ML20102A811

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Proposed TS 3/4.3.3.6 & Tables 3.3-10 & 4.3-7 Re Accident Monitoring Instrumentation
ML20102A811
Person / Time
Site: Summer South Carolina Electric & Gas Company icon.png
Issue date: 07/22/1992
From:
SOUTH CAROLINA ELECTRIC & GAS CO.
To:
Shared Package
ML20102A808 List:
References
NUDOCS 9207270324
Download: ML20102A811 (20)
v • d • e


Text

. _ . . _ _ -

, Enc 16sure 2Lto-Document Control Desk-Letter-TSP 890005 Page 1 PRGPOSED TECHNICAL SPECIFICATION CilANGE - TSP 890005 VIRGIL C. SUMMER NUCLEAR STATION

. LIST OF'AFFECTED PAGES

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-3/4 3 -

3/4 3-57 3/4 3-57a New Page 3/4 3-57b New Page

-3/4 3-58

-3/4 3-58a New Page 3/4-3-59 3/4 3-60 II' t

l

- 9207270324 920722 PDR ADOCK0500gS P

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\

1 INSTRUMENTA110N ACC10EN1 MONITORING INSTRUMENTATION LINITING CONDITION FOR OPERATION _

3.3.3.6 The accident monitoring instrumentation channels shown in Table 3.3-10 shall be OPERABLE.

APPLICA8ILITY
- MBBES i, I nd 3c N S S b O W M l* T6 b k 3. 3 -/O.

ACTION:

A:s hoorn iw TaMc K 3-1o.

a. p.fith the number of 0FERABLE accident monitoring ch4' ,

/ the RequtiTd' umber of ChannelsAhown in Table 3.3-10grestore

/' the dnoperable channc1(s) to

( f at least HOT 5 KIT 00WN thewithip,4PERABL next 12 hoers. / }status withjri 7 d 1

(l b./ With the numbeg of OPERABLE accident monitoring channels l 1

ess thap the Minimum Channels OPE 8A8LE requirements of J ble 3.3-1(!; either restore the inoberable. channels to OPERABLE status witin 48" hours or be in at least HOT 4HUTDOWN within the next-12 hours. -

9 [ The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS l

4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated

.0PERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations 4

j at the frequencies shown in Table 4.3-7. F l

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c TABLE 3.3 E

=

ACCIDENT MONITORING INSTRUMENTATION C

I'i ' #

  • REQUIRED h0. MINIMUM' '

INSTRUMENT OF.

CHANNELS-

' CHANNELS OPERA 8LE -

/ ~

1. Reactor Building Wide Range Pressure 2. I  !

'2. Reactor Coolant . Outlet Temperature-- THOT (Wide Range) 2. .

3.

Reactor Cooiant; Inlet Temperature : TCOLD (Wide Range) '2 .- >

4 Reactor Coolant Pressure - Wide Range lf 1 '

l' I

5. Pressurizer Water Level 2 -

i

6. Steam Line Pressure .' gen.

2 1/ steam generator ,

$ 7. Steam Generator Water, Level - Wide. Range 1/stm. ge 1/ steam generator ,

r

8. Emergency Feedwater Flow d 1/ . gen.

1/ steam generator 9.

/ i(l

-Refueling Water Storage' Tank er le -

2 1

10. Boric Acid. Tank Water L I 2/ tank 1/ tank
11. Reacter Building 6 ay P p cfrarge F1 2 '1
12. Reactor Bui ng Ten,pera 2 1

[h .

t

. , ,, -~.-u-.._..-... . . . . _

V A ILi s 4

g-

. TABLE 3.3-10'(Continued) ~

h'

[g ACCIDENT MONITORING INSTRUMENTATION-

- raa

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e

.)t 5

TOTAL h1 ' /

INSTRUMENT- 0F NNELS CHANNELS PERABLf/f

13. Reactor Building /RifR Sump Level 2 1 14 DELETED.

T

15. . Condensac - Storage Tank' Level 2 -1 '
16. " Reactor Building Cooling Unit Service Water Flow p i
17. g Service Water Temperature-Reactor Building oling Uni 2 pair R

u (Inlet and Discharge) I l 1 pair T 18.' Nacil Storage" Tank Level E 1

l. 19. Reactor Coolant System Subcooli Margin nito 2 1
20. Pressurizer PORV Position idicator 2/ valve
  • 1/ valva *
21. Pressurizer PORV Bio Valve.Pos Indi or 1/ valve 1/ valve
22. Pressurizer Saf Valve P ition Ind" ator 2/ valve 1/ valve
23. In-Core Th ocouples, V 4/ core 2/ core quadrant q quadrant R

g ,24. Re or Vess

2. 1 8
n s g
- Jnot ' L ute% nth associated block valve is closed per Specification 3.4.4

!"r, s

L

_m_. .._._.t___ m

_ , g ___ ., . .,, , ,, ,. g

z.y

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Ex.  : TABLE 4.3-7l

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. ACCIDENT MONITORING INSTRUMENTATION 'SURVEILI_ANCE ' REQUIREMENTS

e

.E. .

" '. CHANNEL CHANNEL-s ' INSTRUMENT.

CHECK - -

CALIBRATION

  • 1 g /. ~ Reactor' Building Wide Range' Pressure f $- R G (. Reactor Coolant Outlet Temperature - TNT (Wide Range) M' S 'R *

~7 E. Reactor Coolant Inlet Temperature - Tg (Wide Range): <W' S R 4

S X. Reactor Coolant Pressure :. . Wide Range :

4

,M . S ' R' ,,

9E Pressurizer Water level ,W b R x

Y

/h4' Steam Line Pressure XS 'R

$ /5X Steam Generator Water level - Wide. Range P S R

/7X Emergency Fee &ater Flow 'wS R

, /s/ JI RWST Water Level M R

10. "cric Acid Tcna S!dti e t c! ?I --

R  ;

F v

II. Etoctui Bullusiy Sp oy 7 @ Oischarp Tic. O E . .

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- C 3- --TABLE 4.3-7:(continued) . .

x m

o '

ACCIDENT' MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS - --

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^

yi ' INSTRUMENT .;

' CHANNEL ' CHANNEL

" ' CHECK-

12. CALIBRATION 9:2cter $'ri!di~; TE ;:rrate : "

t_

[JT. Reactor' Building /RHR Sump Level:.

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Reactor Coolant System Subcooling.. Margin "--'ter- ~

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INSTRUMENTATION

-ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION ,

3.3.3.6 The' accident monitoring-instrumentation channels shown in Table 3.3-10 shall_be OPERABLE.

-APPLICABILITY As shown-in Table.3.3-10.

ACTION:

a. As shown in Table 3.3-10.
b. The provisions of Specification 3.0.4 are not applicable. _

SURVEILLANCE REQUIREMENTS _,

.4.3.3.6 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE _by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.*

Surveillance Requirements for the Hydrogen Monitoring Instrumentation are governed by TS 4.6.5.1.

For the Reactor Building Radiation Level Instrumentation, a CHANNEL

-CALIBRATION may consist af an electronic calibration of the-channel, not including the detector, for the range-decades above 10R/hr and a single point calibration of the detector below 10R/hr with an installed or portable gamma source.

SUMMER - UNIT 1 3/4 3-56

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c, TABLE 3.3-10 j c:

25

  1. 1
  • ' ACCIDENT MONITORING INSTRUM14'ATION 8

~* REQUIRED h.NIMUM

"' NO. OF CHANNELS APPLICABLE INSTRUMENT MODES CHANNELS OPERABLE ACTION Reactor Building Pressure - Narrow Range 2 1 1 1, 2, 3 - l

1.  !

Instrument Loop / Indicator: l Channel D IPT-951/IPI-951 l

Channel B IPT-952/IPI-952 1 Reactor Building Pressure - Wide Range 2  ? I 1, 2, 3 2.

Instrument Loop / Indicator:

Channel D IPI-954A/IPI-954A Channel E IPT-95 B/IPI-954B 2 1, 2, 3, 4 l

3. Reactor Building Padiation Level - High 2 1 Id Range - Instrument loop / Indicator:

e= Channel A RMG-18 r to Channel B RMG-7 i 1, 2 I 2 3 3 4. Reactor Building Hydrogen C;r. centration 1 Instrument Loop / Indicator:

Channel A IAE-8263A/ICI-8257 Channel B IAE-8263B/ICI-8258 2 1 1 1, 2, 3

5. Reactor Building /RHR Su=p Level Instrument loop / Indicator:

Channel A ILT-1969/ILI-1969 l Channel B ILT-1970/ILI-1970 1 1, 2, 3

6. Reactor Coolant Outlet Temperature - THot 2 1 '

- Wide Range - Instrument Loop / Indicator:

Channel A ITE-413/ITI-413 l Channel A ITE-423/ITI 423 i Channel E ITE 433/ITR-413

TABLE 3.3-10 (continued)

E ACCIDENT MONITORING INSTFLHENTATION C

5 REQUIRED MINIMLN

" NO. OF CHANNELS AFPLICABLE INSTRUMENT MODES

~ CHANNELS OPER4BLE ACTION q l

1 1, 2, 3

7. Reactor Coolant Inlet Temperature - TCold 2 1

- Wide Range - Instrument / Loop Indicator:

Channel E ITE-410/ITI-410 Channel E ITE-420/ITI 420 Channel E ITE-430/ITR-410 1 1, 2, 3

8. Reactor Coolant Pressure - Wide Range 2 1 Instrument Loop / Indicator:

Channel E IPT-402/IPI 402 Channel A IPT 403/IPI-403 j Pressurizer Water Level 2 I I 1, 2, 3 l 9.

u s

Instrument Loop / Indicator:

a Channel A ILT-459/ILI 459 w Channel D ILT 460/ILI 460 h Channel B ILT 461/ILI 461 l

  • Reactor t.colant System Subcooling Margin 2 1 I 1. 2, 3 10.

Instrument loop / Indicator:

Channel A ITM 499A Channel B ITM-499B

11. Reactor Vessel Level 2 I I 1, 2, 3 Instrument Loop / Indicator:

Channel A ILT-1311/ILI-1311, ILT-1312/ILI-1312 Channel B ILT-1321/ILI-1321, ILT-1322/ILI-1322 i

m

e - _ _

-p TABLE 3.3-10 (ct' .xed) .

ACCIDENT MONITORING INSTRUMENTATION jR EE

REQUIRED MINIMUM

~# NO. OF CHANNELS APPLICABLE )

INSTRUMENT

"' CHANNELS OPERABLE ACTION MODES Core Exit Temperature 4/ core 2/ core 1 1,2,3 12.

Instrument Loop / Indicator: quadrant / quadrant /

l Channel A Channel B channel channel I

ITEs 4,12,27,32, ITEs 1,5,10,30 2,29,31,35,13.22 34,43,44,49,50,3, l 11,36,40,6,7,17,IB, i

47,9,15,25,45,21, 28,33,39,41,19,24, 8.14,38,48,16,20, ,

1 26,42,46,23 37,51 (Primary display is the plant computer)

(Backup displays are ITM 499 A&B) 2 1 1 1, 2, 3

13. Neutron Flux If Instrument toep/ Indicator:

Channel A INM-35 T' Channel B INM-36

?/stm. gen. 1/stm. gen. 1 1, 2. 3

14. Steam Line Pressure Instrument loop / Indicator:

SG A IPTs-474, 475, 476/IPIs-474, 475, 476 SG B IPTs 484, 485, 486/IPIs 484, 485, 485 SG C IPIs-494, 495, 496/IPIs-494, 495. 496

15. Steam Generator Water Level Wide Pange 1/stm. gen 1/stm. gen. I 1, 2, 3 Instrument loop / Indicator SG A ILT-477/ILT-477 }

SG B ILT-487/ILI 487 SG C ILT-497/ILI-497

16. Steam Generator Water Level - Narrow Range 2/stm. gen. 1/stm. gen. I 1, 2, 3 Instrument Loop / Indicator:

SG A ILTs 474, 475, 476/ILIs 474, 475, 476 SG B ILTs 484, 485, 486/ILIs 484, 485, 486 SG C ILTs 't, 495, 496/ILIs 494, 495, 496

TABLE 3.3-10 (continued) .

F' ACCIDENT MONITORING INSTRUMENTATION E MINIMUM

REQUIRED
  1. 0. OF CHANNELS APPLICABLE INSTRUMENT CHANNELS OPERABLE ACTION MODES
17. Emergency Feedwater Flo. - Wide Range 1/stm. gen. 1/stm. gen. I 1, 2, 3 Instrument Loop / Indicator:

SG A IFT-3561/IFI-3561 i SG B IFT-3571/IFI-3571 SG C IFT-3581/IFI-3581

18. Emergency Fecdwater Flow - Narrow Range 1/stm. gen. 1/stm. gen. I 1, 2, 3 4 1

Instrument Loop / Indicator:

SG A IFT-356IA/IFI-3561A SG B IFT-3571A/IFI-3571A SG C IFT-3581A/IFI-3581A 2 1 1 1,2,3 Id 19. Refueling Water Storage Tank Level

^

  • tment Loop / Indicator:

y' Ct. el A ILT-990/ILI-990 g; Channel B ILT-992/ILI-992

lable 3.3.10 (Continued)

ACTION STATEMENTS ACTION 1. a. With the number of OPERABLE accident monitoring channels less than the Required Number of Channels shown on Table 3.3-10, restore the inoperable channel (s) to OPERABLE status within 30 days 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 HOT SHU100WN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the number of OPERABLE accident monitoring channels less than the Minimum Channels Operable requirement of Table 3.3-10, either restore the inoperable channels to OPERABLE status within 7 days 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 HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2. a. With the number of OPERABLE channels less than the required by the Minimum Channels Operable requirement, either restore the inoperable channel (s) to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or:

1) Initiate the preplanned alternate method of monitoring the appropriate parameter (s), and
2) Prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability, and the plans and schedule for restoring the system to OPERABLE status.

ACTION 3. a. With one hydrogen monitor inoperable, restore the inoperable monitor to the OPERABLE status within 30 days or be in at least HOT STANDBY within the next 6 hourt,

b. With both hydrogen monitors inoperable, restore at least one monitor to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.

G SUMMER - UNIT 1 3/4 3-58a

e

~

TABLE 4.3-7 _

'T

.k

' ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS -

e CHANNEL CHANNEL E INSTRUMENT CHECK CALIBRATION w

S R

1. Reactor Building Pressure - Narrow Range S R
2. Reactor Building Pressure - Wide _ Range S R
3. Reactor Building Radiation Level - High Range
4. Reactor Building Hydrogen Concentration ,

S R

5. Reactor Building /RHR Sump Level R R

Reactor Coolant Inlet Temperature - TCold (Wide Range) S R 7.

l Reactor Coolant Pressure - Wide Range S R

! 8.

5 R

9. Presstrizer Water Level

! R'

10. Reactor Coolant. System Subcocling Margin 5 R L. 11. Reactor Vessel Level S R
12. Core Exit Ten!perature 5 R
13. Neutron Flux

~

TABLE 4.3-7 (continued)

. en - -

5' p; ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,

o c: CHANNEL CHANNEL INSTRUMENT-lc > CHECK CALIBRATION
14. Steam Line Pressure S R
15. Steam Generator Wa"ter Leve1L- Wide Range S R
16. Steam Generator Water Level - Narrow Range 5 R
17. Emergency Feedwater Flow - Wide Range S R
18. -Emergency Feedwater Flow - Narrow Range S R
19. Refueling Water Storage Tank level S R

't 15

,1 l

t .

Enclosure 3 to Document Control Desk Letter TSP 890005 Page 1 PROPOSED TECHNICAL SPECIFILATION CHANGE - TSP 890005 VIRGIL C. SUMMER NUCLEAR STATION DESCRIP110N AND SAFETY EVALVATION DESCRIPTION OF AMENDMENT REQUEST SCELG proposes to modify the VCSNS TS to revise TS 3/4.3.3.6. Tables 3.3-10 and 4.3-7, " Accident Monitoring Instrumentation," and " Accident Monitoring Instrumentation Surveillance Pequirements," respectively. This change will reflect the plant variables (key variables) that provide primary information required to permit control room operators to take specified manually controlled actions, for which no automatic control is provided, after initial stages of an accident and that are required for safety systems to accomplish their safety function for design besis accident events.

The instrumeritation, to be included-in Tables 3.3-10 and 4.3-7, is designated type A categor; 1, except for the Subcooling Margin Monitor which is category 2 but meets the intent of category 1, and non-type A category 1. This designation is in accordance with the guidance given in Regulatory Guide (RG) 1.97, revision 3, as reported to the NRC and found acceptable in their Safety Evaluation Reports dated November 13, 1987, and July 27, 1988. The following instrumentation shall be included in Tables 3.3-10 and 4.3-7.*

1. Reactor Building Pressure - Narrow Range
2. Reactor Building Pressure - Wide Range
3. Reactor Building Radiation Level - High Range

~

4.- Reactor Building H2 Concentration

5. Reactor Building /RHR Sump Level
6. Reactor Coolant Outlet Temperature -THot - Wide Range
7. Reactor Coolant inlet Temperature - TCold - Wide Range
8. Reactor Coolant Pressure - Wide Range
9. Pressurizer Water Level
10. Reactor Coolant System Subcooling Margin
11. Reactor Vessel Water Level
12. Core Exit Temperature
13. Neutron Flux
14. Steam Line Pressure
15. Steam Generator Water Level - Wide Range
16. Steam Generator Water Level - Narrow Range
17. Emergency Feedwater Flow - Wide Range
18. Emergency feedwater Flow - Narrow Range
19. Refueling Water Storage Tank Level

- _ - - _ _ - _ - _ _ _ - - - - - - - - - - - - - - - - - ~

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

4

  • Enclosure 3 to Document Control Desk letter TSP 890005 Page ?

The following non-Category 1 instrumentation shall be deleted from Tables 3.3-10 and-4.3-7: ,

1. Boric Acid Tank Water Level
2. Reactor Building Spray Pump Discharge flow
3. Reactor Building Temperature
4. Condensate Storage Tank Level
5. Reactor Building Cooling Unit Service Water flow
6. Service Water Temperature-Reactor Building Cooling Unit (Inlet and Discharge) 7 Pressurizer PORV Position Indicator .
8. Pressurizer PORV Glock Valve Position Indicator .
9. Pressurizer Safety Valve Position Indicator '
10. Sodium Hydroxide Storage Tank Level The action statements for the Reactor Building Radiation Level and the Reactor Building Hydrogen Concentration have been changed to be consistent with existing technical specifications. The allowable outage time for the rest of the instruments has been changed from 7 to 30 days (for conditions in which the number of OPERABLE accident monitoring channels is less than the Required Number of Channels) and from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days (for conditions in which the number of OPERABLE accident monitoring channels is less than the Minimum Channels Operable) both requiring a reduction to mode 3 (H01 STANDBY) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and then to mode 4 (HOT SilV1DOWN) within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

Administrative changes were made to 15 3/4.3.3.6, Tables 3.3-10 ano 4.3-7.

SAFE 1Y EVALUATION

-lhe proposed TS change incorporates the type A category 1 and non-type A category 1 variables into VCSNS's " Accident Monitoring Instrumentation, Table 3.3 10 " and " Accident Monitoring Instrumentation Surveillance Requirements.

Table 4.3 7." VCSNS's selection of category I key variables is documented in its Summary Report on Reguletory Guide 1.97, originally submitted to the NRC on April 15,1985, amended by various letters referenced in this amendment request (Enclosure 1), and evaluated and accepted by the NRC in their letters of November 13, 1987, and July 27, 1988, and their enclosed Safety Evaluations.

The action statements for the Reactor Building Radiation tevel indication and L the_ Reactor Building Hydrogen Concentration have been changed to be consistent with TS 3/4.3.3 and 3/4.6.5, respectively.

- The increase of the allowable outage time from 7 to 50 days in action statement 1, applicable to the condition in which the number of OPERABLE 1

m _ ._ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ . . _ . _ - _ _ _ _ _ _ , _ _ _ _ . . . . _ ___ _ . _

. . .- . - - - - . _ . _ - - - . - . ~ . _ _ . .

Enclosure 3 to Document Control Oesk tetter TSP 090005

. Page 3 i

accident monitoring channels is less than the Required Number of Channels, is  ;

based on: VCSNS's low failure rates for these types of components, the '

availability of a remaining operable channel, the passive nature of the channel, and the low probability of an event requiring accident moni'oring ,

instrumentation during this interval, t r

The increase of the allowable outage time from 48 'iours to 7 days in action l

. statement 1, applicable to the condition in which the number of OPERABLE accident monitoring channels is less than the Minimum Channels Operable, is based on the relatively los probability of an event requiring accident monitoring instrument operation, and the availability of alternate means to

. obtain tne required information. Continuous operation with two required i channels inoperable in a function is.not acceptable because the alternate indications may not fully meet the requirements applied to the accident monitoring instrumentation. Therefore, requiring restoration of one t inoperable channel of the function limits the risk that the accident L monitoritg function will be degraded should an accident occur. '

.The change.of action statement 13 requiring the plant to go to.IIOT STANDBY and i then to HOT SHUTDOWN has no impact on the safety of the plant since it merely ,

identifies the proper stages of plant shutdown which assumes that the plant  ;

is at 0% power.  :

i z

9, m-r u-we-,rew,.---,,w,-es,v-wwm>ew+--w.-w-n-, ,-r~r,,w*,,-ry--m*wwe%#t-,.w-+%--vi.,,, . --ye-o ,,,,,-w,, w, a.,--yv---am+-w.ewpry3,yy.,v -=--,-vw-, = .-e m my-et

Enclosure 4 to Document Control Desk letter

' TSP 89005 Page 1 PROPOSED 1ECHNICAL SPECIFICATION CHANGE - TSP 890005 l

VIRGIL C. SUMMER NUCLEAR STATION DETERMINATION OF NO SIGNIFICANT HAZARDS CON 510 ERAT 10N DESCRIPTION OF AMENDMENT REQUEST SCE&G proposes to modify the VCSNS TS to revise TS 3/4.3.3.6, Tables 3.3-10 and 4.3-7, " Accident Monitoring Instrumentation," and " Accident Monitoring Instrumentation Surveillance Requirements " respectively. This change will reflect the plant variables (key variables) that provide primary information required to permit control room operators to take specified manually controlled actions, for which no automatic control is provided, after initial stages of an accident and that are required for safety systems to accomplish their safety function for design basis accident events.

The instrumentation, to be included in Tables 3.3-10 and 4.3-7. is designated type A category 1, except for the Subcooling Margin Monitor which is category 2-but meets the intent of category 1, and non-type A category 1. This

-designation is-in accordance with the-guidance given in Regulatory Guide (RG) 1.97, revision 3 as reported to the NRC and found acceptable in their Safety Evaluation Reports dated November 13, 1987, and July 27, 1988. The following instrumentation shall be included Tables 3.3-10 and 4.3-/:

1. Reactor Building Pressure - Narrow Range
2. Reactor Building Pressure - Wide Range
3. Reactor Building Radiation Level - High Range
4. Reactor Building H2 Concentration
5. Reactor Building /Rt!R Sump level
6. Reactor Coolant Outlet Temperature -THot - Wide Range
7. Reactor Coolant inlet Temperature - TCnid - Wide Range
8. Reactor Coolant Pressure - Wide Range
9. Pressurizer Water Level
10. Reactor Coolant System Subcooling Margin
11. Reactor Vessel Water Level
12. Core Exit Temperature
13. Neutron Flux
14. Steam Line Pressure
15. Steam Generator Water Level - Wide Range
16. Steam Generator Water Level - Narrw Range
17. Emergency Feedwater Flow - Wide Range
18. Emergency Feedwater Flow - Narrow Range
19. Refueling Water Storage Tank Level I

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Enclosure 4 to Document Control Desk Letter l

' TSP 890005 Page 2 The following instrumentation shall be deleted from Tables 3.3-10 and 4.3-7:

1. Boric. Acid. Tank Water Level

.2. Reactor Building Spray Pump Discharge flow

3. . Reactor Building Temperature
4. Condensate Storage Tank Level S. Reactor Building Cooling Unit Service Water Flow
6. Service Water Temperature-Reactor Building Cooling Unit (Inlet and Discharge)
7. Pressurizer PORV Position Indicator
8. Pressurizer PORV Block Valve Position Indicator
9. Pressurizer Safety Valve Position Indicator
10. Sodium Hydroxide Storage Tank Level l The action statements for the Reactor Building Radiation Level and the Reactor Building Hydrogen Concentration have been changed to be consistent with existing technical specifications. The allowable outage time for the .

-rest of the instruments has been changed from 7 to 30 days (for conditions in which the number of OPERABLE accident monitoring channels is less than the Required-Number of-Channels) and from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days (for conditions in -

which the numbec of. OPERABLE accident monitoring channels is less than the i Minimum Channels Operable).both requiring a reduction to mode 3 (HOT STANDBY) within6 hours,andthentomode4(HOTSHUT00WN)withinthenext6 hours.

Administrative changes were made to TS 3/4.3.3.6, Tables 3.3-10 and 4.3-7.

BASIS FOR DETERMINATION OF N0 SIGNIFICANT HAZARDS CONSIDERATION SCE&G has-evaluated the proposed TS change and has determined that it does '

not represent a 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 the consequences of an accident previously evaluated.

Regulatory Guide 1.97 furnishes standards acceptable to the NRC for providing instrumentation to monitor plant variables and systems during i and following an accident. The purpose of the accident monitoring  :

instrumentation is to display plant variables that provide information required by the control room operators for manual actions and long term

-recovery. -Determination of variable types and category designations-for VCSNS was accomplished from a review of the Emergency Response l Guidelines (ERGS), the Final Safety Analysis Report, and the

--Westinghouse Owners Group (WOG) ERGS. The WOG_ ERGS were used at VCSNS -

as a basis for the Emergency Response Procedures. Operability of the instruments used for accident monitoring ensures there is sufficient

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Enclosure 4 to Document Control Desk tetter TSP 890005 Page 3 information available on selected plant parameters to monitor plant status during and following an accident. The changes proposed do not affect components that can cause an accident. The increase in allowable outage times from 7 to 30 days or from 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to 7 days does not significantly affect the consequences of an event previously evaluated.

The channel redundancy and the relatively short outage times, coupled with the low probability of an event requiring accident monitoring instrumentation during this interval, ensure that sufficient information is available for operator manual actions. The condition of the plant in either HOT STANDBY or HOT SHUTDOWN the first stage of the plant shutdown process, has no impact on the assumptions made in the accident analysis.- Therefore, the proposed change does not increase the probability or consequences of any accident previously evaluated.

- (2) Create the possibility of a new or different kind of accident from any previously analyzed.

The proposed change is consistent with the requirements of RG 1.97. The accident monitoring. instrumentation will-make available reliable information to plant control room operators to mitigate the consequences of a design basis accident. The first stage of plant shutdown. HOT STANDBY and HOT SHUT 00WN, are plant modes for which VCSNS has been analyzed. Therefore, the changes proposed do not create the possibility of a new or different kind of accident from any previously analyzed.

(3) Involve a significant reduction in a margin of safety.

The inclusion of category 1, type A or 8, instrumentation in the TS provides assurance that adequate information is available to the operators to maintain VCSNS in a safe condition during and following a design basis accident. Accomplishment of specific manual action by the control room operators is enhanced due to the availability and reliability of the indications. The proposed changes do not affect the design or operation of safety related components relied upon to automatically mitigate the consequences of a design basis event. The proposed change from HOT SHUTDOWN to HOT STANDBY as the first stage of plant shutdown will not affect the design or operation of any safety <

related system or component. Therefore, the changes proposed would not involve a reduction in any margin of safety.

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