ML20053B897

Forwards marked-up Proof & Review Copy of Tech Specs Re Seismic & Accidental Monitoring Instrumentation,Electrical Power Sys & Reactor Trip & ESF Actuation Sys Instrumentation

ML20053B897
Person / Time
Site:	Summer
Issue date:	05/24/1982
From:	Nichols T SOUTH CAROLINA ELECTRIC & GAS CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8206010325
Download: ML20053B897 (9)

Text

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.(e SOUTH CAROLINA ELECTRIC & GAS COMPANY POST OFFICE Box 764 COLUMBI A. S. C. 29218 May 24, 1982 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Virgil C. Summer Nuclear Station Docket No. 50/395 Technical Specifications

Dear Mr. Denton:

Attached are miscellaneous changes to the Proof and Review Copy of the Technical Specifications for the Virgil C. Summer Nuclear Station.

The following is a brief explanation of the reasons for the requested changes:

PAGE EXPLANATION 3/4 3-48 The position of the triaxial peak accelograph (item 2.b) and has been relocated as described in our May 3,1982, letter 3/4 3-49 to Mr. Denton.

3/4 3-58 Pocition switches have been installed in each pressurizer and safety valve to provide valve position indication.

3/4 3-60 3/4 8-13 Vital buses 5907 and 5908 are not required to be backed by an inverter connected to a D.C. bus because the momentary loss of these panels will not violate any analysis or detract from the safety of the plant. These panels feed j

control channels 1 and 3 and the auxiliary relay racks. An analysis of loss of power to control systems was included in ou February 26, 1982, letter to Mr. Denton.

B3/4 3-1 This statement is added to the bases to clarify that although listed in Table 3.3-3, Items 3.c.3, 5.a. 6.g, 6.h, and 8. are not Engineered Safety Features.

Please contact Mr. L. D. Shealy if you have any questions.

Very truly yours, 0 6 01.0 95 A

[ d' T. C. Nichols, Jr.

Senior Vice-President LDS/vtw Power Operations Attachments cc: Page 2 J

e Mr. H. R Denton l

May 24, 1982 Page 2 cc:

V. C. Summer (w/o attach.)

G. H. Fischer (w/o attach.)

H. N. Cyrus T. C. Nichols, Jr. (w/o attach.)

O. W. Dixon, Jr.

M. B. Whitaker, Jr.

J. P. O'Reilly H. T. Babb D. A. Nauman C. L. Ligon (NSRC)

W. A. Uilliams, Jr.

R. B. Clary O. S. Bradham L. D. Shealy A. R. Koon M. N. Browne G. J. Braddick M. J. Virgilio J. L. Skolds J. B. Knotts, Jr.

B. A. Bursey NPCF File J

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INSTRUMENTATION I-d TABLE 3.3-7

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SEISMIC MONITORING INSTRUMENTATION h)

MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1.

Triaxial Time-History Accelerographs xY System, including the following components:

a.

Reactor Building Foundation Hat 0.1 to 40 Hz Accelerometer 0.01-to 1.0g i

b.

Reactor Building Ring Girder 0.1 to 40 Hz Accelerometer 0.01 to 1.0g 1

c.

Reactor Building Foundation Mat 1 to 10 Hz Trigger 0.005 to 0.02g 1*

2.

Triaxial Peak Accelographs a.

Top of Steam Generator 0-32 Hz.

-5g to +5g 1

P-es w ieen s~ me L: e b.

Scttem of Reictor '!cere!

0-32 Hz

-5 to +5g i

c.

RHR System Heat Exchanger

~0-20 Hz

-2g to +2g 1

3.

Triaxial Seismic Switches a.

Reactor Building Foundation Mat 0.1 to 30 Hz 0.01 to 0.25 g 1*

4.

Triaxial Response-Spectrum Recorders

)

a.

Reactor Building Foundation Mat (1) 1*

b.

Steam Generator Support (1) 1 c.

Intermediate Bldg., Elev. 463 (1) 1 d.

Auxiliary Bldg. Foundation (1) 1 b)

With control room, indication and/or alarm.

(1) Range varies for the multiple elements of the instrument, i.e.,1.6g at 2 Hz, 10g at 5 Hz, 34g at 10 Hz, 12g at 16 Hz.

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SUMMER - UNIT 1 3/4 3-48 J

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l INSTRU' MENTATION

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TABLE 4.3-4

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SEISMIC MONITORING INSTRUMENTATION SURVEILL NCE REOUIREMENTS ANALOG CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENTS AND SENSOR LOCATIONS CHECK CALIBRATION TEST

/

1. Triaxial Time-History Accelerographs, including the following components:

a. Reactor Building Foundation Mat M

R SA Accelerometer

b. Reactor Building Ring Girder M

R SA Accelerometer

c. Reactor Building Foundation Mat M

R SA Trigger *

2. Triaxial Peak Accelerographs

a. Tcp of Steam Generat NA R

NA 6 ary ine b.

ssu yiz.e rT!aactc. g..ma,ci NA R

NA cm c

c. RHR System Heat ' Exchange'r NA

.R NA

3. Triaxial Seismic Switches

a. Reactor Building Foundation Mat

• M R

SA

4. Triaxial Response-Spectrum Recorders

a. Reactor Building Foundation Mat

• M R

SA

b. Steam Generator Support NA R

NA

c. Intermediate Bldg. Elev. 463' NA R

NA

d. Auxiliary Bldg. Foundation NA R

NA G

With control room indications and/or alarm.

SUMMER - UNIT 1 3/4 3-t.9 J

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TAatE 3.3 10 (Continued) 2

• C N

ACCIDElli MotlITORIrlG IllSTRuttEllTATI0ft rn

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TOTAL HIttIMUM h

0F CllAtillELS

-4 CllAtitlELS OPERABLE ItJSTRUt4EllT 2

1 13.

Reactor BuildinD RilR Sump Level 2

1 14.

Reactor Building Level 2

1 15.

Condensate Storage Tank Level 2

1 16.

Reactor Building Cooling Unit Service Water Flow 17.

Service Water Temperature-Reactor Building Cooling Unit 2 pairs 1 pair R'

(Inlet and Discharge) 2 1

Y 18.

Ita0ll Storage' Tank Level

.t.

19.

Reactor Coolant System Subcooling Margin Monitor 2

1 1/ valve

• 1/ valve

• 20.

PORV Position Indicator 21.

PORV Block Valve Position Indicator 1/ valve 1/ valve Pa;kon Id.c Jou 1/ valve 1/ valve 22.

Safety Valve Acc= tical !!ca! tar a.

4/ core 2/ core 23.

In-Core Thermocouples quadrant quadrant 2~

l 24.

Reactor Vessel Level

o G

Aflot required when the associated block valve is closed per Specification 3.4.4.

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-s TADIE 4.3-7 (continued) c i

5 ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS m

g

=

c CilANNEL CilANNEL 5

CilECK CALIBRATION INSTRllMENT e

H R

12.

Reactor Buildin0 Temperature H

R 13.

Reactor BuildinD RiiR Sump Level H

R i

14.

Reactor Building Level i

H R

15.

Condensate. Storage Tank Level H

R 16.

Reactor Building Cooling Unit Service l

Water Flow w

H R.

17.

Service Water Temperature - Reactor Building Y

Cooling Unit (Inlet and Discharge)

H R

18.

Na0lt Storage Tank Level 19.

Reactor Coolant System Subcooling Margin Monitor M

R H*

R*

20.

PORV Position Indicator M

R 21.

PORV Block Valve Position Indicator h;kon Ind.rdo y 22.

Sef aty Valve.,. ::::t!::? "^-! to.-

a.

H R

H R

23.

In-Core Thermocouples se H

R N

24.

Reactor Vessel Level Hot requ, ired when the associated block valve is closed per Specification 3.4.4.

A g

.I l

i i

I t

• 9 N

s

e.

12 0 volf 11.'C. Vi}al But

1. 5907 en e G l2ecl.

f.

120 volf A.C. Vih l gas 4 590 2 en erg ze l.

ELECTRICAL POWER SYSTEMS 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION 3.8.3.1 The following electrical busses.sh'all be energized in the specified

.nanner with tie breakers open between redundant busses:

a.

Train A A.C. Emergency Busses censisting of:

1.

7200 volt Emergency Busses # 10A and 1EA.

2.

480 yolt Emergency Busses # 1DA1, IDA2 and 1EA1.

b.

Train B A.C. Emergency Busses consisting of:

1.

7200 volt Emergency Busses # 1DB and 1EB.

4 2.

480 volt Emergency Busses # 10B1,1DB2, and 1EBl.

c.

120 volt A.C. Vital Busses # 5902,-Si30P and 5901 energized from an associated inverter connected to D..C. Bus # 1HA*.

d.

120 volt A.C. Vital Busses # 5904,-MBB-and 5903 energized from an~

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associated inverter connected to D.C. Bus # 1HB*.

3.f.

125 volt D.C. Bus 1HA energized from Battery Bank #1A.

h P.

125 voit D.C. Bus 1HB energized from Battery Bank #1B.

APPLICABILITY:

MODES 1, 2, 3, and 4.

b. with o-e 4.c. tt;fal Fus ru,+ eua q ;,.al ne - en evs. a' f4e 4.G I/ihl ou be in af lee.sf Ifor s7MD BY wit 4:n 14e. nett ACTION:

But wtf in 2 k.>s

& hos. s a ml,*,.

cot p spurpeaw w,tf;,, t<e $ ff.u,,q s o fios.rs,

With one of the required trains of A.C. Emergency busses not fully a.

energized, re-energize the division within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

4 guases e s9o b 5901, f90 h or 54 0 'l C.)f.

With one4A.C. Vital -BurAeither not energized f rom its associated inverter, or with the inverter not connected to its associated D.C.

Bus:

(1) rc cncrgiz; thc A.C. Vitsi Su; within 2 hauc5 s 6 i.. ei o

ht liOT STAHOBY-within the ned C hab. 5 end in COLD SHUTDOWN

. within the fcilcwing 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />; and (2) re-energize the A.C. Vital Bus from its associated inverter connected to its associated D.C.

Bus within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

d. t.

With one D.C. Bus not energized from its associated Battery Bank, re energize the O.C. bus from its associated Battery Bank within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The

• .TAcee inverters may be' disconnected from their D.C. Bus for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> as necessary for the purpose of performing an equalizing charge on their associatec battery bank provided (1) their vital buses are energized, and (2) the vital busses associated with the other battery bank are energized from their associated inverters and connected to their associated 0.C. Bus.

SUMMER - UNIT 1 3/4 8-13

.fdAR 0 3 52

D 9

3/4.3 INSTRUMENTATION i45ES 3/4.3.1 and 3/4.3.2 ' REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION The' OPERABILITY of the Reactor Protec5 ion System and Engineered Safety Feature Actuation. System instrumentation ~and interlocks ensure that 1) the associated action and/or reactor trip will be initiated when the parameter monitored by'each channel or combination thereof reaches its setpoint, 2) the specified coincidence logic is maintained, 3) sufficient redundancy is main--

tained to permit a. channel to be out of service for testing or maintenance, and,4) sufficient system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity: assumed available in the facility design for the protection and mitigation. of accident and transient. conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable. to the original design standards. The periodic serveillance-tests performed at the minimum frequencies are sufficient to denionstrate this capability.

'The measurement of response time at the specified 'frequencie's provides assurance that the reactor. trip and the engineered safety feature actuation associated with each channel is completed within the time limi.t assumed in the acci, dent analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of. sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite, or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combina-tions indicative of various accidents, events, and transients.

Once the required logic combination is completed, the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition.

As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident 1) safety injection pumps start and automatic valves position, 2) reactor trip, 3) feedwater isolation 4) startup of the emergency diesel generators, 5) containment spray pumps start and automatic valves position, 6) containment isolation, 7) steam line isolation, 8) turbine trip, 9) auxiliary feedwater pumps start and automatic valve position, 10) containment cooling fans start and automatic valves position,

11) essential service water pumps start and automatic valves position, and

12) control room isolation and ventilation systems start.

ZA/ SERT 8

SUMMER - UNIT 1 B 3/4 3-1

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INSERT A Several automatic logic functions included in this specification are not necessary for Engineered Safety Feature System actuation but their functional capability at the specified setpoints enhances the overall reliability of the Engineered Safety Features functions. These automatic actuation systems are purge and exhaust-isolation from high containment radioactivity, turbine trip and feedwater isolation from steam generator high-high water level, initiation of emergency feedwater on a trip of the main feedwater pumps, automatic transfer-of the suctions of the emergency feedwater pumps to service water on low suction pressure, and automatic opening of the containment recirculation sump suction valves for the RHR and spray pumps on low-low refueling water storage tank level.

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