/15/1992
From:
SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML20095A688 List:
References
NUDOCS 9204200133
Download: ML20095A689 (19)
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l Enclosure 1 to Document Control Desk Letter TSP 890022 Page 1 PROPOSED TECHNICAL SPECIFICATION CHANGE - 1SP 890022 VIRGIL C. SUMMER NUCLEAP. STATION LIST OF AFFECTED PAGES i

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3/4 3 29 3/4 3-30 3/4 3-31 '

3/4 3 33 3/4 3-34 .

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l 9204200133 920415 L PDR .ADOCK 05000395 PDR l

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INSTRUMENTATION TABLE 3.3 5

. t ENGINEERED SAFEH FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 1.- Manual'

a. Snfety Injection (ECCS) Not Applicable
b. Raactor Building Spray Not Applicable
c. Containment Isolation Phase "A" Isolation Not Applicable
d. Steam Line Isolation Not Applicable
e. 'Feedwater Isolation Not Applicable
f. Emergency Feedwater Not Applicable
g. Essential Service Water Not Applicable
h. Reactor Building Cooling Fans Not Applicable
i. Control Room Isolation Not Applicable 2.- Reactor Building Pressore-High
a. Safety. Injection (ECCS) . 14hDO r! /.17.o
b. -Reactor Trip (from SI)' 5 3.0
c. Feedwater. Isolation 1 10.0
d. Contain.aent Isolation-Phase "A" 1 45.0(4)/55.0(5) l-SIM4ER - UNIT 1 3/4 3-29

- .. a . .- - . - _ _.. - _. - .- . - . - ----_--.__--..:

INSTRUMENTATION TABLE 3,3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TlHES INITIATING SIGNAL AND FUNCTION RESPONSE TlHE IN SECONDS

e. Reactor Building Purge and Exhausc Isolation Not Applicable
f. Emergency Feedwater Pumps Not Applicable
g. Service Water System 71.5(41/81.5(5)
h. Reactor Building Cooling Units 76.5(4)/86.5(b)
1. Control Room Isolation Not Applicable
3. Pressurizer Pressure-Low
a. Safety Injection (ECCS) <-1216 E72714 - @ rl /.n
b. Reactor Trip (from SI) < 3.0
c. Feedwater Isolation < 10.0
d. Containment Isolation-Phase "A" [ 45.0(4)/55.0(5)
e. Reactor Building Purge and Exhaust Isolation Not pplicable
f. Emergency Feedwater Pumps Not Applicable
g. Service Water System < 71.5(4)/81.5(5)
h. Reactor Building Cooling Units [ 76.5(4)/86.5(5)
1. Control Room Isolation Not Applicable
4. Differential Pressure Between Sto.am Lines High
a. Safety Injection (ECCS) <.1 M N 7d E b @ n

,, 37.0

b. Reactor Trip (from SI) < 3.0
c. Feedwater Isolation < 10.0
d. Containment Isolation-Phase "A" h 45.0(4)/55.0(5)

SUMER - UNIT 1 3/4 3-30 Amendment No. 67

INSVRUMENTATION TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSF TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

e. Reactor Building Purge and Exhaust Isolation Not Applicable
f. Emergency Feedwater Pumps Not Applicable
g. Service Water System < 71.5(4)/81.5(5)
h. Reactor BuilrHng Cooling Units [ 76.5(4)/86.5(5)
1. Control Room Isolation Not Applicable
5. Steam Line Pres @ e-Low i bb
a. Safety Injection - ECCS d-12r6[2)pg(3) 27.Q:b 37.o
b. Reactor Trip (from SI) 1 3.0
c. Feedwater Isolation < 10.0
d. Containment Isolation - Phase "A" h45.0(4)/55.0(5)
e. Reactor Building and Purge and Exhaust Iselation Not Applicable
f. Emergency feedwater Pumps Not Applicable
g. Service Water System ( 71.5(4)/81.5(5)
h. Reactor Building Cooling Units [76.5(4)/86.5(5)
1. Steam Line Isolation i 10.0 J. Control Room Isolation Not Applicable 6, Steam Flow in Two Steam Lines - High Coincident with T,yg--Low-Low
a. Steam Line Isolation 1 12.0
7. Reactor Building Pressure-High-2
a. Steam Line Isolation 5 9.0 SUP94ER - UNIT 1 3/4 3-31 Amendment No.67

.i .

INSTRUMENTATION

~

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES J u1TT. ATieJ(s SIGNAL Ant Nuer iod gf_ f,9e Ng vgic ps fecp4L5

13. Trip of Main Feedwater Pumps
a. Motor-dn .en Emergency Not Applicable feedwater Pumps
14. Loss of Power
a. 7.2 kv Emergency Bus 5 10.3 Undervoltage (Loss of Voltage)
b. 7.2 kv Emergency Bus 5 13.3 Undervoltage (Degraded Voltage) 15, OntainmentRadioactivity-Nich
a. Purge and Exhaust Isolation Not Applicable
16. RWST level low +1ow
a. Automatic Switchover to Not Applicable Containment Sump l
17. AUX FEED SUCTICN PRESSURE LOW
a. Suction transfer Not Applicable l

l fiote: Response time for Motor- 1 60.0 driven Emergency Feedwater Pumps on all S.I. signal starts SUMMER - UNIT 1 3/4 3-33 l

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. 1 1

INSTRUMENTATION TABLE 3.3 5 (Continued)

TABLE NDTATION ,__

6pIcb7,U (1) Diesel generator starting and seovence loading delays from under voltage included. Response tirie limit includes oper.irt of valves to establish $1 path and attainment of disenarge pressure f or centrifugal charging pumps _

Laert A->and RHR pumps, ykif<osa,0, (2) Diesel generator starting delay not included. Seouence loading delay I r

included. Off site peer availarE Response time limit incluoes +oentag of valves to establish $1 path and attainment of discharge pressure for

- centrifigal charging pumps

  • 3mp[ g (gse gjeney (3) Diesel generator starting and sequence loading / delays from under voltage included. Response time limit includes s>oen4sg of vales to establish $1 path and attainment of disenarge pressure f or centrifugal charging pumps.

ZW.W[ f>' - r>

' (4)~ Diesel generator starting delay not included, Seouence loading delay included, Offsite power availab E (5) Diesel generator starting and secuence leading delays from undervoltage included.

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1 SU+ER UNIT 1 3/4 3-34

- [NSTRUMENTATION 1AELf__33 3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES JNITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDj

1. Manual

-a. Safety injection Not Applicable

b. Reactor Building Snray Not Applicable
c. Containment Isolation Not Applicable Phase "A" Isolation d.- Steam Line Isolation Not Aaolicable e.- feedwater Isolation Not Applicable *
f. Emergency feedwater Not Applicable
g. Essential Service Water Not Applicable
h. Reactor Building Cooling fans Not Applicable
1. Control Room Isolation Not Applicable  ;
2. Reactor Buildina Pressure-Hiah ,
a. SafetyInjection(ECCS) $27.0(2)/27.0(1)
b. PeactorTrip(fromSI) 53.0
c. feedwater Isolation $10.0
d. Containment Isolation-Phase "A" $45.0(4)/55.0(5) s i.

SUMMER - UNIT-1 3/4 3-29

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INSTRUMENTATION TABLE 3.3-5 (Continue #1 l ENGINEERED SAFETY FEATURES RESPONSE TIM [ji j LNITI ATING SIGNAL, AND FUNEl.0J MSPONSE 11ME IN $ ECON 05 l

e. Reactor Building Purge and Not Applicable ,

Exhaust isolation i

f. -Emergency feedwater Pumps Not Applicable
g. Service Water System 71.5(4)/81.5(5)
h. Reactor Building Cooling Units 76.5(4)/86.5(5)
1. Control Room isolation Not Applicable

?

3. Pressurizer Pressure-tow
a. Safety _ Injection (ECCS) 527.0(2)/27.0(1)
b. Reactor Trip (from SI) 53.0 t
c. Feedwater Isolation $10.0
d. Containment Isolation -Phase "A" $45.0(4)/55.0(5) i
e. Reactor Building Purge and Not Applicable Exhaust Isolation
f. Emergency feedwater Pumps Not Applicable
g. Service Water System 71.5(4)/81.5(5)
h. Reactor Building Cooling Units 76.5(4)/86.5(5)
1. Control F;oom Isolation Not Applicable
4. Differential Pressure Between Steam Lines-Hiah
a. Safetyinjection(ECCS) 527.0(2)/37.0(3)

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-b. Reactor Trip (from SI) 53.0

c. Feedwater Isolation $10.0
d. Containment Isolation -Phase "A" $45.0(4)/55.0(5)

SUMMER - UNIT 1 3/4 3-30

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. - INSTRUMEtjiRQ.!!

TABLE 3.3-5 (Continued 1 1

ENGINEEREp SAFETY FEATURES RESPONSE TIMES INITIATING $1GNAL ANp FUNCTION RESPONSE TIME IN SECONDS

e. Reactor Building Purge and Not Applicable
  • Exhaust Isolation
f. Emergency Feedwater Pumps Not Applicable
g. Service Water System 571.5(4)/81.5(5) i h. Reactor Building Cooling Units $76.5(4)/86.5(5) ,
i. Control Room Isolation Not Applicable
5. Sig m Line Pressure-Low
a. Safety Injection - ECCS 527.0(2)/37.0(3)
b. Reactor Trip (from SI) 53.0
c. Feedwater Isolation- $10.0
d. Containment Isolation - Phase "A" 545.0(4)/55.0(5)
e. Reactor Building and Purge and-Exhaust isolation Not Applicable
f. Emergency feedwater Pumps Not Applicable 9 Service Water System 571.5(4)/81.5(5)
h. Reactor Building Cooling Units 576.5(4)/86.5(5) ,
1. Steam Line Isolation $10.0
j. Control Room Isolation Not Applicable i.

- 6._ Steam Flow in Two Steam Lines - Hich Coincident with Tavg--Low-low

a. Steam Line Isolation 512.0

- 7. Reactor Buildina Pressure-Hich-2

a. Steam Line Isolation 59.0 SUMMER - UNIT 1 3/4 3-31 i

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. INS'TRUMENTATlp!!

i TABLE 3.3-5 (Continued)

ENGINEERED _ SAFETY FEATURES RESP (LNSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

13. Trio of Main feedwater Pumpi ,

a._ Motor-driven Emergency Not Applicable feedwater Pumps

14. Loss of Power
a. 7.2 kv Emergency Bus $10.3 undervoltage(Lossof Voltage)
b. 7.2 kv Emergency Bus $13.3 l Undervoltage(Degraded Voltage)
15. Containment Rsjioaciivity--Hich
a. Purge and Exhaust Isolation Not. Applicable
16. RWST Level--Low-Low
a. Automatic Switchover to Not Applicable Containment Sump

'17. -Aux Feed Suction Pressure low I

a. Suction transfer Not Applicable 1

Note: Response time for Motor- 160.0

,- driven Emergency Feedwater

, Pumps on all SI signal starts SUMMER - UNIT 1 3/4 3-33 .

. .= _ _ ... _ _ .. . _ _ .___, . . - . _ . _ . _ . . . _ _ _._..._..___.._.___:.__.

. . _ _. _ _ .. _ .._._ _ _ _ ._ _ _ ..- ___ - _ .. _ _ __. _ ._ .~ ._ - . _ ._-

, JNSTRUMENTATION i

1ABLE 3.3-5_fContinued) ,

6 TABLE NOTATION (1) Diesel generator starting and sequence loading delays from under voltage included. Response time limit includes positioning of valves to establish 51 path and attainment cf discharge pressure for centrifugal charging pumps and RHR pumps. Sequtntial transfer of ,

-centrifugal charging pump suction from the VCT to the RWS1 (RWST valves open, then VCT. valves close) is not included.

(2) Diesel generator starting delay not included. Sequence loading delay included. Offsite power available. Response time limit includes positioning of valves to establish SI path and attainment of '

discharge pressure for centrifugal charging pumps. Sequential

-transfer of-centrifugal charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is included.

. (3) Diesel generator starting and sequence loading delays from urder voltage included. Response time limit includes positioning of valves to establish 51 path and attainment of discharge pressure for centrifugal charging pumps Sequential transfer of centrifugal charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is. included.

(4) Diesel generator starting delay nel included. Sequence loading delay ,

included. Offsite power available, ,

(5) Diesel generator starting and sequence loading delays from undervoltage included.

I SUMMER -UNIT l' 3/4 3 34 L

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Enclosure 2 to Document Control Desk Letter TSP 890022 Page 1 PR01OSED TECHNICAL SPECIFICATION CHANGE - TSP 890022 VIRGIL. C. SUMMER NUCLEAR STATION DESCRIPTION AND SAFETY EVALVATION DESCRIPTION Of AMENDMENT REQUEST SCE&G proposes to modify the VCSNS TS to revise TS 3/4.3.2. Tabic 3.3-5,

" Engineered Safety Features Response Times," to reflect the closing time associated with the Volume Control Tank outlet isolation valves, subsequent to a safety injection signal ("S").

Technical Specifications Table 3.3-5, " Engineered Safety features Response Times," provides the time interval from when a monitored Engineered Safety Features-(ESF) parameter exceeds its actuation setpoint at the channel sensor untti the ESF equipment is capable of performing its intended safety.

function. These response times are measured at least once per eighteen months to verify that the ESF actuation associated with each channel is completed within the time limit specified in the tabic. This verification assures that the assumptions used for the Loss of Coolant Accident (LOCA) and non-LOCA accident analyses remain valid.

Inthenormal.configurationoftheChemicalandVolumeControlSystem(CVCS),

the high-head safety injection outps (charging pumps) take suction from the VolumeControlTank(VCT). When an "S" signal is generated from the protection logic, a signal it sent to >+ art the high-head safety injection pumps (charging pumps), and to open the Refueling Water Storage Tank (RWST) outlet isolation valves to align the borated water source for injectim to the Reactor Coolant System (RCS). Once the RWST outlet isolation valves indicate fully opened, the outlet isolation valves on the VCT begin to close.

The sequential valve stroke time of the RWST and the VCT valves can be as long as 25 seconds and is designed to ensure that a net positive suction head is maintained on the suction of the pumps. The hydrogen pressurized VCT is the source of safety injection (SI) flow until the VCT outlet isolation valves ar': closed, and the suction is changed to the RWST. The VCT outiet isolat' z, valve closure time affects the time assumed at which the borated water in the RWST is available to the suction of the high-head safety injection pumps.

Current TS times only address the opening of RWST outlet isolation valves to maintain a flow path and not the closing of the VCT outlet isolation valves.

The proposed TS change will increase the allowable response times in Table 3.3-5 to incorporate the additional time required for the sequential-stroking of-the RWST and VCT valves.

i Enclosure 2 to Document Control Desk Letter '

TSP 890022 Page 2 SAFETY EVALVATION LARGE BREAK LOCA FSAR SECTION 15.4.1 Large break LOCA an9 yses are performed under the assumption that the '

immediate safety funt tion of the SI System is to supply water to the RCS.

The time at which water (from either the YCT or the RWST) is available to the suction of the high-head safety injection pumps depends on the "S" signal generation time and the time delay for the pumps to attain full coced. This time, the time at which water is available to the suction of the g mps, will not be affected by the time delay for closure of the VCT outlet isolation valves, since the valve alignment for the supply of borated water to the RCS is not considered in the analyses. Although the pressure at the charging pump inlet will be higher when the VCT valves are open, the charging pump flow rate will not be degrade 6. Also, negative reactivity insertion due to core voiding causes the nuclear chain reaction to stop and reduce the core power and decay heat levels without reliance on the injected fluid boron concentration.

Therefore, for 51 actuation signals intended to provide protection against a LOCA, the additional delay for injection of borated water does not have to be considered since boron is only required to maintain subcrittet:11ty in the long term following a LOCA.

From the above discussion it can be concluded that a delay in borated water injection will have no impact on the VCSNS large break LOCA analysis results.

Consequently, current margins to 10CFR50.46 criteria are not decreased.

SMALL BREAK LOCA - FSAR SECTION 15.3.1 Small break LOCA analyses are performed under the assumption that the immediate safety functinn of the SI system is to supply water to the RCS.

This is similar to the large break LOCA analyses assumption. The small break LOCA analysis assumes that shutdown of the reactor core is achieved by insertion of all but the most reactive of the rod control cluster assemblies, and no credit is taken for the boron concentration of the SI flow.

Therefore, as indicated above for the SI actuation signals intended only to provide protection against a LOCA,-the additional delay for injection of borated water will not affect the small break LOCA analysis results.

Boron is only required for maintaining subcriticality in the long term following a LOCA.

From the above discussion, it can be concluded.that a delay in borated water injection will have no impact on the VCSNS small break analysis results.

Consequently, current margins to 10CFR50.46 criteria are not decreased.

l

Enclosure 2 to Document Control Desk tetter TSP 890022 Page 3 ROD EJECTION MASS AND ENERGY RELEASE FOR DOSE CALCULATIONS -

FSAR SECTION 15.4.6 TFe delay in the borated water injection has negligible impact on the Rod Ejection accident analysis $1nce the S1 flow to the RCS is modeled under similar assumptions as in the large break and small break LOCA analyses. The current FSAR Rod Ejection accident analysis for VCSNS was performed with the WFLASH Evaluation Model. The impact on the ftAR Rod Ejection accident will be negligible for reasons discussed under the 60CA analyses.

CONTAINMENT INTEGRITY - (SHORT AND LONG TERM MASS AND ENERGY RELEASE) -

FSAR SECTION 6.2 _

FSAR section 6.2 considers the containment subcompartments, mass and energy, for postulated LOCAs and containmen+ heat removal systems, for the containment subcompartment analyses, a delay in the injection of borated water has no impact on the calculated results, since the short duration of the transient (<3 seconds) does not consider any 51 flaw to the RCS.

Therefore a delay in the injection of borated water would have no impact on the long term mass and energy releases calculated for VCSNS.

STEAM GENERATOR TUBE RUPluRE - FSAR SECTION 15.4.3 for the Steam Generator Tube Rupture (SGTR) accident, primary to secondary break flow was assumed to be terminated at 30 minutes after initiation of the SGTR event, and operator recovery action to cool down the RCS was not modeled 3

in the analysis. Without RCS cooldown, sufficient shutdown margin is assumed to be available initially, and maintained for the long term by borated water.

The increase in delay time for injection of borated RWST water will not change the assumption regarding the maintenance of the long term shutdown margin. Therefore, the additional delay for injection of borated water would -

have no impact on the SGTR cnalysis for VCSNS.

BLOWDOWN REACTOR VESSEL ,JD LOOP FORCES - FSAR SECTION 3.9.3 The blowdown hydraulic loads resulting from a LOCA are considered in Sect'on 3.9.3 (Reactor 'iessel Loss of Coolant Accident Analysis and Dynamic Analysis of Reactor Internals Under faulted Conditions) of the VCSNS FSAR. The increase in delay time until borated water is available, as a result of the VCT/RWST valve interlock logic, will not affect the LOCA blowdown hydraulic loads since the maximum loads are generated within the first few seconds after break initiation, for this reason the ECCS and associated valve interlock logic are not considered in the LOCA hydraulic forces modeling, and thus the additional water delivered from the VCT during the switchover to suction from the RWST does not affect the results of the LOCA hydraulic forces calculation.

l I

Enclosure 2 to Document Control Desk Letter TSP 890022 Page 4 i l

1 POST LOCA LONGTERM COR~ COOLING - FSAR SECTION 15.4.1 ,

l VCSNS licensing position for satisfying the requirements of 10 CfR Part 50, 1 Section50.46, Paragraph (b), item (5),"LongTermCooling,"isdefinedin WCAP-8339. The Westinghouse Evaluation Model commitment is that the reactor will remain shutdown by borated ECCS water residing in the sump post LOCA.

Since credit for the control rods is not taker for a large break t0CA, the borated ECCS water provided by the accumulators and t, RWST must have a boron concentration that, when mixed with other water sources, will result in the reactor core remaining suberitical, assuming all control rods out.

In the normal configuration of the CVCS, the high-head safety injection pumps take suction from the VCT. The VCT is pressurized and serves as a source of Si flow until its outlet isolation valves close. Since the delay between the time RWST valves begin to open--on an "S" sigt.al--and the VCT valves are closed can be as much as 25 seconds, it is appropriate to conservatively assume that, for 25 seconds, delivery of non-borated water occurs before introduction of borated water from the RWST.

The effect of injecting additional non-borated water into the RCS during the switchover to the suction from the RWST has been considered with respect to  ;

the long term core cooling evaluation. Assuming maximw flow from all high-head safety injection pumps, the water from the VCT that could make its way to the RCS and the sump has been estimated at about 3363 lbs. Adding this inventory as non borated water in the boron evaluation, the indicated sump boron concentration would be reduced by only 2.1 ppm (maximum). It is i' concluded that the amount of additional non-borated water does not significantly reduce-the sump boron average concentration; therefore, it does not affect the ability of the core to remain shutdown by borated ECCS water.

The additional delay time in availability of borated water, and the resultant injection'of non-borated water from the VCT, is acceptable from the standpo kt of long term core cooling.

HOT LEG SWITCHOVER TO PREVENT P0TENTIAL BORON PRECIPITATION -

FSAR SECTION 6.3 Post-LOCA hot leg recirculation switchover time is determined for inclusion in Emergency Operating Procedures to ensure no boron precipitates in the reactor vessel following boiling in the core. This time is dependent on power level, the RCS, RWST, and Si accumulators water volume and boron concentration. A delay in the injection of borated water to the RCS would have no impact on the power-level, or volumes assumed for the RCS, RWST, and accumulators, and will-have negligible impact on the boron concentrations.

Therefore, there is negligible impact on the post-LOCA hot leg switchover time for VCSNS.

l. Enclosure 2 to Document Control Desk Letter TSP 890022

. Page 5 NON-LOCA TRANSIENTS The current TS (Table 3.3-5) were originally supported by non-LOCA analyses which assumed the following delays for the delivery of borated water to the RCS:

1. SI signal generation (2 seconds)
2. Diesel start - including time to come up to speed (10 seconds)
3. Valve stroke times and pumps to full speed (10 seconds)

However, this assumes that the VCT aco F '~ autlet isolation valves ,troke sim itaneously rather than sequentially. Yhe valve interlock lcgic increases the Jelay time for the availability of borated water by 15 seconds (conservatively) to 27 seconds with offsite power and 37 seconds witiicu cifsite power, The only non-LOCA transient impacted by the increased time delay is the Steam D

Line Break event. No other FSAR Chapter 15 transient relies on short-term borotion from the RWST to mitigate the event. Based on the current Steam Line Break analysis for VCSNS ar.a sensitivities performed for other plants, the additional time delay is acceptable. Specifically:

1) The additional delay in the availability of borated water occurs early
  • in the Steam Line Break transient when RCS pressure is relatively high and SI flow rates are relatively small due to head versus Si flow '

characteristics.

2) Previous sensitivities have shown that delays of this magnitude result in small changes in the analysis results. A comparison of cases with and without the additional SI delay has shown, over the limiting portion of the transient, maximum differences of 0.2% in power, 0.6'F in ~

RCS temperature, and 10 psia in RCS pressure. A VCSNS specific review of the Steam Line Break analysis demonstrated that there is sufficient nargin available in the analysis such that the conclusions presented in the FSAR remain salid.

3) The analysis assumes only one high-head SI pump is available. From analyses performed for other plants, it has been shown that Si boron concentration reducticn has little effect on the Steam Line Break Mass / Energy Release Inside Containment Analysis. The additional time delay is small when compared to the large change in available boron concentration. Therefore, the lirpact on the Steam Line Break Mass / Energy Release :nside Containaent Analysis is negligible.

Sensitivities performed for the Steam Line Break Superheated Mass / Energy Pelease Ou'. side Containnent Analysis show that the results are not sensitive to large changes in 51 flow (reference WCAP-10961, Rev. 1). The additional time delay is small when compared to the large change in total SI flow; therefore, it is concluded that the impact on the VCSNS Superheated Ma3/ Energy Releases Outside Containment is insignificant.

- . - - _ _ . - - _ _ _ _ . _ _ - . _. __ __ , _ ___ . . _ _.. .- __. ~._

Enclosure 2-to Document Control Desk letter '

TSP 890022 Page 6 In conclusion, the described changes to the ESF response times for Reactor Building Pressure - High, Pressurizer Pressure - Low, Differential Pressure Between Steam Lines - High,'and Steam Line Pressure - Low in Technical Specification Table 3.3-5. Items 2a, 3a, e,a, and Sa, will not invalidate the analyses or subsequent conclusions in the FSAR for all non-LOCA transients.

-The proposed changes reflect the bases for the current TS and are supported by current accident analyses.

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. Enclosure 3 to Document Control Desk letter TSP 890022 Page 1 PROPOSED TECHNICAL SPFCIFICATION CHANGE TSP 890022 VIR31L C. SUMMER NUCLEAR STATION DETERMlHATION OF N0 SIGNIFICANT HAZARDS CONSIDERATION DESCRIPTION OF AMENDMENT REQUEST SCE&G proposes to modify the VCSNS TS to revise 15 3/4.3.2 Table 3.3-5,

" Engineered Screty Features Response Times," to reflect the closing time associated with the Volume Control Tank outlet isolation valves, subsequent to a safety injection signal ("S").

Technica! tr . <

ns Table 3.3-5, " Engineered Safety Features Response Times," irovidt me interval from when a monitored ESF parameter exceed. - 'c r etpoint at the channel sensor until the ESF equipment

-is cap 4 -

~

,9 its intended safety function. These response times are meas - ince per eighteen months to verify that the ESF actuatin ith each channel is completed within the time limit specifics ie. This verification assures that the assumptions used for the L0u. _ ..on-LOCA accident analyses remain valid.

In the normal configuration of the CVCS, the high head safety injection pumps take suction from the VCT, When an "S" signal is generated from the protection logic, a signal is sent to start the high head safety injection (charging pumps) pumps, and to open the RWST isolation valves to align the borated water source for injection to the RCS. Once the RWST outlet itolation valves indicate fully opened, the outlet isolation valves on the VCT begin to close. The sequential valve stroke time of the RWST and the VCT outlet valves can be as long as 25 seconds and is designed to ensure that a net positive suction head is maintained on the suction of the pumps. The hydrogen pressurized VCT is the source of Si flow until the VCT isolation valves are closed, and the suction is changed to the RWST. The VCT isolation valve closure time affects the time assumed at which the borated water in the RWST is available to the suction of the high head safety injection pumps.

Current TS times only address the opening of RWST outlet isolation valves to maintain a flow path and not the closing of the VCT outlet isolation valves.

The proposed TS change will increase the allowable response times in Table 3.3-5 to incorporate the additional time required for the sequential stroking of the RWST and VCT valves.

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Enclosure 3 to Document Control Desk Letter TSP 890022 Page 2 BASIS FOR PROPOSED hU SIGNIFICANT HAZARDS CONSIDERATION SCE&P has evaluated the proposed TS change and has determined that it represents a no significant hazards consideration based on the criteria established in 10 CFR 50.92(c). Operation of VCSNS in accordance with the proposed action will not:

(1) Involve _a significant increase in the probability or consequences of any accident previously evaluated. The proposed change increases the response time of certain ESF functions to account for the sequential stroking of the outlet isolation valves for the VCT'and the RWST. The increase in response time is supported by the current accident analyses. This change is needed to ensure thtt assumptions utilized in the Steam Line Break accident analysis are properly addressed in the Technical Specifications.

(2) Create the possibility of a new or different kind of accident from any previously evaluated. This Technical Specification change is requested to ensure that the Technical Specification requirements support the assumptions utilized in the present safety analyses.

The change does not introduce the potential for new or different accidents from those currently analyzed.

(3) Involve a significant reduction in a margin of safety. The proposed change is requested to incorporate into the Technical Specifications the response times associated witn SI signals which support the plant's current safety analyses and margins of safety.

Therefore, the change will not reduce the margin of safety.

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