Text

Forwards marked-up Pages of Facility Draft Tech Specs & Bases (NUREG-1381) Provided W/Low Power OL
	ML20012F041
Person / Time
Site:	Comanche Peak
Issue date:	03/28/1990
From:	William Cahill, John Marshall; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	RTR-NUREG-1381 TXX-90061, NUDOCS 9004090358
	Download: ML20012F041 (29)
	v • d • e

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1-3 .

'- =M Log # TXX-90061 !

'L J File # 916

/, C C Ref. # 10CFR50.36 ,

l nlELECTRIC March 28, 1990 j i

, U. S. Nuclear Regulatory Commission l' Attn Document Control Desk Washington, D. C. 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NO. 50-445 TECHNICAL SPECIFICATION CHANGES

, REF: NRC Letter from Mr. C. I. Grimes to Mr. W. J. Cahill, Jr., dated September 12, 1989 Gentlemen Attached are the marked up pages of the draft CPSES Unit 1 Technical i

Specifications and BASES (NUREG 1381) provided with the low power operating license. ' These marked up pages reflect changes requested by TV Electric to be included into the full power operating license for CPSES Unit 1. Also attached is a description of these changes. FSAR changes associated with chapter 6,:" Administrative: Controls,"'of the Technical Specifications will be provided in a separate submittal.

! Sincerely.

William J. Cahill. Jr.

l l

By:

L J. S. Marshall Generic Licensing Manager MCP/vid Attachments c - Mr. R. D. Martin. Region IV Resident Inspectors. CPSES (3) 9004090358 900328 hDR ADOCK 05000445 PDC 400 N. Olive Dallas, Texas 75201 ggC 3lI _

I Attachment I to TXX-90061 - ;l

'Pagel'ofJ CPSES TS REVISION 1

!e i . [. ,

4 DETAILED DESCRIPTION Page 1- )

k' ,

l

- is' rese I (as amended) Stagt Description !

Table 2.2-1 2 Revise NIS Power Range "S" value in Table 2.2 1 sh 1 of

5 to allow calibration using the installed meters ,

instead of the DMN originally assumed in the daily ,

recalibration. '

Revisions t

The change in the "S" value reflects the uncertainty l associated with the front panel meters. Adequate allowance exists between the setpoint and the safety ;

analysis' limit so that only the "S" value needs change. l TS Change Request Number: TS 90008.1 ,

SER/SSER Impact: No '

t Table 2.2-1 2 Revise Steam Generator Level Low Low due to removal of '

! unnecessary uncertainties.

l Revisions

! Westinghouse has quantified the uncertainty due to the '

velocity head created by the fluid flowing past the lower NR Level tap. This uncertainty is no longer

( required in this setpoint calculation.

TS Change Request Number: TS-90008.2 SER/SSER Impact: No 3/4-1-8, 10 4 See Page No(s):3/4 5 7. 8. and 9 ,

Clarification: i l

Change wording to clearly express that this condition !

l 1s allowed by Specification 3.4.8.3 only when the- '

l Specification is not applicable. '

l TS Change Request Number: TS-90 012 SER/SSER Impact: No 3/4 1-11, 13 4 See Page No(s):3/4 5 1.10 and 3/4 7 5 I Clarification:

This change provides consistency between the words used in the Surveillance Requirements and the LCO. /

TS Change Request Number: TS90-013 '

Related SER Section: 16 ,

SER/SSER Impact: No .

3/4 1-20. 22 .

2 Implementation of Westinghouse recommended contpol rod wear mitigation techniques. -

Revision: .

/

IE Information Notice No. 87-19 notified Westinghouse f acilities of potential for perforation and cracking of RCCAs. The proposed changes to the Technical Specifications allows the implementation of ,

recommended wear mitigation techniques. Similar changes have been accepted by the NRC for the Wolf t , - . , . .

~ ~

2 L.p .'. ' Attach 5ent1.toTXX-90061<

Page. 2 of 4 1

" 'O m '

~*

s- L CPSES TS REVISION 1-

.) , DETAILED DESCRIPTION Page 2 ;

ze '

J. TS Page

< "' (as amended) Group bescrintion ;

Creek facility. Westinghouse has assessed the impact of the change for CPSES and has concluded that i s, the shutdown margin will not be violated and that the :

Y ' NTC values are boundad by values assumed in the safety !

o analyses. Additionally the effect on Fz and Axial

Offset is at most +0.5% relative to the all-rods out '

,', calculation (when the RCCAs are parked furthest in the core).

TS Change Request Number: TS90-007 :

?: '

SEk/SSER Impact: No- #

Table 3/4 3 27 2 Revise S/G Level Hi-Hi to include increased instrument uncertainty. !

Revision: '

Magnitude of the velocity head bias higher than the assumed value used in the S/G hi hi setpoint calculation Adequate allowance exists between safety analysis limit and setpoint so that only "Z" term affacted.

TS Change Request Number TS-90008.3 SER/SSER Impact: No Table 3/4 3-28 2 Revise Steam Generator Level low-low due to removal of I unnecessary uncertainties '!

Revision: ;

l Westinghouse has quantified the uncertainties due to :

velocity head created by the fluid flowing past the lower NR level tap. This oncertcinty is no longer ,

required in this setpoint calculation. .

TS Change Request Number: TS-90008.4 -

L SER/SSER Impact: No !

l ,

y ;

3/4:6-13 2 Added PORV exemption to the note of the LCO.

Revision: I Added to the LCO note the exemption of Specification 3.7.1.7 since these velves are adequately covered by .

Specification 3.7.1.7. This Ts consistent with Specifications 3.7.1.1. 3.7.1.5. and 3.7.1.6. all of which handle their respective valve (s) outside of ,

L Specification 3/4.6.3.

TS Change Request Number: TS-90-014 Related SER Section: 16 SER/SSER Impact: No l

i 3/4 7-24 2 See Page No(s):B 3/4 7-6 Revise Containment Bldg Area Temperature Monitoring.

Revision:

,. These areas do not have remote reading instrumentation p

IU

.y Attachment 1,to TXX-90061

' 'Page 3 o f 4 '

hg p

4 L .. .

'CPSES TS REVISION 1 DETAILED DESCRIPTION Page 3- !

l TS Page l

. las. amended) Group Descript1on i i

that would allow monitoring these specific areas. The .

Reactor Coolant Pipe' Penetration area has been ,

demonstrated by ' calculation to be;1ess that 200 F if the bulk average temperature of the containment is less !

than 120 F. The bulk area temperature is monitored by .

Specifications 3/4.6.1.5 and 3/4.7.10 Table 3.7 3 item j 6 (General Areas). The CRDM'Pintform Barrier area is

~

enveloped by the CRDM Shroud Exhaust. It has been :

demonstrated by calculation that'the platform barrier a will be less than 140 F when the shroud exhaust is less :

than 163 F. '

Reactor Cavity Exhaust Abnormal Conditions has been :

changed from 175 F to 190 F, which is consistent with !

the previous change in the normal conditions from 135 F !

to ISO F (an increase of 15 F based on the location of i temperature monitor). !

Additonal SSER #'s: 3,4,23

[

SSER Section numbers: 9.4 TS Change Request Number: TS-90 006 ;

Related SER Section: 9.4; SSER22 9.4 SER/SSER Impact: No ,

B-3/4 8-1 3 Clarifies the day fuel tank and fuel oil storage' tank ,.

fuel' oil volumes as specified in' Technical Specifica- '

tions 3,8.1.1 and 3.8.1.2.

Clarification:

s The volumes for fuel oil specified in Technical Speci- 'I fications 3.8.1.1 and 3.8.1.2 for the day fuel tank and fuel storage tank needed clarification since the combined volumes are used to meet the requirements for NRC Regulatory Guide 1.137, January 1978, for the '

minimum required on-site fuel oil storage capacity. ;

For purposes of definition, the Fuel Storage System at '

CPSES consists of the Fuel Oil Storage Tank and is I equivalent to the ANSI N195-1976 definition for supply L tank. The bases for the minimum cap city of the Fuel '

Storage System and the day fuel tank volumes is to meet the seven day on-site fuel oil storage capacity _

criteria requirements of NRC Regulatory Guide 1.137.

January 1978, and ANSI N195-1976. The minimum day fuel -

tank capacity also meets the requirements of 60 minutes c of diesel generator operation at continuous rating plus 10 percent margin as required by NRC Regulatory Guide 1.137. January 1978, and ANSI N195-1976. For added conservatism, the minimum volume required for 60 min-utes plus 10 percent is excluded from the minimum required volume to operate the diesel generator at rated capacity for seven days.

TS Change Request Number: TS-90-015 Related SER Section: 9.5.4: SSER22 9.5.4 SER/SSER Impact: No r

F Attachment 1"to IXX-90061

, ' ' 'Page 4 of 4 L 'L- .f !

,- J CPSES TS REVISION 1 l f[ l< DETAILED DESCRIFTION Page 4 !

o- :

1 TS 'Page '

(as amended) Group Descrintion '

7 s

' '< g 65,6 '

2 Revise SORC quorum and membership requirements. .

Revision i The SORC charter is to advise the Vice President.

Nuclear Operations on matters of nuclear safety. As -

such, the core of the committee is'being changed to those functions / disciplines which are involved in the i daily / routine operation of the plant. Also, the

membership limiting restriction has been removed.

thus permitting.the SORC membership, designated by j the VP, NO. to expand in expertise and background.

As described, the SORC quorum can not be less than what was originally accepted and actually becomes a larger group since the quorum will be based I on the membership designated by the VP. NO.

TS Change Request Number: TS90-005

Commitment Register Number: Y8-0366 & NL-2837

Related SSER Section: SSER22 13.4.1

,SER/SSER Impact: Yes i SSER-22. section 13.4.1. defines the SORC quorum as the Chairman or Vice-Chairman 'plus five members. This is incorrectly stated as the FSAR (A 76) lists the i SORC quorum as the Chairman (or Vice Chairman) and FOUR :

members. SORC membership, as descibed, is unchanged.

r i

1 l(

1, a

6 i

n TABLE 2.2-1^ .

REACTOR TRIP SYSTEN INSTALSENTATION TRIP SETPOINTS 'jg

.n h TOTAL SENSOR *E ALLOIMNCE ERROR "a FUNCTIONAL UNIT (TA) _

Z (5) TRIP SETPOINT ALLOWABLE VALUE

%{

, 1. Manual Reactor Trip M.A. N.A. N.A. N.A. N.A. Ero

2. Power Range, Neutron Flux *

~

a. High Setpoint 7.5 4.56 [I' 1109% of RTP* 1111.7% of RTP*

4 o

b. Low Setpoint 8.3 4.56 [ /' # 125% of RTP* $27.7% of RIP

j

3. Power Range, Neutron Flux, 1.6 0.5 0 $5% of RTP* with $6.3% of RTP* with High Positive Rate a time constant a time constant

>2 seconds 12 seconds-

'T 4. Power Range, Neutron flux, 1. 6 0.5 0 $5% of RTP* with High Negative Rate $6.3% of RTP* with._-

a time constant a time constant

12 seconds 12 seconds

5. Intermediate Ringe, 17.0 8.41 0 $25% of RTP* $31.5% of RTP*

Neutron Flux

6. Source Range, Neutron Flux 17.0 10.01 0 $105 cps $1.4 x 105 cps

7. Overteeperature M-16 5.8 3.65 1.2+0.8 III See Note 1 See Note 2

8. Overpower N-16 4.0 1.93 0 <115.1% of RTP*

1112% of RTP*

9. Pressurizer Pressure-Low 4.4 0.71 2.0 11880 psig >1863.6 psig

10. Pressurizer Pressure-liigh 7. 5 5.01 1.0 $2385 psig $2400.8 psig

RIP = RATED THERMAL POWER (1) 1.2% span for delta-T (RIDS) and 0.8% for iressurizer pressure.

_ __ .,,,n, - -- ,- - - - --_.- ~ - - ------n'--~ ' = ~~ ^-

' ~

-e n TABLE'2.2-1 (Continued) .

g REACIOR TRIP SYSILM INSIRtNENIAll0N IRIP SEIPOINIS ,W E E 3:

2 10fAL SENSOR **

Att0WANCE ERROR "i A

IUNCll0NAL UNil (IA) Z (S) IRIP SEIP0lNT AttOWABLE VALUE 3,3

e 11.- Pressurizer Water Level-Higt 8.0 2.18 2.0 <92% of instrument <93.9% of instrumest u span span _ ,

G o a 12. Reactor Coolant Flow-Low 2. 5 1.18 0.6 >90% of loop- >88.6% of loop g

,_. design flow ** design flow"* 7 99* :

97 0% 25,0 .t3,1 w

13. Steam Generator Water M g 2.0 >28 4 of. narrow >2 W of narrow $.

Level - Low-tow range instrument range instrument ~

span span

14. Undervoltage - Reactor 7. 7 0 .0 14830 volts- >4753 volts- ,

Coolant Pumps each bus _ each bus

15. Underfrequency - Reactor 4.4 0 0 357.2 Hz >57.1 Hz. -

Coolant Pumps -

16. Turbine Trip a ., Low Trip System Pressure N.A. N.A. N.A. >59 psig >46.6 psig

b. Turbine Stop Valve N.A. M.A. N.A. >1% open >l% open Closure

17. Safety Injection input N.A. N.A. - N.A. N.A. N.A.

from ESF 1

- Loop design flh = 95,700 gpe.

k

__ _ _ _ g- ~m 3

- > g- &- e a v- g s

% 9 y +r- --N~r w4- y- t -y--m="'M4 - -.p-w -"

w.-ga ww- r- 4,w-=q" --~ww rg e v=-e 'uv y-.+g -

'*--M -g-*w7we t g-+--*e-wm

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, 414a sin.o d Jt. fe TJ[Y- 9 dd 6 /

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' ILye. 3 c.( A 'l

. . . REACTIVITY CONTROL SYSTEMS 1 FLOW PATHS OPERATING

' J LIMITING CONDITION FOR OPERATION !

i 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE: '

a. 'The flow path from the boric acid storage tanks via either a Doric '!

acid transfer pump or a gravity feed connection and a charging pump ;

to the Reactor Coolant System'(RCS), and

b. Two flow paths from the refueling water storage tank via centrifugal charging pumps to the RCS.

APPLICABILITY: MODES 1, 2, 3, and 4.* i ACTION: !

t With only one of the above required boron injection flow paths to the RCS :

OPERABLE, restore at least two boron injection flow paths to the RCS 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 STANOBY and borated to a SHUTDOWN MARGIN equivalent to at least 15 Ak/k at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS ,

l-4.1.2.2 At least two of the above required flow paths shall be demonstrated l OPERABLE: ;

a. At least once per 7 days by verifying that the temperature of the '

l , flow path from the boric acid storage tanks is greater than or equal to 65'F when it is a required water source;

b. At Yeast once per 31 days by verifying that each valve (manual, .

powe.r-operated, or automatic) in the flow path that is not locked. l sealed, or othenvise secured in position, is in its correct position; i- and

c. At least once per 18 months by verifying that the flow Dath required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

, pos app h es.ble-l- "A maximum of two charging pumps shal be OPERABLE whenever the temperature of one or more of the RCS cold legs i less than or equal to 350*F except

whe.a o - " by Specification 3.4.8, An inoperable pump may be energized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

COMANCHE PEAK - UNIT 1 3/4 1-8

- - . - - - -m

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

, ^ AtM k m wt 3.to.Vkx~ Host i

' ' % ~ af c4 a ai ;

.-, RFaCTIVITY CONTROL SYSTEMS

... i CHARGING PUMP 5 - OPERATING j l

LIMITING CONDITION FOR OPERATION 3.1.2.4 At least two centrifugal charging pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3*, and 4** ".

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to

~

i 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 STAN08Y anc borated to a !

.SHUTOOWN MARGIN equivalent to at least 1% ak/k at 200*F within the next :

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next ;

-7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

E SURVEILLANCE REQUIREMENTS 4.1.2.4.1 The required centrifugal charging pump (s) shall be demonstrated OPERABLE by testing pursuant to Specification 4.0.5.

l 4.1.2.4.2 The required positive displacement charging pump shall be demonstrated OPERABLE by testing pursuant to Specification 4.1.2.2.c.

[

4.1.2.4.3 Whenever the temperature of one or more of the Reactor Coolant System (RCS) cold legs is less than or equal to 350'F e:rghg pur;; :htM M OPE'A*LE, e.v. cept q v WM, a maximum pecificatian of 3 .t.a ? I.r 44 two -

I

w When required, one charging pump shall be demonstrated inoperable at least qqus. bll * .

y once per 31 days by verifying that the motor circuit breakers are secured in L the open position.

The provisions of Specifications 3.0.4 and 4.0.4 are not applicable for entry !

into MODES 3 and 4 for the charging pump declared inoperable pursuant to l Specification 3.1.2.4 provided the charging pump is restored to OPERABLE :

status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 3 or prior to the temperature of l one or more of the RCS cold legs exceeding 375'F, whichever comes first.

, **In MODE 4 the positive displacement pump may be used in lieu of one of the required centrifugal charging pumps. ,

1

An inoperable pump may be en'ergized for testing provided the discharge of the pump has been isolated from the RCS by a closed isolation valve (s) with l power removed from the valve operator (s) or by a manual isolation valve (s) I s.7 cured in the closed position.

COMANCHE PEAK - UNIT 1 3/4 1-10 l

l l

l

. g 4:m Aw A +. Tu-9es/ ;

, ' ' *Ap fetay. d

'. . '* REACTIVITY CONTROL SYSTEMS 3

BORATED WATER SOURCE - SHUTOOWN ;

e f

LIMITING CONDITION FOR OPERATION t

3.1.2.5 As a minimum, one of the following borated water sources shall be i OPERABLE: !

a. A boric acid storage tank with: ;

1) A minimum indicated borated water level of 10% when using the ;

boric acid transfer pump, ,

'I

2) A minimum indicated borated water level of 20% when using the gravity feed connection. '

3) A' minimum boron concentration of 7000 ppm, and f

l .4) A minimum solution temperature of 65'F. ;

) ,

E b. The refueling water storage tank (RWST) with; j

1) A minimum indicated borated water level of 24%,

2) A minimum boron concentration of 2000 ppe, and [

3) A minimum solution temperature of 40*F.

APPLICABILITY: MODES 5 and 6. i l

l< AC110N:

1 p

With no borated water source OPERA 8LE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

1 i SURVEILLANCE REQUIREMENTS ,

1 .

i 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Varifyir.g the indicated borated water Mand

3) Verifying the boric acid storage tank solution temperature when it is the source of borated water.

l

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it is the sour e of borated water and the outsiijle air temperature is l less than 40'F.

l 1

COMANCHE PEAK - UNIT 1 3/4 1-11 -

1

Avtahmt A 4 TA F a 98 d 61 -l

. 4 Lye'4 of A4 .

)

.;~ ,

REAC7tVITY CONTROL SYSTiMS j i

SURVE!LLANCE RE0V!REMENTS l

4.1". 2. 6 Each borated water source shall be demonstrated OPERABLE:

i

a. At;least once per 7. days.by: j 1

1

1) Verifying the boron concentration in the water, !

2) Verifying the indicated borated water v #of the water source, and

!?

['

3) Verifying the boric acid storage tank solution temperature when it'is the source of borated water, i

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when- >

the outside air temperature is either less than'40*F or greater than

.120*F. .

c' b

L 4

COMANCHE PEAK - UNIT 1 3/4 1-13 I i

\

, ,Am L M A %: Tx t- 4 0 6 t t .

,.' 'fspo 70434 i

S' ' REACTIVITY CONTROL SYSTEMS ,

"' ',' " SHUTOOWN R00 INSERTION LIMIT l LIMITING CONDITION FOR OPERATION i

'3.1.3.5 All shutdown rods shall be fully withdrawn *,

APPLICABILITY: MODES 1** and 2** # .

ACTION: *

, With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4.1.3.1.2, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either:

, a. Fully withdraw the rod, or i

b. Declare the rod'to be inoperable and apply Specification 3.1.3.1.1 r

SURVEILLANCE REQUIREMENTS 3

4.1.3.5 Each shutdown rod shall be determined to be fully withdrawn: ;

e. Within 15 minutes prior to withdrawal nf any ends in Cnntrol ,

Bank A, B, C, or 0 during an approach to reactor criticality, and t

b. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, t

t

Fully withdrawn shall be the condition where s'hutdown rods are at a position within the interval ~of > d i 231 steps withdrawn.

A12.

- See Special Test Exceptions pecifications 3.10.2 and 3.10.3.

With K,ff greater than or equal to 1.

C0MANCHE PEAK - UNIT 1 3/4 1-20 e , ,

F. . ,

. i

~'

. Attach:cnt 2 to TXX 90061 !

Page 8 o 24 J.X,$,22.'2, g j, g pgz l

,231 ii ...... . ... . ....

I6I .. . . ( N..)

see ;: _ _ _:_ a - < - . .

azz .

i Iano;. .:..... .. ...

l BANK B !

l (0, 16 4) '

1 160."- .

' - - i- -

g j (100,146) l .

l W : BANK C . .

-m :

1 120 h ' - - - - - - -

w i i

! . i .

60 V - - ~ + i i

BANK!D I!

(0,!49)'

^ ,;o !_. . ... .' . ..

g L

l (31,0) ,

i i i i ' i i '

O' i -'

0: 10 20 30 40 50 60 70 80 90 100 t l- PRECENT OF RATED THERMAL POWER o t 1 -

Fully withdrawn shall be the condition wh control rods are at a position within the interval of nd < 231 steps withdrawn. AA 1

FIGURE 3.1-1 ROD BANK INSERTION LIMITS VERSUS THERMAL POWER COMANCHE PEAK - UNIT 1 3/4 1-22 g

.: ; N

, . . , =

~ ' '

IABLE 3.3-3 (Continued) -

o -

I ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUDENIATION TRIP SEIP0lNis M 27 E

o . TOTAL stNsot ERROR

y o *-

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

[ IRIP SETPOINI Att0WABLE VALUE [E s ee

[ 4. Steam Line Isolation g i z .

Q a. Manual Initiation -

N.A. N.A. N.A. N.A. N.A.

w [*

b. Automatic Actuation Logic M.A. N.A. N.A. N.A. N.A. I and Actuation Relays - E

c. Containment Pressure--High-2 2. 7 0.71 1. 7 . 16.2 psig 16.8 psig

d. Steam Line Pressure--Low 17.3 15.01 2.0 >605 psig a >593.5 psig*

e. Steam Line Pressure - 8.0 0.5 0 -<100 psi"* $ 178.7 psi ** ~

, y Negative Rate--High

5. Turbine Trip and Feeduater l Isolation l

I

a. . Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays -

b. Steam Generator Water Level--High-High 7.6 g'h Y 2.0 182.4% of <84.3% of narrow narrow range range instrument instrument ~ - span.

Span.

c. Safety Injection See item 1. above for all Safety Injection Trip Setpoints and All(wable Values.

e - _ _ J_ _ -- - - -_ _ _ _ _'- _._w__----

__- -s__..- -_ _ _m._---s--_ -e__22N-_m.+z_.__

_ _ _ _ _ _ -_____.__--__a___.__m__w.e ._ . . _ m.___ __m_..um-.- -__----_m.-,=,..s_-.-s_.._m_..__m-.2-.

~. .-

n lAetLE 3.3-3 (Continued)

Q ,V y ENGINEERED SAFE 1Y FEATURES AC ~UAIION SYSTEN INSTRUpKMTATION 1 RIP SEIPOINTS n

I" E. ON.

m SENSOR E$e

,7, 10TA. .

ERROR -o3 R FUNCTIONAL UNIT AliOUANCE (TA) Z (S) TRIP SETPOINT ALLOWABLEVALUE(( *

c. 6. Auxiliary Feedwater o 1

-4

i. a. Automatic Actuation Logic N.A. M.A. N.A. N.A.

i ~ M.A. E 2

and Actuation Relays 4 ar0 22. 6 25 o5 2N % g

b. Steam Generator Water g M 2.0 > 28-6Y of > .264t of narrow $

Level--Low-Low narrow range range instrument instriment span.

span.

1:' c. Safety Injection - Start See tes 1. above for all Safety Injection Trip Setpoints and Notor Driven Pumps Allowable Values. .

y d. Loss-of-Offsite Power N. A. N.A. N.A. N.A. N.A.

e. Trip of All Nain Feedwater N.A. N.A. N.A. M.A. M.A.

Pumps

7. Automatic Initiation of ECCS Switchover to Containment Sump '

a. Automatic Actuation Logic N.A. N.A. N. A. M.A. N.A.

and Actuation Relays

b. RWST Level--Low-Low 2. 5 0.71 1.25 > 40.0% of 1 38.9% of span span Coincident With See item 1. above for all Safety Injection Trip Setpoints and Safety injection Allowable Values.

8. loss of Power (6.9 kV & 450 V Safeguards System Undervoltage)

a. 6.9 kV Preferred Offsite M.A. M.A. N.A. > 5004 V 1 5900 V Source Undervoltage - > 4900 V i .. "w e w . .A- w-- ., , . ,. m ,,w i.,,y . . . . % .-, . , , _ . - , _ , , .

g- - - --

^- ~

i P pe /f of A4 l E'MERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS l COLD LEG INJECTION m ,

j LIMITING CONDITION FOR OPERATION I 3.5.1 Each cold leg injection accumulator shall be OPERABLE with:

a. The discharge isolation valve open with power removed, f

b. An indicated borated water level of between 395 and 61% l

c. A beron concentration of between 1900 and 2200 ppm, and f

d. An indicated cover-pressure of between 623 and 644 psig. '

APPLICABILITY: MODES 1,'2, and 3*.

ACTION: I

a. With one cold leg injection accumulator inoperable, except as a result .:

of a closed. isolation valve or the boron concentration outside the ;

required values, restore the inoperable accumulator to OPERABLE status '

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l and reduce pressurizer pressure to less than 1000 psig within the !

following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.: With one cold leg injection accumulator inoperable due to the

isolation valve being closed, either immediately open the isolation valve or be in at least HOT STAND 8Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . ;

c. With the boron concentration of one cold leg injection accumulator outside the required limit, restore the boron concentration to within the required limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY l

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1 Each cold leg injection accumulator shall be demonstrated I l OPERABLE: .

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by: y

1) Veritying the indicated borated water nd nitrogen cover pressure in the tanks, and

Pressurizer pressure above 1000 psig.

COMANCHE PEAK - UNIT 1 3/45-1

-~" As54 es A A 7'u- t o o s / . __

C.,

b)t /A o.f .14

. EMERGENCY CORE COOLING SYSTEMS ,

3/4.5.3 ECCS SUBSYSTEMS - T,yg < 350*F l 1

^

ECCS SUBSYSTEMS LIMITING CONDITION FOR OPERATION j l-3.5.3.1 As a minimum, one ECCS subsystem comprised of the following shall be !

OPERA 8LE:

a. One OPERA 8LE centrifugal charging pump,* '

1

b. One OPERA 8LE RHR heat exchanger,

c. One OPERA 8LE RHR pump, and

.' d. An OPERA 8LE flow path capable of taking suction from the refueling f water storage tank upon being manually realigned and transferring suction to the containment sump during the recirculation phase of ;

operation.

APPLICABILITY: MODE 4.

t .

ACTION: '

a. With no ECCS subsystem OPERA 8LE because of the inoperability of either the centrifugal charging pump or the flow path from the ;

refueling water storage tank, restore at least one ECCS subsystem to ;

OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the aevt .

20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months . !

b. With no ECCS subsystem OPERA 8LE because of the inoperability of i either the residual heat removal heat exchanger or RHR pump, restore ,

at least one ECCS subsystem to OPERA 8LE status or maintain the Reac-L tor Coolant System T8 less than 350*F by use of alternate heat removal methods. l

c. .In the event the ECC'S is actuated and injects water into the Reactor '

Coolant System, a Special Report shall be prepared and submitted to !

the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated ;

actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special Report whenever its value exceeds 0.70.

E

A maximum of two charging pumps shall be OPERA 8LE whenever the temperature of ,

one or more of the RCS cold legs is less than or equal to 350*F, except 4t%/ hew >

p!hu; ty- Specification 3.4.8.3[l.r aet yphc 4/v, COMANCHE PEAK - UNIT 1 3/4 5-7

, , . A,..d u A w TAX-Oce / -

l 77 . , ; ' ha i.3 o 4 d '( l EMERGENCY CORf COOLING SYSTEMS j SURVEIL LANCE REQUIREMENTS j r

a 4.5.3.1.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable l requirements of Specification 4.5.2. '

'.g not a #Ue*4/d . . de,J !

4.5.3.1.2 A maximum of two herging pumps shall be OPERABLE except 4e ;:_:to- :

,, by- Specification 3.4.8. hen required, one charging pump shall be demon- '

strated inoperable

by verifying that the motor circuit breaker is secured in 1 the open position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 or prior to ;

the. temperature of one or more of the RCS cold legs decreasing below 325'F, ~{

whichever occurs first and at least once per 31 days thereafter. t t

k j

i i

l t

i "An inoperable pump may be energized for testing provided the discharge of the e

pump has been isolated from the RCS by a closed isolation valve (s) with power removed from the valve operator (s) or by a manual isolation valve (s) secured in the closed position.

l L

l ,

COMANCHE PEAK - UNIT 1 3/4 5-8

- - - . - _ _ - _ - _ _ - - _ _ _ _ _ . - - ~ . - _ _ _ - ,_ - - - - . , - - , , , - - - - - 9

3,

' & 4 a k u TXk '956! L' 4 .

A s y c4 c '!

>Ed3RGENCYCOR) COOLING $YSTEMS I' !

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,. 3/4.5.3' ECCS SUS $YSTEMS - T,yg < 350'F SAFETY INJECTION pumps l u i

,, s LINITING CON 0! TION FOR OPERATION

'i 3.5.3.2 All safety injection pumps shall be inoperable. !

APPLICABILITY: Modes #4 , 5, and 6 with the reactor vessel head on. I

. ACTION: i With a safety injection pump OPERABLE, restore all safety injection pumps to -

an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . i SURVEILLANCE REQUIREMENTS )

1 4.5.3.2 All safety injection pumps shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from M00E 3 or prior to the temperature of one or more of the RCS cold legs decreasing below 325*F, whichever occurs first~

and at least once per 31 days thereafter. i l

e

+,

l ?

'I f

An inoperable pump may be energized for testing or for filling accumulators l provided the discharge at the pump has been isolated from the RCS by a closed l- isolation valve (s) with power removed from the valve operator (s), or by a manual

isolation valve (s) secured in the closed position. -

dea l kle .

Except :: :llnct 5,# Specification 3.4.8. Q ;,5 aof off ic< l COMANCHE PEAK - UNIT 1 3/4 5-9 .

Y a

_, _,.,,.,c., , _ . . - -

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,'. A30 /S of AY

' EMERGENCY CORE COOLING SYSTEMS I

3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION t.

3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

3 4. A minimum indicated borated water level of 95%,

b. A boron concentration of between 2000 and 2200 ppm of boron, c, A minimum solution. temperature of 40'F, and

d. A maximum solution temperature of 120*F.

I.

APPLICABILITY: MODES 1, 2, 3, and 4.

A_C, TION:

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l-(.

1-SURVEILLANCE REQUIREMENTS 4.5.4 The RWST shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) le velV Verifying the indicated borated water h in the tank, and l

i l 2) Verifying the boron concentration of the water,

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when l l .

the outside air temperature is less than 40'F or greater than 120'F.

i l

COMANCHE PEAK - UNIT 1 3/4 5-10 .

l

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F 1 :$ A A . + a.

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[ ,;,'._ [dNTAINMENTSYSTEMS' i

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, . 3/4.6.3' CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION I

3.6.3 The containment isolation valves shall be OPERA 8LE.# !

> w APPLICABILITY: MODES 1, 2, 3, and 4 ACTION: i i

With one or more of the containment isolation valve (s) inoperable, maintain at l least one isolation valve OPERA 8LE in each affected penetration that is open and: ,

a. Restore the inoperable valve (s) to OPERA 8LE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , f or .

I

b. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least .;

one deactivated automatic valve secured in the isolation position, or

c. Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least '

one closed manual valve or blind flange, or

d. 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 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

r SURVEILLANCE REQUIREMENTS 4.6.3.1 The containment isolation valves shall be demonstrated OPERABLE prior to returning the valve to service after maintenance, repair or replacement work :

L is performed on the valve or its associated actuator, control or power circuit

) by performance of a cycling test, and verification of isolation time.

3 1

1 1

The requirements of Specification 3.6.3 do not apply for those valves covered

, , by Specifications 3. 7.1.1, 3. 7.1. 5, g3. 7.1. 66 mA 7. 7. /. 7. -

CAUTION: The inoperable isolation valve (s) may be part of a system (s).

Isolating the affected penetration (s) may affect the use of the system (s).

Consider the technical specification requirements on the affected system (s) ;

and act accordingly.

l' COMANCHE PEAK - UNIT 1 3/4 6-13 ,

.4 e

'' hMimmed O to TKK-9 0% f F > fap siof 29-( 4 ,'

P(, ANT SYSTEMS L

L .

CONDENSATE STORAGE TANK w

g LIMITING CONDITION FOR OPERATION _

t -

t 3.7.1.3 The condensate storage tank (CST) shall be OPERA 8LE with an. indicated wat e level of at least 534.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With the CST inoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

a. Restore the CST to OPERABLE status or be in it least HOT STANOBY 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 , or I b. Demonstrate the OPERA 81LITY of the Station Service Water (SSW)

system as a backup supply to the auxiliary feedwater pumps and restore the CST to OPERA 8LE status within 7 days or be in at least HOT STAND 8Y 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. hours.

SURVEILLANCE REQUIREMENTS leseI

$. 7.1. 3.1 The CST shall be demon rated OPERA 8LE at least once per 12 hoiire hv

t. verifying the indicated water is within its limits when the tank is the j supply source for the auxiliary feedwater pumps. !

~

4.7.1.3.2 The SSW system shall be demonstrated OPERABLE at least once per 12 ,

, bours' whenever the SSW system is being used as an alternate supply: source to '

the auxiliary feedwater pumps by verifying the SSW system OPERABLE and each motor operated valve between the SSW system and each OPERA 8LE auxiliary feed-

. water pump is OPERABLE.

.( .

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I-COMANCHE PEAK - UNIT 1 3/4 7-5

i Attachment 2 to TXX 90061

,' f Page 18 of 24 TABLE 3.7-3 AREA TEMPERATURE MONITORING l

MAXINUM

.._ AREA.- t - . ,,- .

TEMPERATURE LIMIT ('F)

Normal Abnormal -

Conditions Conditions !

1. Electrical and Control Building i

Normal Areas 104 131 c Control Room Main Level (E1. 830'-0") 80 104 i Control Room Technical Support Area ~

(El. 840'-6") 104 104 UPS/ Battery Rooms 104 '

113 Chiller Equipment Areas 122 131 i

2. Fuel Building s Normal Areas 104 131 .

Spent Fuel Pool Cooling Pump Rooms 122 131

3. Safeguards Building Normal Areas 104 131 AFW, RHR, SI, Containment Spray Pump Rooms 122 131 RHR Valve and Valve Isolation Tank Rooms 122 131 RHR/CT Heat Exchanger Rooms 122 131 Diesel Generator Area 122 131 Diesel Generator Equipment Rooms 130 131 >

, Day Tank Room 122 131 1..

4. Auxiliary Building '

Normal Areas 104 131

l. CCW, CCP Pump Rooms 122 131 l CCW Heat Exchanger Area 122 '131 -

l: CVCS Valve and Valve Operating Rooms 122 131 j -

Auxiliary Steam Drain Tank Equipment Room 122 131 l Waste Gas Tank Valve Operating Room 122 131 ,

5. Service Water Intake Structure 127 131 !

1 '

i

6. Containment Building 1

l General Areas 120 129

-C""" " M n r. ISO

^

Reactor Ca ity Exhaust 150 hyiog'A, E s s ;u m e H ~ 16 in l

I COMANCHE PEAK - UNIT 1 3/4 7-24

f , ,4 % h wy 4 Trt.9406i

' ,; 4 9o O d A 'f PLANT $11"fM$

.. .in ,

SA!($- -

SNUB 8tR$ (Continued) - 'T ' " I. "

A list of individual snubbers with detailed information of snubber loca-tion and site and of system affected shall be available at the plant in accora dance with 10 CFR 50.71(c). The accessibility of each snubber shall be determined and approved by the Station Operation Review Committee (SCRC). Th6 determination shall be baseu upon the existing r:;diation levels and the expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temperature, atmosphere, location, etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addiP on or deletion of any hydraulic or mechanical snubber I

1 shall be made in ace ndance ..ith 10 CFR 50.59.

Surveillance to demonstrate OPERABILITY is by performance of the recuirements of an approved inservice inspection program.

Permanent or other exemptions from the surveillance program for incivicual snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performec to qualify the snubbers for the applicable design conditions at ettner the com-pletion of their fabrication or at a subsequent date. Snubbers so exempted shall be listed in the list of individual snubbers indicating the extent of the uemptions.

The service life of a snubber is established via manufacturer input anc information through consiceration of the snubber service conditions ano ar,sociated installation and maintenance records (newly installed snuboers, seal replaceo, spring replaced. in high radiation area, in nign temperature area, etc.). The requirement to monitor the snubber service life is inclucec to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life.

3/4.7.10 AREA TEMPERATURE MONITORING The limitations on nominal area temperatures ensure that safety relatec equipment will not be subjected to temperatures that would impact their environ-mental qualification temperatures. Exposure to temperatures in excess of the maximum temperature for normal conditions for extended periods of time coulo reduce the qualified life or design life of that equipment. Exposure to tor-peratures in excess of the maximum abnormal temperature could degrace the operability of that equipment. ,g See Insut A -

3/4.7.11 UPS HVAC SYSTEM The OFERABILITY of the UPS HVAC System ensures that the uninteruptible power supply and distribution rooms ambient air temperatures do not exceed the i allowable temperatures per Specification 3/4.7.10 for continuous-cuty rating for the equipment and instrumentation , cooled by this equipment.

CCMANCHE PEAK UNIT 1 B 3/4 7-6 i l

p-Attachment 2 to TKK 90061 Pepe 20 of 24 -

r , , i TN5titT- A '

j l

Nov **l and, A4wey n4l *mpe vaiurs lem'efs f,r the foltaWins Aread are a5Wred bJ mniton'n,3 i oky Qvea s with a. Co r re lated.

h7"P'r*f u r Yel*48ansicri l "I AWWI Ay'4-Conditions Gmd if tees Am Manieve(

I cRm Plaiform 14 o \41 6ttoemi A rea.

&"I*f l CRbM Shre d i hkaug l t

3eacter Cav i+J 13 5 17r j kcaefer caveh "Delecter Welt h%k j R c . P ipa. 2eo 2or 6t ene,al Areas !

Penefra%n ExAsur '

heacter Cavdy l (N.ig 6.+e $+ rs) GKhaust !

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I

b*b L. 4 A. 4 7xr M3/

A atof M i q .

3fl. 8 EL(CYR! CAL POWER SYSTEZS l O .

+

BASES l

3/4.8.1, 3/4.8.2, and 3/A.B.3 A.L SOURCES. D.C. SOURCES. and ONSITE POWER :

DI5TRIEUTION l The OPERABILITY of the A.C. and 0.C power sources and associateo distribu-tion systems during operation ensures that sufficient power wi11 be available i to supply the safety relatec equipment required for: (1) the safe shutdown of i the facility, and (2) the mitigation and control of accicent conditions within !

the facility. The minimum specified independent and redundant A.C. and D.C. ,

power sources and distribution systems satisfy the requirements of General ;

Design Criterion 17 of 10 CFR $0 Appendix A.

The ACTION requirements specified for the levels of degradation of the :

power sources provide restriction upon continued facility operation commensurate r with the level of degracation. The OPERABILITY of the power sources are ;

consistent with the initial condition assumptions of the safety analyses ano .

&re cased upon maintaining at least one redundant set of onsite A.C. and D.C. I power sources and associated distribution systems OPERABLE during accident ;

conditions coincident with an assumed loss of offsite power and single f ailure :

of the other onsite A.C. source. The A.C. and 0.C. source allowable out-of- '

service times are based on Regulatory Guide 1.93, " Availability of Electrical Power Sources," December 1974 and Generic Letter 84-15, "Prnposed Staf f Position '

to Improve and Maintain Diesel Generator Reliability." When one diesel generator i is inoperable, there is an additional ACT!0N requirement to verify that all ;

required systems, subsystems, trains, components and devices, that depend on the remaining OPERABLE diece) generator as a source of emergency power, are also OPERABLE, and that the steam-driven auxiliary feedwater pump is OPERABLE.

inis requirement is intenoeo to provide assurance snat a loss of ottsite power i

event will not result in a complete loss of safety function of critical systems '

during the period one of the diesel generators is inoperable. The term, .

verify, as used in this context mans to administratively check by examining !

logs or other information to determine if certain components are out of service !

for maintenance or other reasons. It does not mean to perform the. Surveillance i Requirements needed to demonstrate the OPERABILITY of the component. !

The OPERABILITY of the minimum specified A.C. and D.C. power sources and i associated distribution systems during shutdown and refueling ensures that: i (1) the facility can be maintained in the shutdown or refueling condition for -

axtended time periods, and (2) sufficient instrumentation and control capa- ;

/ZMEAT al-bility is available for monitoring and maintaining the unit status. '

'The Surveillance Requirements for demonstrating the OPERABILITY.of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power ;

Supplies," March 10, 1971; 1.108, " Periodic Testing of Diesel Generator Units t Used as Onsite Electric Pever Systems at Nuclear Power Plants," Revision 1, [

August 1977: and 1.137, " Fuel-011 Systems for Standby Diesel Generators," ;

January 1978, Generic Letter 84-15, and Generic Letter 83 26. " Clarification of Surveillance Requirements f* Diesel Fuel Impurity Level Tests."

COMANCHE PEAK - UNIT 1 B 3/4 8-1 . ,

o ,

' t. .i Attachment 2 to TXX+p0061 -

H ;

tape 22 of 24 '

s ItiftT 8 [

\

The OPERASILITf of the day fuel tank and Fuel 5 terete $ystee are based on the !

fellesing: 1) the einleum day fuel tank values enevres sufficient fuel :

issediately eveilable to operett the diesel ponerater et the contineous reting i for 60 minutes plus 30 Percent, and 2) the remaining day fuel teak volume .

(between that required for (1) ebeve end.the valuee specified in the tietting ,l Conditions for Operation), combined with the minious specified Fuel Storage !

$ysten volume, ensures sufficient on. site fuel et) storage cepeetty to operete j the di.sei gen.reter et the ceasinuous rating for seven days, l

The Fue) $terage $pstes consists of the fuel oil storage tank and is !

equivalent to the AMS! W195 1976 definition for supply tank. i e

l.

i i

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i t

i l

I I

l I

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l 9

I

Y. . M u S = e a. u rn.s:sst Y s h( col. l *,ADN!N!$TRATIVE

%e M o fCONTE A 4 0L$

". UNIT STAFF QUALIFICATION $ (Continued)

V h'c the oualifications of ANSI N18.1-1971, technicians and maintenance personnel f .- .'L 'may be permitted to perfom work in the specific task (s) for which qualification 4.

has been demonstrated.)

6.4 ' TRAINING 6.4.1 A retraining and replacement training program for the unit staff shall y' ' be maintained under the direction of the Vice President, Nuclear Operations and shall meet or exceed the requirements and recommendations of AN$1 N18.1-1971, 10 CFR 55, and shall include familiarization with relevant indust y opera-tional experience.

4.5 REV!EW AND AUDIT 6.5.1 $TATION OPERATIONS REVIEW COMMITTEE (50RC)

FUNCTION

6. 5.1.1 The 50RC shall function to advise the Vice President, Nuclear Operations on all matters related to nuclear safety.

COMPOS 1TTON p n. minimump 6.5.1.2 The S0RC shall be composed of managers or individuals reporting ~

I directly to manecers from the areas listed below and meet the requirements of ANSI N18.1-1971 $ectic'ns 4.2 or 4.4 for required experience.

Operations Maintenance

,. Instrumentation and Controls Technical Support l Radiation Protection Sespity c>

S:t h; -

l The Plant Manager shall serve as the chairman of 50RC. A senior health physi-cist is acceptable for the Radiation Protection representative on 50RC. The 50RC members shall be designated, in writing, by the Vice President, Nuclear i Operations.

ALTERNATES

.6. 5.1. 3 All alternate members shall be appointed in writing by the Vica President, Nuclear Operations to serve on a temporary basis; however, no more l

than two alternates shall participate as voting members in 50RC activities at i any one time. .

COMANCHE PEAK - UNIT 1 -

f-5

h aw ,,,r A 4 TrYate6s I

'e*

,e & D4 *+ M I

i. , ADMIN!$TRATIVE CONTROLS j l

l -

' MEETING FREQUENCY 6.5.1.4 The 50RC shall meet at least once per calendar month and as convened I by the 50RC Chairman or his designated alternate, j i

QUORUM

(

i 6.5.1.5 The quorum of the $0RC necessary for the performance of the 50RC -

responsibility and authority provisions of these Technical Specifications !

I shall usM4consist

-al'-*w.of the Chairman or his designated alternate and 6 -(t) IntM*fybt-9 o- majoriy o reg a lsr mem be rs (o t~ +keir a l+ crode.s ),

RESPONSIBILITIES

6. 5.1. 6 The 50RC shall be responsible for: i

a. Raview of applicable administrative procedures recommended in

}

Appendix A of Regulatory Guide 1.33 Revision 2, February,1978. -

b. Review of the safety evaluations for: (1) procedures, (2) change to procedures, equipment, systems or facilities, and (3) tests or t experiments completed under the provision of 10 CFR 50.59 to verify ;

that such actions did not constitute an unreviewed safety question, s

c. Review of proposed procedures and changes to procedures, equipment, i systems or facilities which involve an unreviewed safety question as defined in 10 CFR 50.59 or involves a change in Technical I;;;if h: tion:; !

d. Review of proposed test or experiments which involve an unreviewed

! safety question as defined in 10 CFR 50.59 or requires a change in j Technical Specifications,

e. Review of proposed changes to Technical Specifications or the :

Operating License; j

~

f. Investigation of all violations of the Technical Specifications including the forwarding of reports covering evaluation and recom- !

mendations to prevent recurrence to the Vice President Nuclear j Operations and to the CRC i

g. Review of reports of operating abnormalities, deviations from ex-l pected performance of plant equipment and of unanticipated defici- l encies in the design or operation of structures, systems or 4 components that affect nuclear safety; !

h. Review of all REPORTABLE EVENTS; I i

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COMANCHE PEAK - UNIT 1 - 6-6 ,

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ML20012F041

Text

SUBJECT:

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