Text

Proposed Tech Spec Changes Re Thermal Design Flow,Fire Detection Instruments & Overpressure Protection Sys
	ML20027C552
Person / Time
Site:	Summer
Issue date:	10/08/1982
From:	; SOUTH CAROLINA ELECTRIC & GAS CO.
To:	;
Shared Package
ML20027C550	List: ML20027C549; ML20027C552;
References
	NUDOCS 8210180162
	Download: ML20027C552 (54)
	v • d • e

,

ATTACIMENT I REQUESTED CHANGE ER ,

i, VIRGIL C. SUMMER STATION i .

TECHNICAL SPECIFICATION 3.2.3 i

._ 8210180162 821008 C ' ~ ~';

PDR ADOCK 05000395 ~l P .PDR

f-.

PAGES AFFECTED The following is an explanation for the requested change to Figure 3.2.3.

PAGE EXPLANATION 3/4 2-10 Current estimates' of total reactor coolant systems (RCS) flow at the Virgil C. Summer Nuclear Station indicate that the measured flow, when measurement uncertainties of 3.5% are included, may be less than the Thermal Design (TD) flow used in the plant accident analyses. The measured flow will be determined at a future date by a flow calorimetric with the plant operating above 50% power. The finite time period allowed for restoring RCS flow to above the minimum required value in the present Technical Specification is based on the assumption that the low flow is due to time limited local effects, e.g. grid voltage / frequency dips. The concern at Summer is not with low flow due to time limited local effects, but rather with the measured RCS flow being less than TD flow when flow measurement uncertainties of 3.5% are included.

The Technical Specification is quite specific in the action to be taken if the measured flow is less than TD flow. The results of the required action is that Thermal Power will be limited to less than 5% of Rated Thermal Power. This is because the Technical Specification does not recognize the possibility of a long term reduction in- flow, nor the various trade-offs allowed by the relationships between flow, DNB, and core power. These trade-offs can be used to justify continued operation at some reduction in the maximum allowed power if the measured RCS flow is less than required.

It is widely recognized that the relationships between core power, flow, and DNB are:

3 Flow "

1%

&DNB l0f (Eq. 1) 3 Power "

1%

BDNB 1.8% (Eq. 2)

Thus the relationship between Power and Flow is:

3 Power "

1%

& Flow 1.8% (Eq. 3)

EXPLANATION ,

Based on a conservative assumption that the measured RCS flow will be no lower than 95% of TD flow, it is requested that a region of acceptable operation be added to Figure 3.2.3 for:

95% TD Flow 1 RCS Flow 5 100% TD Flow (Eq. 4)

Based on the relationship given by Equation 3, it is recommended that the maximum power level for this region be restricted to less than or equal to 95%

of Rated Thermal Power (2636 MWg ). This restriction of core power is the equivalent of an RCS flow increase of approximately 9% in terms of DNB margin. Since in this region the flow deficit is 5% or less, the power i

restriction will result in a minimum increase of approximately 4% in terms of DNB margin. Operation of the plant in this region within the specified power restriction results in no increase in Tavg, thus there is no temperature impact on the DNB margin.

The Technical Specifications and accident analysis results have been evaluated to determine the impact of

, operating within the defined nevregion of Figure 3.2-3 with the imposed restrictions. In all cases sufficient margin exists to allow continued plant operations. No Technical Specification limits require modification, including core limits, OTAT, OPAT, and Power Range Neutron Flux High Setpoint.

The core limits remain the same due to the increased margin to DNB afforded by the power reduction and interpretation that they will be valid for a maximum

, power level of 2636 MW instead of the design 2775 MW .

This implies that under these conditions 2636 MW should beconsideredtobe100%ofRatedThermalPowerkor

Figure 2.1-1. With this restriction applied to the i Safety limits there is no change in the core limits.

l thus the OTAT and OPAT trip setpointsremain unchanged.

l Utilizing the latest Westinghouse data, the uncertainty i

in the instrumentation for the Power Range Neutron Flux

! High trip function is 4.7% span (or 5.7% RTP). With l a normal assumption of reactor trip at 109% RTP the uncertainty analysis verifies that a trip will take l

place at 109% RTP plus 5.7% uncertainty or 114.7% RTP.

A 5% reduction in RCS flow requires a trip at 115.2%

RTP. Therefore, adequate margin exists in the instru-mentation such that no change in the nominal setpoint is necessary.

EXPLANTION

, Obviously, if the measured RCS flow (including 3.5% uncertainties) is equal to or greater than Thermal Design flow, operation will be in the acceptable region of the present Figure 3.2-3 and the requirements of this specification will remain unchanged. The addition of the new region to Figure 3.2-3 is only requested to preclude a need-less reduction to 5% RTP if the measured RCS flow is found to be low.

l i

1 POWER DISTRIBUTION LIMITS 4^'

ACTION: (Continued)

b. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that fj{} the combination of R , R and RCS total flow rate are restored to

"..' i within the above limits,2or reduce THERMAL POWER to less than 5% of RATED T.HERMAL POWER within the.next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .,

, c. Identify and correct the cause of the out-of-limit condition prior ;

to increasing THERMAL POWER above the reduced THERMAL POWER limit 67.s required by ACTION items a.2. and/or b. above; subsequent POWER V.i/ OPERATION may proceed provided that the combination of 3R , R7 and indicattj RCS total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable operati.on shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater tha,n or equal to 95% of RATED THERMAL POWER.

No' SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate and R , R, shall be 3

determined to be within the region of acceptable operation.of F1gurt 3.2-3:

a. Prior to operation above 75% of RATED THERMAL R0WER after each fuel loading, and *

b. At least once per 31 Effective Full Power ' days.

cr}.

4.2.3.3 The indicated RCS total flow rate shall be verified to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the most recently obtained values of R 1

.and 2R , btained per Specification 4.2.3.2, are assumed to exist.

aph 4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL

.'/ CALIBRATION at least once per 18 months.

4.2.3.5 The RCS total flow rate shall be determined by measurement at least once per 18 months.

SUMMER - UNIT 1 3/4 2-9 O

. _ , - - . _ , . . , , - . . _-. _ . . _ ,,_ _ ~_ -. _- -.__ _ _ _ _ ._9____ _ _ _ _ . _ .

I ATTAC10!ENT II REQUESTED CHANGES FOR VIRGIL C. SUMMER STATION TECHNICAL SPECIFICATIONS , 3.3.3.7 -

3.7.9.2 3.7.9.4 .

I 3.7.9.5 i

PAGES AFFECTED The following is a brief explanation for the requested changes to the Fire Protection Technical Specifications.

PAGE EXPLANATION 3/4 3-62 The corrections to the Fire Detection Instruments.

thru Table 3.3-11 are necessary to reflect changes 3/4 3-66d in che installation.

3/4 7-28 Elevations have been eliminated from the

,. identification of the spray and sprinkler system because they do not accurately describe areas protected. The name is sufficient to identify the system.

3/4 7-32 Additional fire hose stations have been added to Table 3.7-5 to ensure adequate fire protection of safety related equipment.

3/4 7-34 An additional yard fire hydrant and hose house have been added to Table'3.7-6 to ensure adequate fire protection for the diesel generators.

l l

1 l

l t

(

{,. TABLE 3.3-11 .

FIRE DETECTION INSTRUMENTS MINIMUM INSTRU- TOTAL NUMBER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS

.b HEAT SM0KE HEAT SMOKE

1. REACTOR BUILDING ' -

Zone 1 Elev. 412' Room 12-01 NE 2 3 Room 12-01 SE 2 3 Room 12-01 SW 2 3 Room 12-01 NW 2 3 Zone 2 Elev. e 436 Room 12-03 1 2 Room 12-07 1 2 Room 12-08 1 2 Zone 3 Elev. 436'

Room 36-01 NE 2 .

3 Room 36-01 SE 2 3 Room 36-01 NW 2

3

. Room 36-01 SW 2 , 3

"'? Zone 4 Elev. 462'

'I Room 12-03 1 2 Room 12-07 1 2 l

Room 12-08 1 2 Zone 000 Elev. 4tf" MF Room 12-08 1 2 Zone 000 Elev. 4GY S/4 Roc'1 63-01 SE 2 3 Room 63-01 SW 1 2 Zone PPP Elev. -+te/95 i Room 12-03 1 2 l

(- Room I?-07 1 2 Zone PPP Elev. 4&F S/Y '

Room 63-01 NE 1

11 Room 63-01 NW 3 65.#

2. CONTROL BUILDING F

fZoneAElev.463' -

- b'f l tuum C2 02 1** 1**

Room 63-03 S t**

"f/ Room 63-04

// s**

1** 2**

Pa3e (

SUMMER - UNIT 1 3/4 3-62 1

1 l

TABLE 3.3-11 (Cont.) '9-

_F/RE DE7A: c r9nA/ TA/S7'R/tM t A/ 7's.

MINIMUM INSTRU- TOTAL :: UMBER I!iSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS HEAT SMOKE HEAT SMOKE

2. CONTROL BUILDING (Cont.) ("n -"

Zone A Elev. 448' '

Room 48-02 (Cable Spreading Room-Upper) 17** 18**

Zone A Elev. 436'

{.3 Ree... OC 02 1" 1:- V Room 36-03 S 1*

J/. 3,**

Roo.n 36-04 1**

' ' 2**

' Zone B Elev. 425' Room 25-d2 (Cable Spreading Room-Lower) 10** 11**

Zone B Elev 412'K ** **

Room 12-hgum ,., /2-03 12-04 M ; /*"*" 23+,* *,

J '

Zone B Elev. 400' #

.ge.1.e7Roem1

--- pooo oo4G-05,,0,0-

- 01 .

1**** I 1**

i

/**

one J Elev 436' g;, .

Room 36-01 2**

Room 36~O2 2"# #

3*,

f Zone J Elev. 448' Room 48-01 21** 22**

Zone K Elev. 412' Room 12-02 2** 3**

Oc c.. l'00 3** 4" Zone K E1.v. 425'

f3 2,5j,g,1, ,, 2 2,*,*,

Rum 2S*0+

Zone L Elev. 436' 23*4 4 l 2***

Room 36-11 Zone M Elev. 436' 4T ' ~ 4 G48**- 7

(

Room 36-10 -M**** 1 F 2 Zone N Elev. 448' Room 48-02 (Cable Spreading Area For HVAC g Cabinets) / 3 -M, 25~31, {

Zone P Elev. 46L, 4 MCB g g.

Room 63-05 (Fain Control Board) 3 6 Zone R Elev. 463' .

Room 63-05 (HVAC Control Board)

/

1 1 {

Zorse O Elev. 4 63 j % *- en SUM. a UNIT Qa, 43-02 3/4 3-63

( c.ua i czev, Room 3 6-10 saw 2 cgg gne i

Room 3 4-11 2 noo gu nk

)

. kne R Elx 436' 2pco 3 9 1t Room 36-01 b,,a Z Elev. 43G, g g g /0 km 36-// .

C..- *,- TABLE 3.3-11 (Cont.)

l FIRE DETBCT/DA/ IA N 'TA lf M B N T s MINIMUM INSTRU- TOTAL NUM8ER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS

, HEAT SM0KE HEAT SM0KE

. h-)I 2. CONTROL BUILDING (Cont.)

Zone S Elev. 448' ', .

Room 48-02 (Cable Spreading Area g for HVA Cabinets) /2 1 23 M h Zone VV Elev. 436' Room 36-09 1 2 Zone W Elev. 463' Room 63-05 # #

4 8 Room 63-d6 ,- 1 1 Room 63-07 -

1 2 Room 63-09 1 1 Room 63-10 1 1 Room 63-11 1 . 1 Room 63-12 1 1 Room 63-13 -

1 1 r.

W.

I Zone XX Elev. 463' C- Room 63-14 1 1 Room 63-15 1 1 Room 63-16 1 1 Room 63-17 1 1 Room 63-18 1 2 Room 63-19 1 1 Room 63-20 1 1 Room 63-21 1. 1 Room 63-22 1 2 Room 63-23 1 1 Zone NNN Elev. 482' Room 82-01 4, 8, Room 82-02 4 7 Room 82-03 2 3 Room 82-04 1 2 Zone TTT Elev. 412' Room 12-11 s 1 2

5. ii;RTii F::4ETRATIO:s AC ESS ARCA Zvue EC Ebv. 430' R;;; 20-01 2 3 Zune Yi Ehv. 412' Rec... 12 01 # #

2 .,

SUMMER - UNIT 1 3/4 3-64 L------

l

.. )

.s ;

F l}P E TABLE 3.3-11 (Cont.)

D E7'A= c 7~JoM J'NSTRUMEN 73 S !

MINIMUM INSTRU- TOTAL NUMBER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS HEAT SMOKE HEAT SM0KE INTERMEDIATE BUILDING (Cont.) h#~5 J.% I EASTPENETRATIONACCNSSAREA !

Zone JJ.Elev. 412'

, Room 12-02 5, 10, ,.

i Zone LL Elev. 436' g Room 36-02 5 10,

! i y, K. WEST PENETRATION AREA '

l Zone Y Elev.3412' Room 12-01 ~

4, g

. 8 Zone DD Elev. 436' Room 36-01 ,

4, 8, l

Zona II Elev. 463' Room 63-01 :

2[ -

3 Room 63-03 ,y

f d [S. s.;

, S'. J/ . INTERMEDIATE BUILDING Zone AA Elev. 412'y,,, , q_ ,z is ** /6 "'

Room 12-62 N g,,,, , . a g /[. 22 e& d ' M *g* P s e Room 12-13 A{$e.ee iz-o9 la 44 lax l, 4

l Room 12-13 8 1* 1**

Room 12-13 C 1** 1**

l Zone BB Elev.112'# m a-o2 /0 d# // ""

Room 12-02 E (,-1y== 74gs=

Room 12-10 1** ina Zone CC Elev. 436' Room 36-02 WSW 3** 4**

Zone FF Elev. 423'-6" Room 236-01 3, 5, (.

Zone GG Elev. 412' Room 12-03 1 1 Room 12-04 1 1 Room 12-05 1 2 4 Room 12-06, Room 12-07 1

1 2

1 h.

Room 12-08 1 1 Room 12-12 1 1 Room 12-14 1 1 l Room 12-15 1 1 '. -

1 -

l l

SUMMER - UNIT 1 3/4 3-65 l

C'G FIRE TABLE 3.3-11 Cont.)

DETECTfDA/ (7A/WANMENTS MINIMUM INSTRU- TOTAL NUMBER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS HEAT SMOKE HEAT SMOKE

. .?

? 5". g. INTERMEDIATE BUILDING (Cont.)

Zone KK Elav. 412' ' -

Room 12-09 1 1 Zone KK $1ev. 423' g C:-c.

9 Room 23-01 .

2 4, Room 23-02 2, 4, Zone KK Elev. 436' Room 36-01 2 3 Zone MM Elev. 463' .

g g Room 63-01 4, 8, Room 63-02 3 6 Room 63-03 1 , 2 Zone PP Elev. 436' Room 36-02

8, 16,

~

. Room 36-03 (CREP) 1 2 Room 36-03 A (CREP) 1 2 Room 36-03 8 1 2 Room 36-04 1 1 l

Zone MMM Elev. 451' y Room 51-01 3 ,g 6, l Room 51-02 3 6 Room 51-03 1 2 Roogi 51-04 1 2 Zo'ne FFF Elev. 436' y Room 36-02 8, 16 Room 36-02 E 1 ,. 1 h Zone UUU Elev. 426' Room 26-01 2g y

, 4, y

Room 26-02 2 4 6.L AUXILIARY BUIDLING l

(- Zone Q Elev. 436' Room 36-18 2 y

4 y

Zone HH Elev. 463' Room 63-01 1 2 Room 63-04 1 1

{' Room 63-06 Room 63-17 1

1 2

1 l

SUMMER - UNIT 1 3/4 3-66

i i

f.}

TABLE 3.3-11 (Cont.) V.-

_ PIRE DETEC7/DN TNS7AMMEWr$

MINIMUM INSTRU- TOTAL NUMBER INSTP.UMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS HEAT SMOKE HEAT SMOKE .

6,T. AUXILIARY BUILDING (Cont.)

Zone QQ Elev. 452' Room 52-01 1 1 Room 52-02 1 1 Zone RR Elev. 463' Room 63-16 2** 3**

h Room 63-19 1** 1**

Zone ZZ Elev. 374' Room 74-12 1 1

. Room 74-16 - 2 3 Room 74-17 -

2 3 Zone AAA Elev. 388' Room 88-23 1 . 1 Room 88-24 1 1 Room 88-25 l 1 2 1

Zone AAA Elev. 397' Room 97-02 3, 6, ';

Room 97-02 N 1 1 Room 97-02 S 1 2 Zone BBB Elev. 388' Room 88-05 1 2 Room 88-13 N 2 3 Room 88-13 NE 1 1 l Rocm 88-13 5 1 1 1

Zone BB"B Elev. 397' .

Room 97-01 1 2 Zone BBB Elev. 400' Room 00-01 W

(

l 1 1 Room 00-02 E 2 3 Zone CCC Elev. 463' Room 63-09 2 3 l Room 63-14 1 2' Zone EEE Elev. 374' Room 74-01 1 2 (C ,

Room 74-07 1 2 1

Room 74-08 2g 3g Room 74-09 3 5 Room 74-18 1 1 Zone GGG Elev. 400' y y

Room 00-01 3 6 SUMMER - UNIT 1 3/4 3-66a

h*o

TABLE 3.3-11 N/RE hETscr/DW(Cont. )

INCTWRMEWTS MINIMUM INSTRU- TOTAL NUMBER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS HEAT SMOKE HEAT SMOKE Cw%'.6,X. AUXILIARY BUILDING (Cont.)

Zone HHH Elev. 412' ',

Room 12 0; 2 73 - 0# W l 1 Room 12-06 1, .

1,

.S " ~

h . 12-27 1 1 Zone HHH Elev. 425' p Room -2; C2 r 26-o/ .

S ,I 10 Room 26-d2 CW . 1 1 l

Room 26-02 S -

1 1 l Zone III Elev. 412' Room 12-02 1 .- 1 Room 12-03 A 1 1 -

Room 12-09

1 1

2. Room 12-11 N 1 1 5f]-

Roem 12-15 1 1 Room 12-18 1 1

( Room 12-28 1 1 l Room 12-31 1 1 Zone JJJ Elev. 412' Room 12-02 Ig ly Roon 12-11 4 7 Roqm 12-11 N 1 1 l

l Zone LLL Elev. 436' Room 36-33 N R**m 36-18 2[

g S

3 Zone RRR Elev. 388' Room 88-05 1 1 Room 88-13 1 2 Room 88-13 5 1 1 Room 88-16 1 1

/

Zor.e WW Elev. -4%- #48 y

Room M-s& fB-ol y 3 5 Zone WW Elev. 446' Room 46-01 1 1 he.

l SUMMER - UNIT 1 l

3/4 3-66b

Eons Y)! Elev, 4 /2 ' g 4 Roont 12 -O I 2 't TABLE 3.3-11 (Cont. 3 F/RE DFrEC T/oM IN MMENTS MINIMUM INSTRU- TOTAL NUMBER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS

~

HEAT SMOKE HEAT SMOKE 6,K. AUXILIARY BUILDING (Cont.) ,

G Zone YYY Elev. 412' Room.12-05 1 1 Room 12-27 . 1 1 Room 12-30 1 1 h

Zone ZZZ Elev. 4te'4fs#

Room &G-Ot f5-O/ 1 1 Zone ZZZ Eleir. 436' f Room 36-01 '

1 1 Room 36-03 1 2 Room 36-31 1 1 Room 36-33 S 1 . 2

%g. FUEL HANDLING BUILDING. .

Zone EE Elev. 436'

'{ Room 36-01 W x m 26-01 E 2

I 3

I o bl 1 1 Zone TT Elev. 463' Room M 43-01 Room 63-of N S -1d

.g= y jod*

lp Zone UU Elev. 463' '

Room 34=9k 63-0I 7 -1$p l Room Gs-et S 5 JS A*

q j 8.K. SERVICE WATER PUMPHOUSE Zone NN Elev. 425' '

l Room 25-04 Room 25-05 1, 2, {

3 6 l

Zone 00 Elev. 441' Room 41-02 1** 1**

Zone 00 Elev. 436' RoomM .96-01 8** 9**

Zone DDD Elev. 441' Room 41-01 3, 5, Room 41-01 M I I Zone VVV Elev. 425' ('~

\.

Room 25-03 1 1 SUMMER - UNIT 1 3/4 3-66c

f. ..

' V

>~1 R E TABLE 3.3-11 (WCont.T DETECr/O fNs 7'RuMEWTS MINIMUM INSTRU- TOTAL NUMBER INSTRUMENT LOCATION MENTS OPERABLE

OF INSTRUMENTS HEAT SMOKE HEAT SMOKE a.

9,.10. DIESEL GENERATOR BUILDING

~

Zone DG Elev. 436' Room 36-01 1** . 2** -

Room'36-02 1** 2**

?: Room 36-03 7 g** g ya*

J Room 36-04 7 z*n g g** - -

Zone OG Elev. 427' 1**

Room 27-03 1**

Room 27-04 1** 1**

~

Zone KKK Elev. 400' ~

Room 00-01 1 1 Room 00-02 1 1 Zone KKK Elev. 427' Room 27-01 1 1 Room 27-02 1 1

- l.f Room 27-03 2 3 Room 27-04 2 ,

3

The fire detection instruments located within the Reactor Building are not l required to be operable during performance of Type A Containment Leakage l

Rate Tests. -

- Automatically Actuates Preaction Sprinkler System.

- - Automatically Actuates Low Pressure CO System.

2 C

I G SUMMER - UNIT 1 3/4 3-66d

INSTRUMENTATION RADIDACTIVE LIOUID EFFLUENT M0NITORING INSTRUMENTATION .

h* ,

LIMITING CONDITION FOR OPERATION 3.3.3.8 The radioactive liquid effluent monitoring instrumentation channels .*

^

shown in Table 3.3-12 shal,1 be OPERABLE with their alarm / trip setpoints set to ensure that the limits of Specification 3.11.1.1 are not exceeded. The alarm /

trip setpoints of these channela shall be determined in accordance with the '

0FFSITE DOSE CALCULATION MANUAL (00CM).

APPLICABILITY: At all times.

ACTION:

a. With a# radioactive liquid effluent monitoring instrumentation channel alarm / trip setpoint less conservative than required by the above specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel or declare the channel inoperable. .

i b. With less than the minimum number of radioactive liquid effluent i monitoring instrumentation channels OPERABLE, take the ACTION shown

., in Table 3.3-12. Additionally if this condition prevails for more than 30 days, in the next semiannual effluent report, explain why this . r' l

condition was not corrected in a timely manner. -

c. The provisions of Specifications 6.9.1.12.b, 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.3.3.8.1 Each radioactive liquid effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-8. :.

i E

SUMMER - UNIT 1 3/4 3-67

I PLANT SYSTEMS l

C4 ^ SURVEILLANCE REOUIREMENTS (Continued)

\

1 4.7.9.1.2 The fire pump diesel engine shall be demonstrated OPERABLE:

h, a. At least once per 31 days by verifying:

1. The fuel storage tank contains at least 150 gallons of fuel, and '

2. The diesel starts from ambient conditions and operates for at

{,S least 30 minutes on recirculation flow.

b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-197!i, is within the acceptable limits specified in Table 1 of ASTM D975-1977 -

when checked for viscosity, water and sediment.

c. At least once per 18'mokths, during shutdown, by subjecting the

, diesel to an ' inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

4.7.9.1.3 The fire pump diesel starting 24-volt batter'y bank and charger

. . shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1. The electrolyte level of each battery is above the plates, and l

2. The overall battery voltage is greater than or equal to ,

l 24 volts. l I

b. At least once per 92 days by verifying that the specific gravity is !

appropriate for continued service of the battery.

c. At least once per 18 months by verifying that:

1

1. The batteries, cell plates and battery racks show no visual

{ indication of physical damage or abnormal deterioration, and j

2. The battery-to-battery and terminal connections are clean, l tight, free of corrosion and coated with anti-corrosion material. i l

b .

i l

l

{..

l l

SUMMER - UNIT 1 3/4 7-27 ,

l PLANT SYSTEMS SPRAY AND/0R SPRINKLER SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.9.2 The following spray and/or sprinkler systems shall be OPERABLE: .

a. Fuel Handling Building Charcoal Filter Plenums. L . "id;. Elc.. ""2' b

b. Control R om Emergency Charcoa.1 Eilter Plenums. Co..t. ; 0;d;.

u.... , m .,

c. Diesel. Fire Pump Room Wet P,ipe Sprinkler System. : . col ot-: . 3 Wete.

,_ .. e,.. ,,c

d. Diesel Generator ,,e.Building

.a m Pr.eaction Sprinkler System.--9irsci G na,o__

....ww.

r,_

E'My #W. idiw3'"' d CeLio j

e. CoLi Sr.fodi..3 Rw w.a. o. . Ci.oaca Preaction Sprinkler System,7 Cer.trel,0 ildir.; Cice. " C2 ' , ". '. 0 ' , " 25 ' , ' 25 ' , '12 ' ar.d 40 0 ' .

f. Service Water Pump House Preaction Sprinkler System. Cersice Wote.-

T w.uy ;;ww . C; . 430 ...d 441.

g. Intermediate Building Preaction Sprinkler System. I..te. diote Cw: Ji..g C1ei. " 12 . - .

h. Auxiliary Building .Preaction Sprinkler System. 400' [1;.. ;

, APPLICABILITY: Whenever equipment protected by the spray / sprinkler system is .

required to be OPERABLE.

g ACTION:

a. With one or more of the above required spray and/or sprinkler systems inoperable, within one hour establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol. Restore the system to OPERABLE status Within 14 days or, in lieu of any other report required by Specifica-tion 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status. _

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.9.2 Each of the above required spray and/ce sprinkler systems shall be demonstrated OPERABLE:

Q.

a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position.

b. At least once per 12 months by cycling each testable valve in the .

flow path through at least one complete cycle of full travel.

SUMMER - UNIT 1 3/4 7-2B 1

V .

PLANT SYSTEMS

FIRE HOSE STATIONS -

LIMITING CONDITION FOR OPERATION '

l l

... 3.7.9.4 The fire hose stations shown in Table 3.7-5 shall be 0PERABLE.

APPLICABILITY: Whenever equipment in the areas protected by the fire hose stations is required to be OPERABLE. .

~

ACTION: ,

7.. a.

With'one or more of the fire hose stations shown in Table 3.7-5 i W

inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye connected to the length of hose at the station, the other outlet of the wye connected to a hose of sufficient length to provide coverage for the area unprotected by the inoperable hose station. This shall be accomplished within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the inoperable fire hose is the primary means of" fire suppression; otherwise, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the fire hose ,

station to OPERABLE status within 14 days or, in lieu of any other report req'uired by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the.inoper-ability, and plans and schedule for restoring *the station to OPERABLE status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS -

! 4.7.9.4 Each of.the fire hose stations shown in Table 3.7-5 shall be demonstrated OPERABLE:

! a. At least once per 31 days by a visual inspection of the fire hose stations accessible during plant operation, to assure all required equipment is at the station.

b. At least once per 18 months by:

l. Removing the hose for inspection and re racking, and

',h 2. Inspecting all gaskets and replacing any degraded gaskets in the couplings. ,

3. Visual inspection of the fire hose stations not accessible during plant operation to assure all required equipment is at the station.

' ih . c. At least once per 3 years by:

1. Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage.

2. Conducting a. hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater. .

SUMMER - UNIT 1 3/4 7-31

i PLANT SYSTEMS h

TABLE 3.7-5 FIRE HOSE STATIONS REE4/R8CK LOCATION

ELEVATION HOSENAGK-IDENTIFICATION **

Auxiliary Building 374 4138, 4140, 4142 -

388 '

4144, 6764; 4/ 5*

6 412 -

6766, 6761, 3 436 6783, 6769, 4148, 4147 463 6784, 6755, 4149 397 4143 p.

Reactor Building 412 6778, 6780, 6776

.)

436 6777, 6782 463 6781, 6779 Fuel Handling Building 436 6802 -

463- 6804, 6807 Control Building 463 68093 6T88 482 6815, 6810 :

Intermediate Building 412 4128, 4121, 4122 436 4129, 4124, 4123 463 4130 (.

c ,dn I Bkn'iding 412. 4812] 48/3

425- Afoto?, 4049 4yg . 6808, 4 814 448 4068, 4070 4

^ List all Fire Hose Stations required to ensure the,0PERA8ILITY of safety related equipment.

l ** Identified by isolation valve number

. c SUMMER - UNIT 1 3/4 7-32

PLANT SYSTEMS YARD FIRE HYDRANTS AND HYDRANT HOSE HOUSES s

LIMITING CONDITION FOR OPERATION

. 3 07 .5. The yard fire hydrants and associated hydrant hose houses shown in

., Table 3.7-6 shall be.0DERABLE. $ .

APPLICABILITY: Whenever equipment in the areas protiected by the yard fire

, hydrants is, required to.be OPERABLE. '

, ~

ACTION:

h a. With one or more of the yard fire hydrants or associated hydrant

hose houses shown in Table 3.7-6 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months have sufficient additional lengths of 21/2 inch diameter hose located in an adjacent.0PERABLE hydrant hose house to provide service to the unprotected area (s) if the inoperable fire hydrant or associated hydrant' hose house is the primary means of fire suppression; otherwise, provide the additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the hydrant or hose house to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specifica-tion 6.9.2 within the next 30 days outlining the action taken, the

. cause of the inoperability, and the plans and. schedule for restoring the hydrant.or hose house to OPERABLE status.

b. The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REOUIREMENTS 4.7.9.5 Each of the yard fire hydrants and associated hydrant hose houses shown in Table 3.7-6 shall be demonstrated OPERABLE:

a. At least once per 31 days by visual inspection of the hydrant hose house to assure all required equipment is at the hose house.

b. At least once per 6 months (once during March, April or May and once during September, October or November) by visually inspecting each 1

, ( c.

yard fire hydrant and verifying that the hydrant barrel is dry and that the hydrant is not damaged.

At least once per 12 months by:

1. Conducting a hose hydrostatic test at a pressure of 150 psig or at least 50 psig above maximum fire main operating pressure, whichever is greater.

b 2. Inspecting all the gaskets and replacing any degraded gaskets in the couplings.

3. Performing a flow check of each hydrant to verify its OPERABILITY.

U SUMMER - UNIT 1 3/4 7-33 1

PLANT SYSTEMS fg.t

. TABLE 3.7-6 \?'

YARD FIRE HYDRANTS AND ASSOCIATED HYDRANT HOSE HOUSES LOCATION .

HYDRANT NUMBER

EcutFMENT ID.

Waf of Service Water Pumphouse 3 XFX- JN - FS h N:.+k

F De'esel Genenfoy Butidi,,y .

2 gyy_ y g _p$

~

l l

l Q:

^These hydrant numbers are the numbers physically indicated on the hydrant houses.

h e

SUMMER - UNIT 1 3/4 7-34

~

I ATTACIDENT III REQUESTED CHANGE FOR VIRGIL C. SUMMER STATION TECHNICAL SPECIFICATION 3.4.3 ,

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)

PAGES AFFECTED The following is a brief explanation for the requested change.

PAGE EXPLANATION 3/4 4-9 As described on Page 22-84 of the Safety Evaluation Report (NUREG-0717), the two backup pressurizer heater groups are supplied from separate independent emergency f,. busses backed by diesel generators. The power for the

,* backup heaters is always from the emergency power supply. This power supply cannot be transferred to a balance-of-plant (normal) power supply. Therefore, Surveillance Requirement 4.4.3.3 is not applicable for the Virgil C. Summer Nuclear Station. .

D e a

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REACTOR COOLANT SYSTEM 3/4.4.3 PRESSURIZER LIMITING CONDITION'FOR. OPERATION 3.4.3 The pressurizer shall be OPERABLE with a water volume of less than or equal to 1288 cubic feet,',(92% of indicated span) and at least two groups of pressurizer heaters each having a capacity of at least 125 kw.

APPLICABILITY: MODES 1, 2 and 3

~

ACTION: . :

a. With one group of pressurizer heaters inoperable, restore at least two groups 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 the' pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 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 .

ly

SURVEILLANCE REOUIREMENTS 4.4.3.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

( 4.4.3.2 The capacity of each of the above required groups of pressurizer

( heaters shall be verified by erergizing the heaters and measuring circuit current at least once per 92 days.

4.4.3.3 T;is ums , w e..m y mm. mpij fer th; pr;;nchcr h;;tcr; ;ha'i L;-

i ;n;tr:::d CPEMSLE t h::t :nce per I? enth: by m:ne:113 tr:n:ftr ' ; przer fro . th; c.;r ;l t; th; cmcr;;ncy p::::r :upply 2nd :n:rgizing th: h::t:r;.

C l

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SUMMER - UNIT 1 3/4 4-9 L ,

REACTOR COOLANT SYSTEM s.

3/4.4.4 RELIEF VALVES "

. LIMITING CONDITION TOR OPERATION 3.4.4 All power operated valves shall be OPERABLE. , relief valves ,(PO.RVs) and their associated block APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a.

h With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to OPERABLE status or close the associated block valve (se), and remove power from the block valve (s); otherwise, 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 COLD SHUTDOWN within the following,30> hours.

t, With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the block valve (s) to OPERABLE status or close the bicek valve (s) and remove power from the block valve (s); otheivise, be in at least HOT STANDBY within the next 6' hours 'and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . . .-

c. The provisions of Specification 3.0.4 are not applicable. @Y SURVEILLANCE REQUIREMENTS 4.4.4.1 In addition to the requirements of Specification 4.0.5, each PORV shall be demonstrated OPERABLE at least once per 18 months by:

a. Performance of a CHANNEL CALIBRATION and

b. Operating the valve through one complete cycle of full travel.

4.4.4.2 Each block valve shall be demonstrated OPERABLE at least once per (;g.

92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with the power removed in order to meet the W

requirements of 3.4.4.a or 3.4.4.b.

l 0

SUMMER - UNIT 1 3/4 4-10

ATTACHMENT IV REQUESTED CHANGE MR VIRGIL C. SIMIER STATION TECHNICAL SPECIFICATION 3.4.9.3 e

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PAGES AFFECTED The following is a brief explanation for the requested change.

PAGE EXPLANATION 3/4 4-34 This change is requested to avoid notifying the Commission and preparing a Special Report each t time a PORV is made inoperable in order to perform

, required surveillance tests or preventive maintenance.

It will have the desirable effect of limiting reporting to failures which render the cold overpressure system inoperable.

e I

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION hi) 3.4.9.2 The pressurizer temperature shall be limited to: .

a. A maximum heatup',of 100'F in a'y n one hour period,

b. A maximum cooldown of 200*F in any one hour period, and

c. A maximum auxiliary spray water temperature differential of 625*F.

APPLICABILITY: At all times. .

ACTION: .

With the pressurizer temperature l{mits in exce s of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out of-limit condition on the fracture toughness. properties of the pressurizer; determine that the pressurizer remains acceptable for continued operation 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 rJuce the pressurizer pressure to'less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOUIREMENTS 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation.

{

6 C

SUMMER - UNIT 1 3/4 4-33

.s REACTOR COOLANT SYSTEM

- {,

OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION

)

3.4.9.3 At least one of t'he 'following o' verpressure protection systems shall be OPERABLE:

a. Two power operated relief valves (PORVs) with a lift setting of less .....

than or equal to the maximum setpoint defined by Figure 3.4-4, or "

b. The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 2.7 square inches.

APPLICABILITY: MODE 4 when the temperature of any RCS cold leg is less than or equal to 300*F, MODE 5, and. MODE 6 with the reactor vessel head on.

he reasons other than sneveillance ACTION:

tesfly *

_ P ra.v en h've matos feneice

a. IntheeventeitherPORVbecoNsinoperable}notifytheCommission within 7 days. In the event both PORVs are inoperabl , ndtify the Commission within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In both cases a Special Report shall be prepared and submitted to the Commission pursuant to Specifica- 3':/..

tion 6.9.2 within 30 days. The report shall describe the cause of the inoperability, plans for restoring the valves to OPERABLE status and any corrective action necessary to prevent recurrence.

b. The provisions of Specification 3.0.4 are not applicable.

I 1

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l l C SUMMER - UNIT 1 3/4 4-34

i ,

ATTACID1EST V MISCELLANEOUS CORRECTIONS

- FOR VIRGIL C. SGDIER STATION TECHNICAL SPECIFICATIONS

I r .

PAGES AFFECTED The following is a brief explanation for each miscella.:eous corrections requested.

PAGE EXPLANATION 3/4 2-11 The rod bow penalty curve is plotted incorrectly.

The coordinate points are correct.

3/4 4-5 The number for this Specification should be 3.4.1.4.1.

3/4 8-18 Typographical error. Test setpoints for all three reactor coolant pump breakers are the same, 3960 amps.

3/4 9-9 ALTERATIONS -misspelled.

6-3 Typographical error. The correct title is Emergency Coordinator. Also AMD should be ADM as shown.

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0 5 10 15 20 25 30 33 -'

.F .

3 C' REGION AVERAGE BURNUP (10 MWD /MTU)

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Figure 3.2-4 Rod Bow Penalty as a Function of Durnup 25 i to I

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s

. .s POWER DISTRIBUTION LIMITS

~

3/4.2.4 QUADRANT POWER TILT RATIO (.'i LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

ACTION: ~

i

a. With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but less than or equal to 1.09:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either:

a) TheQUADRANTh0WERTILTRATIOisreducedtowithinitslimit, or .

," b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. i -

n

2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either: p;,s a) Reduce the QUADRANT POWER TILT RATIO to within its limit, or (.5/

~

b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

h.

4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95% [J;d or greater RATED THERMAL POWER. '"

See Special Test Exception 3.10.2.

/'P..

SUMMER - UNIT 1 3/4 2-12

REACTOR COOLANT SYSTEM h COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4,0 At least one ~ residual heat removal (RHR) loop shall be OPERABLE

};'. and in operation *, and either:

a. #

One additional RHR loop shall be OPERABLE , or

b. The secondary side water level of at least two steam generators shall be greater than 10 percent of wide range indication.

h h APPLICABILITY: MODE 5 with Reactor Coolant loops filled".

~

ACTION:

t a. With less than the above required loops OPERABLE and/or with less than the required steam generator level, immediately initiate corrective action to return the required loops to OPERABLE stat 6s or to restore the required level as soon as possible.

b. With no residual heat removal loop in operation, suspend all operations involving a reductior; in boron concen-tration of the Reactor Coolant Systea and immediately

initiate corrective action to return the required residual

, heat removal loop to operation.

SURVEILLANCE RE0VIREMENTS 4.431.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.4.1.2 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0ne residual heat removal loop may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for g., surveillance testing provided the other RHR loop is OPERABLE and in "7 operation.

A Reactor Coolant pump shall not be started with one or more of the Reactor Coolant System cold leg temperatures less than or equal to 300 F unless

~1) the pressurizer water volume is less than 1288 cubic feet and/or 2) the f ..

sect,ndary water temperature of each steam generator is less than 50 F above QO eac1 of the Reactor Coolant System cold leg temperatures.

A The RHR pump may be de energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron cont entration, and 2) core outlet temperature is maintained at least 10 F 3 below saturation temperature.

SUMMER - UNIT 1 3/4 4-5

REACTOR COOLANT SYSTEM COLD SHUT 00WN - LOOPS NOT FILLE 0 G

LIMITING CONDITION FOR OPERATION 4 3.4.1.4.2 Two residual heat removal (RHR) loops shall be OPERABLE # and at least one RHR loop shall be in operation."

APPLICABILITY: MODE 5 with Reactor Coolant loops not filled.

ACTION: .,

O 2

a. With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible.

b. With no RHR loop in opsration, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation.

.e e .*

SURVEILLANCE REOUIREMENTS 4.4.1.4.2.1 At least one RHR loop shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

0ne RHR loop may be inoperable for up t 3 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing provided the other RHR loop is OPEMBLE and in operation. .

h.p A

The RHR pump may be de-energized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided 1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and 2) core outlet temperature is maintained at least 10 F below saturation temperature.

SUMMER - UNIT 1 3/4 4-6

ELECTRICAL POWER SYSTEMS SURVEILLANCE REOUIREMENTS (' Continued) -

(c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable T%~ type shall also be functionally tested until no more.-

failures are found or all circuit breakers of that type.

have been functionally tested.

2.

By selecting and functionally testisg.a representative sample h ,' .

of at least 10% of each type of lower voltage circuit breakers.

Circuit breakers selected for functional testing shall be -

selected on a rotating basis. Testing of these circuit breakers shal+ <:onsist of* injecting a current in excess of the breakers. '

nominal setpoint and measuring the response time. The measured response time will be comp.ared to the manufacturer's data to insure that it is less than or equal to a value specified by the manufacturer. Circuit breakers found inoperable during -

functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found - ~~

inoperable during these functional tests, an additional repre-sentative sample of at least 10% of all the circuit breakers of . ~ .

the inoperable type shall also be functionally tested until no .

more failures are found or all circuit breakers of that type

,. , have been functionally tested. .

3. 'By selecting and functionally testing.a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses shall include at least 10% of all fuses of that -

type.

The functional test shall consist of a non-destructive ~-

~

resistance measurement test which demonstrates that the fuse -

meets its manufacturer's design criteria. Fuses found inoperable l ' during these functional tests shall be replaced with OPERABLE .

fuses. prior to resuming operation. For each fuse found inoperable i .

during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be l

.] .

~

functionally tested until no more failures are found or all

. fuses of that type have been functionally tested.

b. .At least once per 60 months by subjecting each circuit breaker to an

' inspection and preventive maintenance in accordance with procedures prepared in conjunction with its m,anufacturer's recommendations.

~: . .

SUMMER - UNIT 1 -

3/4 8-17 '

a

TABLE 3.8-1

- '. g .. '

W t i CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICE TEST SETPOINT CRIT C .

5 TEST SETPOINT. RESPONSE TIME EQUIP N0.-SYS/0ESCRIPTION DEVICE LOCATION

.s...~ '

7.2 KV_Swgr.

.I PRIMARY XSW1A #9 LONG TIME l 3960 Amps < 15.75 Sec.

XPP0030A-RC 1)' Reactor Coolant Pump A INSTANT 5808 Amps N/A

. i-

- GROUND INST. 11 Amps N/A .

t XSW1A #5- LONG TIME 5544 Amps < 15.33 Sec.

BUSIA Normal Feed BACKUP BUS 1A Emergency Feed BACKUP XSW1A #3 LONG TIME < 5544 Amps 315.33Sec.

XPP0030B-RC PRIMARY XSW1B #7 LONG TIME Amps < 15.75 Sec.

w 2) 5808 Amps N/A

> Reactor Coolant Pump B INSTANT GROUND INST. 11 Amps . N/A

= '

b

~

LONG TIME f,544 Amps < 15.33 Sec.

BUS 1B Normal feed BACKUP XSW18 #5 BUS 1B Emergency Feed ' BACKUP , . XSW1B #3 LONG TIME 5544 Amps 315.33Sec.

~

PRIMARY. XSW1C #3 LONG TIME 39604mps < 15.75 Sec.

3) XPP0030C-RC Reactor Coolant Pump C .

INSTANT 5808 Amps N/A

~' GROUND INST.11 Am'ps N/A BUSIC Normal Feed BACKUP XSW1C #9 LONG TIME 5544 Amps < 15.33 Sec.

BUSIC Emergency Feed BACKUP XSW1C #13 LONG TIME 5544 Amps $15.33Sec.

U. -

=

g -

g g.

e

+. . ,. . - .,

Q O O u -

.0. O.- .

6< -

. REFUELING OPERATIONS

. s : . .-

3/4.9.8 REACTOR BUILGING FURGE SUPPLY AND EXHAUST ISCLATIO.'i S STE4

(,; LIMITING CONDITION FOR OPERATION

3.9.8 The Reactor Building Purge Supply and Exhaust Isolation System styall be OPERABLE.

('

APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. '

. ACTION: *

~

h 'With the Reactor Building Purge Supply and E'xhaust Isolation System inoperable,'

close each of the Purge and Exhaust penetrations providing direct access from ' "

the reactor building..atmospheresto the outside atmosphere. 'The provisions o'f

,Specificatic.n 3.0.4 are not applicable. . .

,. . S' UitVEILLANCE RE0_UIREMENTS .

~

4.9.8 The Reactor Building Purge Supply and Exhaust Isolation System shall be .

demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once .

per.7 days during CORE.ALTEPXATIONS by verifying that Reactor Building Purge t Supply and Exhaust isolation occurs on manual initiation and on a high radiation test signal from each of the containment radiation monitoring instrumentation

. channels, and by verifying .that isolation occurs on the 36-inch lines of the -

.p.

V Purge' Supply and Exhaust Isolation System on a high radiation test signal from -

, the reactor building. manipulator crane area channels.

l Q;,.. , -

w . .

1

r. .

l

(.c ,

, SUMMER,- UNIT.1 i .'. ,. ,. ,, 3/4 9-9 .. Amend' ment'No. 2

REFUELING OPERATIO'NS - -- - - V-3/4.9.9 VATER LEVEL - REFUELING CAVITY AND FUEL TRANSFER CANAL '

' ~

, LIMITING CONDITION FOR OPERATION -

., y. w 3.9.9 At least 23 feet of water shall be maintaiced over the top of the

, reactor pressure vessel flange. - - .

h.'

~

APPLICABILITY: During movement of fuel assemblies or control rods within the reactor pressure vessel or the refueling cavity when either +.he fuel assemblies being moved or the fuel. assemblies seated within the reactor ;.ressure vessel *

. are irradiated. ,. -

ACTION: -

W.ith the requirements of the above specification not satisfied, suspend all. -

. operations involving-movement of fuel assemblies or control rods within the pressure vessel. . . .

.k u: ..

SURVEILLANCE ilEOUIREMENTS - .

. 4.9.9 The water level shall be determined to be at least its minimum re, quired' .

. depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during mov.ement of fuel assemblies or control rods. - -

, g-Q. -

9

- ' 3-W,.

' ~

SUMMER - UN,I.T_.l.. . 3/4 9 .-

,' .: . ;j i . . _. .

i -)

3. -

m

ru-- .. - .) ,'

(:

ANAGln ~ is eg V. C. SUMMER NUCE E AR STATION - e LA g DEPUTY M ANAGER s1 g r-- - - _ _ _ m M I DIRECTOR l g

=====M STATION l 1 OuALitY CONrROL l E

_ c____ L_____J DIRECTOR OF I~~~~7

! SURV(tL L ANCE }= = = = = = * ,_____

SECURITV g SysitMS l SUPE RVISOR P l L____J L____d 1

1 I I I ASSISTANT MGR A338sTANT MGR ASSISTANT MGR. AS$1ST ANT MGR OPE R ATIONS M AINT SERvtOES TECHNICAL SUPPORT SUPPOttT SE nvsCE S SUPE RVISOR OF OPE R A TION S I I

= MA%TENANCE FIRE PL %NER OsdECTOR DIRECTOR DIRECTOR PROF f CTION OF 00 DINATOR CHLMtSTRY SITE ENGR. TECH SERV EA4 A E d N .

1 I l,D COOHD SHsFT WE LDING j g suPr Rvi$ ors suP*RvesOR P CHEM 6STRY ENGINEERS m M AINT E N A NC E SUPVS. AND l ~

SPECLS.

I E NGIN E E R CONTROL ROOM '

f0REMIN l l l ]

CHE MISTRV 1 SPECLS. STA.NUC. SHIFT TECH COMP. IECH SERV ENGR ADV. SUPV. COORD REACTOR MEL a (Lic.

l l l l OPERATOR $ SUPV Se' i l I I I I l l I TCCHNICAL PERSONNEL SUPV TEST DIR. OF DsR OF OsR SCHED AS SIST A N T MECH ELEC COORD. HP mese. OUTAGE OPE R ATORS PE RSONNE L PE RSON Nf L A D A%

I I I i TECH. H P. ADMIN SCHED b AU RILI ARY 1 1 PE RS. SUPVS SUDGET OUT.PERS OPER ATOR$ RECOROS tbC MAftRIALS I

$UPV EUPV l 5

a*

I I EPE CL S.

Ibc STORESb r PERSONNEL PROCUREMENT PERSONNEL ,

Figure 6.2 2 i

Virgil C. Summer NuCtear Station Functional Organisation .

TABLE 6.2-1 v

. MINIMUM SHIFT CREW COMPOSITION SUMMER UNIT 1 .

POSITION NUMBER OF INDIVIDUALS REOUIRED TO FILL POSITION h3 MODES 1, 2, 3, & 4 MODES 5 & 6 SS 1 1 CRF 1 None R0 A0 STA

.2 2

1 1 h 1 None SS - Shift Supervisor with a Senior Reactor Operators License on Unit 1 CRF - Control Room Foreman with a Senior Reactor Operators License on Unit 1 R0 - Individual with a Reactor Operators License on Unit 1 A0 - Auxiliary Operator .

STA - Shift Technical Advisor Except for the Shift Supervisor, the Shift Crew Composition may be one,less than the minimum requirements of Table 6.2-1 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the Shift Crew /A.

Composition to within the minimum requirements of Table 6.2-1. This provision QW does not permit any shift crew position to be unmanned upon shift change due to an onccming shift crewman being late or absent.

During any absence"of the Control Room Foreman from the Control Room while the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shift Technical Advisor) with a valid SRO license shall be designated to assume the Control Room ccmmand function. During any absence of the Shift Supervisor from the Control Room while the unit is in MODE 5 or 6, an individual with a valid R0 or SR0 license shall be designated to assume the Control Room command function. ,

SUMMER - UNIT 1 6-4'

=

a e

ATTACHMENT VI REQUESTED CHANGE FOR VIRGIL C. SUMMER STATION TECHNICAL SPECIFICATION 3.6.4

i PAGES AFFECTED The following is a brief explanation for the requested changes to Table 3.6-1 of Technical Specification 3.6.4 These changes exclude 14 additional contain-ment isolation valves from Type C testing. Also enclosed is a corresponding change to Table 6.2-53a of the FSAR which will be included in the next amendment.

These changes have been discussed with the NRC reviewer, Mr. Hearn.

PAGE EXPLANATION 3/4 6-20a It is requested that valves in the reactor coolant pump seal injection lines be excluded from Type C testing because these lines are not isolated during an accident and remain filled and pressurized when-ever there is water in the reactor coolant system.

3/4 6-20b It is requested that the eight motor operated valves located outside containment on the safety injection lines be excluded from Type C testing because in each line there are redundant check valves in series inside containment to prevent leakage and during an accident the lines remain filled and pressurized by the safety injection pumps.

3/4 6-20c See explanation for page 3/4 6-20a.

TABLE 3.6-1 (Continued)

E

l C0tlTAltiMENT IS01ATI0tl VALVES eti ItAXIMUM 7 VALVE flUMBER REAC10R BUILDING PURGE SUPPLY.

ISOLATION TIME c C.

= AttD EXilAUST ISOLATI0ft (Continued) FUNCTI0tt (SEC)

[ 4. 00028-All Reactor Building Purge Exhaust 5

5. 6056-liR Alternate Reactor Building Purge Supply Line 5

6. 6057-IIR Alternate Reactor Building Purge Supply Line 5

7. 6066-ilR Alternate Reactor Building Purge Exhaust Line 5

8. 6067-IIR Alternate Reactor Building Purge Exhaust Line 5 D. MANUAL (1)

1. 8767-DN Demineralized Water Line N/A

2. 8768-DN Demineralized Water Line -

N/A

3. 6772-FS Fire Service llose Reel Supply . H/A

4. 6773-FS Fire Service llose Reel Supply' N/A g 5. 2679-IA Breathing Air Supply Line ll/A

= 6. 2600-IA Breathing Air Supply Line N/A p 7. 6587-NG Nitrogen Supply To Steam Generators -

fl/A g 8. 8090A-RC Dead Weight Tester N/A

.> 9. 80908-RC Dead Weight Tester il/A Reactor Building Service Air

10. 2912-SA N/A

11. 6671-SF Refueling Cavity Drain Line N/A

12. 6672-SF Refueling Cavity Drain Line ll/A

13. 6697-SF Refueling Cavity Fill Line il/A

14. 6698-SF Refueling Cavity Fill Line -

il/A

15. 7135-WL Reactor Coolant Drain Tank Discharge To Waste N/A E. REMOTE MANUAL (2)

1. 9602-CC Component Cooling To R. C. Pumps . II/A

2. 8102A-CS # Seal Injection To Reactor Coolant Pump A N/A

3. 81028-CS -l# Seal Injection To Reactor Coolant Pump B fl/A

4. 8102C-CS .g$:. Seal Injection To Reactor Coolant Pump C il/A e o e .

@ e e @

@ e e w TABLE 3.6-1 (Continued) o o e- '

E 3 CONTAINMEllT ISOLATION VALVES HAXIMUM c- VALVE flUMBER ISOLATION TIME j*i E. REMOTE 11ANUAL (Continued) FUNCTION (SEC)

,_. 5. 8107-CS Charging Line To Regenerative lleat Exchange il/A

6. 60508-ilR liydrogen Analyzer Return Line il/A

7. 6051A-IIR llydrogen Analyzer Supply Line il/A

8. 60518-ilR llydrogen Analyzer Supply Line N/A

9. 6051C-IIR llydrogen Analyzer Supply Line il/A

10. 6052A-IIR liydro0en Analyzer Return Line fl/A

11. 60528-ilR liydrogen Analyzer Return Line il/A

12. 6053A-IIR llydrogen Analyzer Supply Line N/A .

13. 6053B-flR llydroGen Analyzer Supply Line N/A

14. 8701A-Ril RilR Pump Suction From Reactor Coolant Loop A N/A

15. 87018-Ril HilR Pump Suction From Reactor ~ Coolant Loop C N/A w 16. 8801A-SI $1- Boran Injection Tank To Reactor Coolant Loops N/A h 17. 88010-SI .$$: Boran Injection Tank To Reactor Coolant Loops il/A m 18. 8811A-SI RilR Pump A Suction From Recirculation Sump N/A 4 19. 88118-SI RilR Pump B Suction From Recirculation Sump N/A

@ 20. 8884-SI $ liigh llead Safety Injection To Reactor Coolant Loops N/A

21. 8085-SI sc liigh Ilead Safety Injection To Reactor Coolant Loops N/A

22. 8886-51 C liigh llead Safety Injection To Reactor Coolant Loops N/A

23. 8888A-SI Low IIcad Safety Injection To Reactor Coolant Loops N/A

24. 88888-51 @ Low Ilead Safety Injection To Reactor Coolant Loops N/A

25. 8889-SI -$- Low Ilead Safety Injection To Reactor Coolant Loops - it/A

26. 3003A-SP Supply To Reactor Building Spray Nozzles N/A

27. 30038-SP Supply To Reactor Building Spray Nozzles il/A

28. 3004A-SP Spray Pump A Suction From Recirculation Sump N/A

29. 30043-SP Spray Pump B Suction From Recirculation Sump N/A

30. 3103A-SW Service Water From Reactor Building CoolinD Unit A N/A

31. 3103B-SW Service Water From Reactor Building Cooling Unit B N/A

32. 3106A-SW Service Water To Reactor Building Cooling Unit A N/A

33. 3106B-SW Service Water To Reactor Building Cooling Unit B N/A

34. 3110A-SW Service Water To Reactor Building Cooling Unit A N/A

35. 3110B-SW Service Water To Reactor Building Cooling Unit B N/A i

TABLE 3.6-1 (Continued)

E 3 C0flTAltlMEllT ISOLATION VALVES 9

MAXIMUM e VALVE P1UNDER ISOLATI0ll TIME

{

e F. CllECK

1. 7541-AC FUtiCTION CRDM Coolant Water Inlet Line (SEC) il/A

2. 7544-AC CROM Coolant Water Outlet Line ,

il/A

3. 9570-CC Component Cooling To R. C. Pump Bearings il/A

4. 9689-CC Component Cooling From R. C. Pump Bearings il/A

5. 8103-CS Reactor Coolant Pump Seal Water Return N/A

6. 8368A-CS @ -

Seal Injection To R. C. Pump A il/A

7. 83688-CS # Seal Injection To R. C. Pump B ti/A

8. 8368C-CS $ Seal Injection To R. C. Pump C N/A

9. 8381-CS Charging Line To Regenerative lleat Exchanger N/A

10. 6799-FS Fire Service Delu0e To Charc;oal Filters N/A

11. 2661-IA Instrument Air Supply To Reactor Building fl/A w 12. 6588-l1G tiitrogen Supply To Steam Generators N/A A 13. 8046-RC Pressurizer Relier Tank Makeup Water Line il/A m 14. 2913-SA Service Air Supply To Reactor Building il/A A 15. 3009A-SP Supply To Reactor Building Spray flozzles N/A R 16. 30098-SP Supply To Reactor Building Spray Nozzles N/A Accumlator Nitrogen Supply

17. 8947-SI N/A

18. 8861-SI Fill Line To Accumulators N/A Valve not subject to Type "C" leakage test.

(1) Manual valves may be opened on an intermittent basis under administrative control.

(2) Remote manual valve positions are maintained by administrative control.

l

@ e e @ e o e

I 3 */

AMENDMENT e

. . . . _ _ _ _ , o ,,

TABLE 6.2-53a (Continued)

( CONTAINMENT ISOLATION VALVES St'BJECTED TO TYPE C LEAXACE TISTS Type C Leaksee Test Penetratio.

Nur.be r Service Svstem Inside Contairzent Outside Contairment I2)

[ 210 O) Reactor Building laak LR No( } No s Kate Test Pressure Sensing Line 211(7) Reactor Building Leak LR No( No I2} 22 Rate Test Blowdown I

a Reactor Building 1Aak LR No II No I2) 212 Rate Test Bloudew Emergenev Feedvater C EP N/A No 3 213 214 Spare - N/A N/A 215 Spare - N/A N/A Reactor Building 1.eak LR No I2} No I2) 22 216

. [' Rate Test Pressurization

Line 217 Reactor Building Pressure - iA N/A Sensing 216 Spare - N/A N/A 219 Steam Generator Elowdewn BD N/A No Loop C (4) N/A sc I3}

220 Stea= Generator Blowdom SS I too; C Sampling Line

8 N*

221 (7) Seal Injection to Reacter CS M* 4ee 3y Coolan t Puer Loop C M 222 (7) High Head Safety Injection St NO(5) #*

h (6)

J to Reactor Coolant Loops 223I ') Sampling Line f rom Feactor Coolant Loop C SS Yes Yes l5 v

6.2-206b

TABLE 6.2-533 (Continued) "J y CONTAINMENT ISC:.ATICH VALVES SUBJICTID TO IYPE C IIAKAGI TESTS II7EITE , 1M Type C Leabre Test i Pcnetration Ntche r Servies System Inside containwnt outside Contair-*nt 224 Steam Generator Blow- BD N/A No I3) down Loop B

225 Steam Generator Blow- SS N/A No I) down Loop B Sampling Line 226.(7) Residual Heat Removal RH Yes N/A Pump Suetion from Reactor Coolant Loop C Q Q M 227 (7) Lov Bead Safety injection SI Eas= h (5) 21

! to Reactor Coolant Loops 228 Spare -

i N/A N/A ,

h.N N# M 229 (7) Seal Injectien to Reactor CS ,

% -C-O-Coolant Pu=p Loop B W" seaster Va==cl M'. WA N/A Level Sesfog L'ee 1308 Reacter vess..J M w/A Level Semina une 19 uce neu e,r v ,,i ac "I^ %

Level Scrutng U ne nops t Spare - W/A N /A 231I ') Denineralized Water DN -

Yes Yes l5

~

232 Spare - '

N/A N /A 301A H2 Analyzer Supply HR Yes Yes ,

B H Analyzer Di; charge ER Yes Yes ;

302 b) Atcorn,c ,n, s, n yes its Perpt Lint a

6.2-1 66s

F 3'i AMENDMENT : :::::, ^^;

TABLE 6.2-53a (Continued)

(# OTTA11estrf TSot.ATION VALWS St BJECTED YO TYPE C iIAKACE TESTS P m tration Type C leakage Test Numbe r Services System fa mide Containment Out side Containment

. e

+ 322 low Head Safety Injection St No(5) NO x (6) W IM .

to Reactor Coolant loops 323(') Accumulator Sampling Line SS , N/A Yes l5

' - 314 Breathing Air IA Yes Yes 23 315 ( 7 ) low Head Safety injection Sr . No(5) Na(6) u X yy to Reactor Coolant loop Not Legs ,

326 Steam Cenerator Blowdown BD N/A No II}

loop A 327b) Spray Pu=p A Suction f rom SP N/A Yes Reactor Building Recircu-7

( lation Su=p 328 Spray Pump F Suction f rom SP N/A Yes Beactor Building Facircu-lation Sump 32 9 Safety Injection Puce A SI N/A Yes Suc tion f rom Re ac tor Building Recirculation Sump 330(7} Ceeponen t Cooling Water CC Yes 7es 22 irom Beactor Coolant Pump .

v .

Bea ring s l

401 Supply to Reactor Building SP Yes Yes Spray Nozzles Train A v

402 (4 ) Reactor Building Furge AH Yes(6) Yes Supply 403 (7) Beactor Building Cooling SW N/A Yes

(. 11 nit B Supply v

404 N Fire Service Hose Feel FS Yes y'*

5 Supply

e M.

AMENDMENT &

( TABLE 6.2-534 (Continued) <Y- - -i .361 CONTAINMENT ISCLATIC': VALVES St2JECTED TO TYPE C IIAKACE TISTS Typ e C le alr a t e Te s t P;netration Fumber Service System .

Inside Containment outside containment w

405 Sampling Line f rom SS Yes Yes l5 Pre ssu rize r g) 22 406A Dead Veight Tester RC N/A Yes s

B hactor Building Pressure -

N/A N/A Sensing the 407A Radiation Monitor Supply SS Yes Yes B(4) and Return ,

YO HO f 3I 406 3eal Injection to Reactor CS .2i.e,. ai Coolant Pmp Loop A 409 Charging Line to Pagenera- CS Yes Yes 22

('

4, tive Heat Exchanger 410 (7) Pasetor Coolant Ptap Seal CS Yes Yes Vater Return 411 b) Steam Generater Blowdown SS N/A No S)

Loop 4 Sarpling 412(7) High Head Safety Injection SI Mo(5) N#(6)

.M.oo.

3y to Reacter Coolang Leops 413 Spare -

N/A N/A 1

'v 414 Spare -

N/A N/A 415 (7) High Nead Safety Injection $1 No(5) M8( 6)

.s yD to Raactor Coolant Loop v 416 Spare -

N/A N/A 417A.B.C Reactor Vessel Level RC N/A N/A Sensing i

417D Sample Return to SS Tes Yes 23 l PRT

.v ,

417E Spare - N/A N/A i 6. 2-2 06 f l

r AMENDMENT tT* N

O'?!!'"E'. , 1 CO-( YABLE 6.2-534 (Con tinue d)

CONTAINMENT ISCIATION VALVES SUBJ"CTED 10 TYPE C 1.EAKACE TESTS Pinetration ,

Type C 14akage Test Numbtr Service System Inside Containment Outside Containment 418 I) Reactor Coolant Drain WL Yes Yes l5

'Tartk to Vent Reader and H2 41t(') Refueling Cavity Drain SF Yes Yes l5 Line

~ 420I ') Pressure Relief Tank RC Yes Yes l5 421I ') Refueling Cavity Till Line SF Yes Yes l5 422I ') Pressure Relief Tank RC Yes Yes l$

Makeup 423I ') haetor coolant Drain WL Yes Yes l5 s Tank 1

424I ') Faactcf Building Su=p ND Yes Yes l5 Drain Safety injection Pump B SI N/A Yes 425 Suction fron F4 actor Building Facirculatica Sump 22 Baron Injection to SI No N 3 426 Reactor Coolant kops 427I ') Fire Service Deluge FS Yes Yes l5

~

Main Stean Loop A MS N/A No 428 Reactor Building - N/A N/A 500 Pressure Sensing 703 Peactor Building - N/A N/A Pressure Sensing 6.2-206

.

ML20027C552

Text

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