Text

Forwards marked-up Proof & Review Tech Specs,Per 850904 Request,Indicating Typos & Corrections to Info Submitted in 840928 & s
	ML20137V848
Person / Time
Site:	Clinton
Issue date:	10/02/1985
From:	Spangenberg F; ILLINOIS POWER CO.
To:	Butler W; Office of Nuclear Reactor Regulation
References
	U-600275, NUDOCS 8510040140
	Download: ML20137V848 (85)
	v • d • e

__ _

r y . e 8 U-600275 058-85 (10- 02)-L 1A.120 ILLIN0/8 POWER COMPANY IP CLINTON POWER STATION, P.o. BOX 678. CLINTON. ILLINol$ 61727 October 2, 1985 Docket No. 50-461 Director of Nuclear Reactor Regulation Attn: Mr. W. R. Butler, Chief Licensing Branch No. 2 Division of Licensing U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

Clinton Power Station Unit #1 Technical Specifications

Dear Mr. Butler:

Consistent with your request in your letter dated September 4,

,e 1985, to Mr. F. A. Spangenberg, Illinois Power Company is providing h information relating to inconsistencies identified as a result of our review of the " Proof and Review" copy of the Clinton Power Station Technical Specifications (CPS-TS).

The enclosure consists of marked up pages from the " Proof and Review" copy of CPS-TS. These pages identify such items as typographical errors and corrections to information previously submitted in letters dated September 28, 1984 (U-0739) and May 23, 1985 (U-600079). These corrections have been discussed with your Mr. C. S.

Schulten of the Technical Specification Review Group.

On-going programs such as the Technical Specification Commitment Tracking System, Preoperational testing, surveillance procedure writing and the-Final Safety Analysis Report (FSAR) certification will provide final confirmation ~that the information in the CPS-TS is consistent with

-the Safety Evaluation Report (SER), FSAR, and the As-Built design. We are confident that final certification can take place in accordance with the schedule provided to Illinois Power Company in a letter from A. Schwencer to F. A. Spangenberg dated January 31, 1985. We will keep Mr. C. S. Schulten inforced of any changes to the " Proof and Review" copy of CPS-TS as a result of our continued review. We will work with Mr. Schulten as requested in order to resolve concerns of the Technical Branches and the Region as they v cur.

/'

B510040140 851002 PDR ADOCM 05000461 A PDR

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L U-600275 058-85( 10 0@-L 1A.120 Please contact us if you~have any questions regarding this information.

Sincerely yours, F. A. Spar kenbe Manager - Licensing and Safety FAS:RFP/kaf Attachments cc: B. L. Siegel, Clinton Licensing Project Manager, w/o enclosure NRC Resident Office, w/o enclosure Regional Administ'rator, Region III, USNRC, w/o enclosure Illinois Department of Nuclear Safety, w/o enclosure C. S. Schulten, NRC Technical Specification Review Group, w/ enclosure it- USNRC

% Mail Stop 509 Washington, DC 20555

4 Attachment O

\

i 9 100F & REYiEW COPY '

1 0FFINIT!ONS SECONDARY CCNTAINMENT INTEGRITY 1.38 SECONDARY CONTAINMENT INTEGRITY shall exist when: g 1

a. All secondary containment penetrations required to be closed .'during i accident conditions are either: '

1. Capable of being closed by an OPERABLE secondary containment automatic isolation system or

2. Closec by at least one manual valve, blind flange, or deactivated automatic valve or damper as applicable secured in its closed posi-tion, except is provided in Table 3.6.6.2-1 of Specification 3.6.6.2.

b. All secondary containment equipment hatches are closed and sealed.

c. The standby gas treatment system is in compliance with the requirements of Specification 3.6.6.3.

d. %At least one door in each access to the secondary containment is closed, except for normal entry and exit..

o. The sealing mechanism associated with eacli secondary containment penetra-tion, e.g. , welds, bellows or 0 rings, js OPERABLE. ,

f. The pressure within the secondary containment is less than or equal to the value required by Specification 4.6.6.1.a.

SELF TEST SYSTEM 1.39 The SELF TEST SYSTEM shall be that automatic system that injects short- .

duration pulses into the solid state nuclear system protection system (NSPS) circuits and The SELF TESTSYSTEM%verifiesproperresponsetovariousinputcombinations." Wesigned to mainta cuttry essential to the Reactor Protection System, Emergency Core Cooling Sys-tens, and the Nuclear Steam Supply Shutoff System on a continuous cyclic basis.

Tha SE.'.F TEST SYSTEM may be used to perform various surveillance testing functiuns to satisfy technical specifications requirements for those components it is designed to monitor. The STS may be used to augment conventional testing methods to perform CHANNEL CHECKS, CHANNEL FUNCTIONAL TESTS, CHANNEL CALIBRA-4 TIONS, RESPONSE TIME TESTS AND LOGIC SYSTEM FUNCTIONAL TESTS.

SHUTDOWN MARGIN 1.40 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be suberitical assuming all control rods are fully inserted, except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn, and the reactor is in the shutdown condition; cold, i.e. ,

68*F; and xenon-free.

SITE BOUNDARY

! 1.'41 The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensee.

CLINTON - UNIT 1 1-7 a..,....

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. PR00F & REVIEW COPY i

SAFETY LIMITS

BASES

=

!, 2.1.3 REACTOR COOLANT SYSTEM PRESSURE $

The Safety Limit for the reactor coolant system pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the -

system is not endangered. . The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code 1971 Edition, including Addenda through Summer 1973, which permits a maximum pressure transient of 110%, 1375

'W. psig, of design pressure 1250 psig. The Safety. Limit of 1325 psig, as measured

' by the reactor vessel steam dome pressure indicator, is valer.t to 1375 psig at the lowest elevation of the reactor coolant system. e reactor coolant

. system is designed to the ASHE Boiler and Pressure Vessel Code, 1974 Edition, including Addenda through the Summer of 1974, for the reactor recirculation piping which permits a maximum pressure transient of 120% equaling to 1500

' (suction) psig and 1980 (discharge) psig of design pressures of 1250 psig for suction piping and 1650 psig for discharge piping from the recirculation pump i discharge co the outlet side of the discharge u off valve and 1550 psig from

,, the discharge shutoff valve to the jet pumps. he pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes.

4 2.1.4 REACTOR VESSEL WATER LEVEL

) '

With fuel in the reactor vessel during periods when the reactor is shut down, '

consideration must be given to water level requirements due to the effect of l decay heat. If the water level should drop below the top of the active irra-p' .

' diated fuel during this period, the ability to remove decay _ heat is reduced.

This reduction in cooling capability could lead to elevated cladding tempera-tures rnd clad perforation in the event that the water level became less than two-tFirds of the core height. The Safety Limit his been established at the top of tne active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action.

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CLINTON - UNIT 1 B 2-5 Aud 231985 I

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c i MC:&hTEWCDPY l REACTIVITY CONTROL SYSTEMS -

SURVEILLANCE REQUIREMENTS (Continued) _

4.1.3.1.4 The scram discharge volume shall be determined OPERABLE'by demonstrating: .

h

. g-

a. The scram discharge volume drain and vent valves are OPERABLE, when con- 4 trol rods are scram tested from a normal control rod configuratfor. of less o than or equal to 50% ROD DENSITY at least once per 18 months, by veri- of fying that the drain and vent valves: 2 N

1. Close within 30 seconds after receipt of a signal for control rods @

to scram and , o

2. Open when the scram signal is reset. h

-o

b. ' Proper level sensor response by performance of a CHANNEL FUNCTIONAL TEST 2 for each scram discharge volume scram and control rod block level instru- 'I mentation at least once per 31 days.

4.1.3.1.4 The scram discharge volume shall be determined OPERABLE by

..- - demonstrating the scram discharge volume drain and vent valves OPERABLE,

) %

at least once per 18 months, by verifying that thit drain and vent valves:

a. Cicse within 30 seconds after receipt of a signal for c'entrol rods to scram, and g

b. Open when the scram signal is reset. =

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k CLINTON - UNIT 1 3/4 1-5 AUG 2 s 1985 -

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l PR00F & REVIN WI 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION _

3.3.1 As a minimum, the reactor protection system instrumentation channels shown.in Table 3.3.1-1 shall be OPERA 8LE with the REACTOR PROTECTION SYSTEM

, RESPONSE TIME as shown in Table 3.3.1-2. -

APPLICABILITY: As shown in Table 3.3.1-1. C/

Y.

M ACTION:

As shown in Table 3.3.1-1. v"V Ajr -

pt SURVEILLANCE REQUIREMENTS '

(46 /1I.d 4 m i

1

..oe 3L 4.3.1.1 Each reactor protection system instrumentation channel shall be demon-strated OPERABLE by, the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL

-TEST and CHANNEL CALIBRATION operations for the OPERATIONAL CONDITIONS and at the frequencies shown in Table 4.3.1.1-1.

4.3.1.2 LOGIC. SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.

4.3.1.3 The REACTOR-PROTECTION SYSTEM RESPONSE TIME of each reactor trip functional unit shown in Table 3.3.1-2 shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least two logic trains such that all logic trains are tested at least once per 36 months and one channel per trip function such that all channels are tested at least once every N times 18 months vhare N is the total number of redundant channels in a specific reactor trip function. .

CLINTON - UNIT 1 3/4 3-1 AUG 2 e 545

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TABLE 4.3.1.1-1 (Continued) *

p . REACTOR PROTECTION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS .!

i !i! i CHANNEL OPERATIONAL

! d CHANNEL. FUNCTIONAL CHANNEL CONDITIONS IN WHICH l 7 FUNCTIONAL UNIT CHECK TE'ST CALIBRATION (a) .

SURVEILLANCE REQUIRED l,

} c 9. Scram Discharge Volume Water i

! i5 Level - High j f

a. Level Transmitter S M -

R I9) 1, 2, 5 I-

b. Float Switches NA Q R 1,2,5 l

l 10. Turbine Stop Valve - Closure NA' M R 1

11. Turbine Control Valve Fast !'

i Closure Valve Trip System 011 l Pressure - Low NA H R 1

] 12. Reactor Mode Switch -

l ,

Shutdown Position NA R NA 1,2,3,4,5 (

13. Manual Scram

NA M NA 1,2,3,4,5

w (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. -

4 (b) The IRM and SRM channels shall be determined to overlap for at least I decade during each startup I

after entering OPERATIONAL CONDITION 2 and the IRM and APRM channals shall be determined to overlap

l. for at least 1 decade during each controlled shutdown, if not performed within the previous 7 days.

(c) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to startup, if not performed within the previous 7 days.

l g, j,

(d) this calibration shall consist of the adjustment of the APRM channel to conform to the power values ::>. s ,

calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERNAL C2 !

l!

POWER. Adjust the APRM' channel if the absolute difference is greater than 2% oT RATED THERMAL POWER.

Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in Cd$

}, t determining the absolute difference.

(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a p l~il'

l. calibrated flow signal.

(f) The LPRMs shall be calibrated at leasLonce per 1000 effective full power hours (EFPH) using the TIP system.

@:=3 F 4 j* *

(g) Calib ate the analog trip module unT Cat least once per 31 days. *

'N %)

4 (h) Verify measured core (total core flow) flow to be greater than or equal to established core flow.at the c :n existing loop flow control (APRM % flow). C3 i

{,mR (i) This calibration shall consist of verifying the 610.6 second simulated thermal power time constant.

(j) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per M!jil

!2 a Specification 3.10.1. .

j t (k) With any control rod withdrawn. Not appilcable to control rods removed per " ecification 3.9.10.1 or 3.9.10.2. r

!I'g! (1) This function is not required to ha OPERABLE when DRYWELL INTEGRITY is not required to be OPERA 8LE per Special Test Exception 3.10.1. {'

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f, , TABLE 3.3.2-1 ,

ISOLATION ACTUATION INSTRUMENTATION n l

.C HINIMOM OPERABLE APPLICABLE l

$ ISOLATION CHANNEg)PERTRIP OPERATIONAL i E TRIP FUNCTION SIGNf.L tt SYSTEM CONDITION ACTION

1. PRIMARY CONTAllWENT ISOLATION I 2i. a. Reactor Vessel Water Level- ' '-

Low Low, Level 2 B(b)(c)

C 2 li 2 3 and # 20

b. Drywell Pressure - HiCh L 2 1, 2, 3 20

c. Containment Building fuel Z 2 1, 2, 3 and

21 r.

Transfer Pool Ventilation -

Plenum Radiation - High

d. Containment Building M 2 .1, 2, 3 and

21 Exhat:st Radiation - High

' . t

e. Cor.Lainment Building 5 ,

2 1, 2, 3 and

21 '

Continuous Containment d.

t'

Purce (CCP) Exhaust ,

tab Radiation - High' '

i A H

f. Manual Initiation.

NA 1/shstem 1, 2, 3 and *# 26 ISOLATION HINIMUM APPLICARLE l

SIGNAL TOTAL NO. CHANNELS OPERABLE OPERATIONAL j

. I 2. MAIN STEAM LINE ISOLATIONt y VidVE '"*". OF CHANNELS TO TRIP CHANNELS CONDITIONS ACTION

~

a. Reactor vessel Water Level- !

Low Low Low,. Level 1 0 . 4 2 3 1,2,3 20 '

b. Main Steam 1.ine ~~~!

Radiation - High Id) C 4 2 3 1,2,3 23 i I i

c. Main Steam Line. i;f

. Pressure - Low H 4 2 3 1,2,3 23 E,. ,

d. Main Steam Line

!; ' Flow - High 0 4/M5t.-

2g /NW 1'

' -l85 23 5,* .

j .e. Condenser Vacuus - Low J 4 2 3 1, 2,** 3** 23 :n ,

Main Steam Line Tunnel E

f.

Temp. - High E 4 2g 3 1,2,3 23 f

i: .

g. Main Steam Line Tunnel j

A Temp. - High F 4 2g 3 1,2,3 23 e,

h. Main Steam Line Turbine i c.-

Bldg. Temp. - High G 4 1,2,3

. m 1. Manual Initiation NA 2/ system 2

2/ system 3

2/ system 1, 2, 3 23 22

[-

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'THIS PAGE OPEN PENDING REC $fT "d7 PRQQf & REhf# COPI -

IMORMATION FROM THE APPLICANT 1 TABLE 3.3'.2-1 (Continued)

ISOLATION ACTUATION INSTRUMENTATION ACTION

_.?

i- ACTION 20 -

Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN

]

* ~

within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

j ACTION 21 -

Close the affected system isolation valve (s) within.1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or:

a.t In OPERATIONAL CONDITION 1, 2 or 3, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

E b. In Operational Condition *,' suspend CORE ALTERATIONS, handling of irradiated fuel in the containment and operations with a potential for draining the reactor vessel.

ACTION 22 -

Restore the manual initiation function to' OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 23 -

Be in at least STARTUP with the associated isolation valves closed within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 24 -

Be in at least STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 25 -

Establish SECONDARY CONTAINMENT INTEGRITY with the standby gas

. treatment system operating with'in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 26 -

Restore the manual initiation function to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN.within-the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

' ACTION 27 Close the hffected syttem isolation valves within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable.

ACTION 28 -

Lock the sffected syste:'t isolation valves closed within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and declare the affected system inoperable... -

NOTES When handling irradiated fuel in the primary or secondary containment and during CORE ALTERATIONS and operations with a potential for draining the reactor vessel. meM88H(h Ocf in M y May be bypassed with reactor N r F ;;;r; G U 4 and all turb.ine l stop valves closed.

- - When handling irradiated fuel in the secondary containment.

During CORE ALTERATIONS and operations with a potential for draining the reactor vessel.

t Main Steam line isolation trip functions have 2 out-of 4 isolation logic.

tt See Specification 3.6.4 Table 3.6.4-1 for valves which are actuated by these isolation signals.

(a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for required surveillance without placing the trip system in the tripped con-

. dition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.

(b) Also actuates the standby gas treatment system.

(c) Also actuates the control room emergency filtration system in the isolation -

mode of operation. -

(d) Also trips and isolates the mechanical vacuum pumps.

(e) This note deleted. -

(f) Also actuates secondary containment ventilation isolation dampers and _-

valves per Table'3.6.6.2-1. -

CLINTON - UNIT 1 .

3/4 3-15 AUG 2 91885 t . . _ . _ _ _ . _ . . . . .

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. TABLE 3.3.2-2

' ISOLATION ACTUATION INSTRUMENTATION SETPOINTS

~1 n

! C TRIP FUNCTION -

TRIP SETPOINT ALLOWABLE VALUE 1.

h PRIMARY CONTAINMENT ISOLATION

. a. Reactor Vessel Water Level - -

Low Low, Level 2 -

> -45.5 in.* 1 -47.7 in.

[ b. Drywell Pressure - High 1 1.68 psig 1 1.88 psig

c. Containment Bldg. Fuel Pool Transfer . !

Ventilation Plenum Radiation - High 1 100 mR/hr 1 500 mR/hr

d. Containment Bldg. Exhaust ;

Radiation - High 5 100 mR/hr 1 500 mR/hr i

e. Contains.ent Bldg. Continuous containment Purge (CCP) .

Exhaust Radiation - High 1 100 mR/hr $ 400 mR/hr ;

2" f.-

Manual Initiation NA i NA '

g 2. MAIN STEAM LINE ISOLATION

a. Reactor Vessel Water Level - '

Low Low Low, Level 1 2 -145.5 in." 1 -147.7 in.

I b. Main Steam Line Radiation - liigh 1 3.0 x full power background 1 3.6 x full power background ,

90 l

c. Main Steam Line Pressure - Low 1 849 psig 1 837 psig y

' ' j

d. Main Steam Line Flow - liigh 5 170 psid*

$ 178 psid Q, .

. e. Condenser Vacuum - Low t 8.5 in. lig vacuum 17.6 in lig vacuum Qo ,

f. Mdin Steam Line Tunnel "

'S

?. !

Temp. - High I

l .

g

g. Main Steam Line Tunnel A Temp. - I!igh 5 165*F 1 54.5*F 5 176*F hl C.c1 i

1 60*F . c2 f c) 1 7 h. Main Steam Line Turbine Bldg.

Q

g. Temp. - High 1 131.2*F

, 5 138*F u -

, 1. Manual Initiation NA NA i 1, i .

l

$$$$$N$R1 ' E APP l _

TABLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#

1. PRIMARY CONTAINMENT ISOLATION '

[ a. Reactor Vessel Water Level - Low Low, Level 2

b. Drywell Pressure - High (< 10(a))

c. Containment Building Continuous Containment (510(a))

Purge (CCP) Exhaust Radiation - High (b) ~( s 10(,))

d. Manual Initiation NA

2. MAIN STEAM LINE ISOLATION -

a. Reactor Vessel Water Level - Low Low Low, Level 1 < 1.0*(/< 10(a),,)

,, b. Main Steam Line Radiation - High (b)

c. Main Steam Lin.e Pressure - Low (51.0*(/<10(*)**))

d. Main Steam Line Flow - High < 1.0*(/< 10(*)**) '

e. Condenser Vacuum - Low 7 0.S*(/< 10(*)**)

NA

f. Maia Steam Line Tunnel Temp. - High NA

g. Main Steam Line Tunnel A Temp. -:High -

NA

. h. Main Steam Line Turbine Bldg. Temp. - High 3

1. NA u Manual Initiation -

NA ' .S

3. SECONDARY CONTAINMENT ISOLATION t

a. Reactor Vessel Water Level - Low Lcw, Level 2

b. Drywell Pressure - High (< 10(*))

c. Containment 81dg. Fuel ({10(a)) -

Plenum Radiation - Highgjnsfer Pool Ventilation

d. Containment Bldg. Exhaust Radiation - High(b)

(1 10(,))

(1 10(,)) g

e. g ContainmentBldg.Continuouscontainggt Purge (CCP) Exhaust Radiation - High (< 10(3)) L f.

Fuel 81dg.VentilagnExhaust Radiation - High (1 10(,)) it

.. g. Manual Initiation NA D

4. REACTOR WATER CLEANUP SYSTEM ISOLATION

~

a. A Flow - High (< 10(*)##)

b. A Flow Timer NA

c. Equipment Area Temp. - High NA

d. Equipment Area A Temp. - High' NA

e. Reactor Vessel Water Level - Low Low, Level 2 (< 10(,))

f. Main Steam Line Tunnel Ambient Temp. - High

f NA

g. Main Steam Line Tunnel A Temp. - High NA

h. SLCS Initiation NA --

1. Manual Initiation NA W

CLINTON - UNIT 1 3/4 3-21 AUG 2 91985

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M00F & HEW COPY THIS P GE OPEN PENDING RECEIPT OF i INFORMATION FROM THE APPLICANT TABLE 3.'3.2-3 (Continued)

ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds),

! 5. ~REkCTOR CORE ISOLATION COOLING SYSTEM ISOLATION __-

]

a. RCIC Steam Line Flow - High (< 10(a)ggg)

b. RCIC Steam Line Flow High - Timer i NX

c. RCIC Steam Supply Pressure - Low

d. RCIC Turbine Exhaust Diaphragm Pressure - High (< 10(a))

NX

e. RCIC Equipment Room Ambient Temp. - High f NA 3

f. RCIC Equipment Room A Temp. - High NA b

g. Main Steam Line Tunnel Ambient Temp. - High

h. Main Steam Line Tunnel A Temp. - High NA .Y

1. Main Steam Line Tunnel Temp. Timer NA U NA

~ j. RHR Equipment Room Ambient Temp. - High NA D

k. RHR Equipment Room A Temp. - High NA

1. Drywell Pressuie - High (< 10(,))

m. Manual Initiation , NA

6. RHR SYSTEM ISOLATION kg

a. RHR Equipment Area Ambient Temp. - High '

NA~

b. RHR Equipment Area A Temp. - High NA c.

d.

e.

RHR/RCIC Steam Line Flow - High Reactor Vessel Water Level - Lcw, Level 3 NA

(< 10(,)) d Reactor Vessel Water Level - Low Low Low, Level 1

f. ('<' 10(3))

Reactor Vessel (RHR Cut-in Permissive)

Pressure - High NA

. g. Drywell. Pressure - High NA

h. Manual Initiation NA

' (a) Isolation system instrumentaticn response time specified includes the diesel generator starting and sequence loading delays.

(b) Radiation detectors are exempt from response time testing. Response time shall be measured from detector output or the input of the first electronic component in the channel. ,

" Isolation system instrumentation response time for MSIVs only. No, diesel generator delays assumed. I

- Isolation system instrumentation response time for associated valves except MSIVs. ,

Isolation system instrumentation response time specified for the Trip Func-tion shall be added to isolation time shown in Tables 3.6.4-1 for valves and 3.6.6.2-1 for dampers to obtain ISOLATION SYSTEM RESPONSE TIME-for each -

valve / damper.

MTime delay of 45 to 47 seconds. .

M # Time delay of 3 to 13 seconds. -

CLINTON - UNIT 1 3/4 3-22 AUG I $ 1985

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l TABLE 4.3.2.1-1 (Continued) i E i' j

'2 ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS :

E! . !

i CHANNEL OPERATIONAL' i CHANNEL FUNCTIONAL CHANNEL CONDITIONS IN'WHICH

! E TRIP FUNCTION CHECX ' TEST CALIBRATION SURVEILLANCE REQUIRED 6.

RHR SYSTEM ISOLATION

a. RHR Equipment Area Ambient ;

Temp. - High S M R 1, 2, 3 I

b. RHR Equipment Area 1 a Temp. - High S M -

R, 1, 2, 3 ;

c. RHR/RCIC Steam Line

, Flow - High S .M R (b) 1, 2, 3 I

d. Reactor Vessel Water Level - .

Low, Level 3 S M R (b) 1, 2, 3

e. Reactor Vessel Water Level -

l Low Low Low, Level 1 S M R (b) 1, 2, 3 ,

w f. Reactor Vessel (RHR Cut-in D Permissive) Pressure - High S M R (b) 1, 2, 3 I w g. Orywell Pressure - High S H R (b) 1, 2, 3 j A m

h. Manual Initiation NA R NA 1, 2, 3

When handling irradiated fuel in either the secondary or the primary containment and during CORE ALTERATIONS ,

and operation i

- When reactor g:gp...,

gotential

^ for drainpg

. ; 0012) theany e i and/or reactor turbinevessel.

stop valve is open. - - -

, #0uring CORE ALTERATION and operations with a potentjal for draining the reactor vessel.

(a) Each train or logic channel shall be tested at least every other 31 days, l yu i

(b) Calibrate the analog trip modules at least once per*31 days. ' c .2 IM

.; : o. >

'l I

n et s

i=

g JJa, -Io q aper: N k laru 9/r/dr.: ader we/cnh. g! ,-

j g, g .

i

= m b

1 , !

I j

, I

,Li -

. PROOF & REVIEW COPY l

TABLE 3.3.3-1 (Continued)

EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION

. ACTION 30 - With the number of OPERABLE channels.less than requ' ired by the Minimum OPERABLE Channels per Trip Function requirement:

a. With one channel inoperable, place the inoperable channel in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

or declare the associated system inoperable.

b. With more than one channel inoperable, declare the. associated system inoperable.

ACTION 31 - Deleted.

ACTION 32 - With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement,

~

declare the associated ADS trip system or ECCS inoperable.

ACTION 33 - With the number of OPERABLE channels less than the Minimum OPERABL1 Channels per Trip Function requirement, place the g Q inoperable channel (s) in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . y ACTION 34 A fwith the numoer of UFtKAtilt Channels less than required by tne 1 Minimum OPERABLE Channels.per Trip Function requirement, verify bus power availability at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or declare j L the associated ECCS inoperable.

ACTION 35 - With the number of OPERABLE channels less than required by the .

Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or

, declare the associated ADS valve or ECCS inoperable.

ACTION 36 - With the number of OPERABLE channels less than required by the

~

Minimum OPERABLE Channels per Trip Function requirement:

a. For one trip system, place that trip system in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or declare the HPCS system inoperable.

b. For both trip systems, declare the HPCS system inoperable.

ACTION 37 - With the number of OPERABLE channels less than required by the Minimum OSERABLE Channels per Trip Function requirement, place at least ene inoperable channel in the tripped condition within I hour

or declara the HPCS system inoperable.

ACTION 38 With the nuroer of OPERABLE channels less than the Total Number of Channels, declare the associated emergency diesel generator inoperable and take the ACTION required by Specifications 3.8.1.1 or 3.8.1.2, as appropriate.

ACTION 39 - With the number of OPERABLE channels one less than the Total Number of Channels, place the inoperable channel isthe tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months *; operation may then continue until performance of the next required CHANNEL FUNCTIONAL TEST. ---

"The provisions of Specification 3.0.4 are not appliable. _

CLINTON . UNIT 1 3/4 3-31 AUG 2 91985 g g6 96#e hmwe g M **w. ***** ' ' ' ' * * * * ' * * -* ~

-mm.-. - ---p3- --- % ---*,s. -- . - , i.----.q.,- . ,.---.,.,,,,,,,3-.,

, ,-g --.w-- , . - y we_ ,. - ,.e, c- m ,-w-- g w. ---- p

PROOF & L REYlEW.r:- m..,

COMe TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES , _r

ECCS RESPONSE TIME (Seconds) 1.

, LOW PRESSURE CORE SPRAY SYSTEM {37

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM

a. Pumps A, dB, and 6 37

5. " u..p C 37

3. AUTOMATIC DEPRESSURIZATION SYSTEM NA -

4.

HIGH PRESSURE CORE SPRAY SYSTEM f27 S. LOSS OF POWER ,

NA a

CLINTON - UNI.T 1 3/4 3-35 AUG 1.S 1985

, ,, y . - - - --, m -m ,-, - - - . -

. =

I l

l ,

j i TABLE 3.3.4.1-2 ,.

l n l C ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION SETPOINTS -4

- D 2x ri -

E TRIP ALLOWABLE C) C') 1 l

c-TRIP FUNCTION SETPOINT VALUE

-5 c3 e

5

1. Reactor Vessel, Water Level -

Low Low, Level 2

}-45.5in.* M in.

1 ag ;, J

. dm t_. o l

, e 2o

i 2. Reactor Vessel Pressure - High { M psig ,

{ M psig -[j $

o 1o65

/oso 9,c' d 2 i

, '. - i -.3 1 rr. Z i

. p G2 1

a x :

I-, ~k :

Lc m, c 2 rr1 3 w

) >%

2 -t 1

[*

w o j 1 ~'1 i t

.e I

t i J .

. ~~ -

. i 3

ca.

, ; S:3 -

' ! ri

J St*

l ,

s. . ..q, -

"See Bases Figure 83/4 3-1. g

I re ,

- 2-C

,i ,,

m

, to 5 s

j si .

!2it .

i 1, 6 .

____..m__.._.

7._..____...-____7__.n_ ,

i P200F & EEW COPY I

INSTRUMENTATION END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION

~=

LIMITING CONDITION FOR OPERATION 3.3.4.2 The end-of cycle recirculation pump trip (EOC-RPT) system instrumen-tation channels shown in Table 3.3.4.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.4.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.4.2-3.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is > to 40% of RATED THERMAL POWER.

ACTION:

a. With an end-of-cycle recirculation pump trip function instrumentation channel trip se.tpoint less conservative than the value shown in the Allowable Value column of Table 3.3.4.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

b. An inoperable channel may be left'in inoperable status for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months

_ prior to placing it in the tripped condition, provided at least three

, OPERABLE channels'in the same trip' logic are~ monitoring that parameter.

c. With the number of OPERABLE channels one less than the Minimum OPERABLE Channels per Trip Function requirement:

1. For one trip function, place one channel in the tripped condition and restore the inoperable trip function to OPERABLE , status within 48 hcurs

- or reduce THERMAL POWER to less than 40% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

2. For both trip functions, place one channel in each trip function in the tripped condition and, restore at least one trip function to i OPERABLE status within one hour or reduce THERMAL POWER to less than 1 40% of RATED THERMAL POWER within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . l l

& cfslkItc&

M jaw.; 66~

~ &Mw h s; ,n 1 4 .+ c t/4 loS c i m

CLINTON - UNIT 1 3/4 3-43 AUG 2 9 to85

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

PiEF & EViB'l COPY j TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION ,e ACTION

. ACTION 70 -

a. With one of the required monitors inoperable, place the inoperable channel in the (downscale) tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ; restore the inoperable channel to OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , initiate and maintain operation of the control room emergency filtration system in the (i.oleti ..;f mode of operation. r ec*leculaMoo

b. With both of the required monitors inoperable, initiate and .

maintain operation of thej antrol room emergency filtration

- systeminthe(helo2l.w..;eh.

rt<#c enmode of operation within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 71 - With the required monitor inoperable, perform area surveys of the monitored area with portable monitoring instrumentation at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

e e

e O l

- )

i CLINTON - UNIT 1 3/4 3-63 .

. ,5

_ , . . . . . . . - - . . . ~ . . - . . . . - - .. - = = = . . . - * - "' --

m

- v. _ .s - . n., ,. - -. _ . . - ..

'I MCF -.-

& RB18Y CCPY INSTRUMENTATION REMOTE SHUTDOWN MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION ~ ~

t .

I i 3.3.7.4 The remote shutdown system instrumentation and controls shown in i Table 3.3.7.4-1 and 3.3.7.4-2ishall be OPERABLE.

3 % --- r*5P4 cfiW *Iy APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

a. With the number of OPERABLE remote shutdown system instrumentation channels less than required by Table 3.3.7.4-1, restore the inoperable channel (s) to OPERABLE status within 7 days or be in at least fiOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

. b. With the number of OPERABLE remote shutdown system controls less than required by Table 3.3.7.4-2, restore the inoperable control (s) to OPERABLE status within 7 days or be in at least HOT SHUTDOWN

. *w within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

~

I c. The provisions of specification 3.0.4 are not applicable.

.- SURVEILLANCE REQUIREMENTS .

4. 3. 7. 4. i Each of the above required remote shutdown system instrumentation channels shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3.7.4-1.

4.3.7.4.2 Each of the above remote shutdown control switches and control circuits shall be demonstrated OPERABLE by verifying its capability to perform its intended fenction(s) at least once per 18 months.

~ -

CLINTON - UNIT 1 3/4 3-72 29 E l

,-.-*--*-+~y-'-**.--r

--.- e

-r -,

^

r-- -- - w

P!!00F & REVIEW COPY TABLE 3.3.7.4-1 REMOTE SHUTDOWN MONITORING INSTRUMENTATION _ ,e

' DIVISION I

DIVISION II

~ .. .

MINIMUM MINIMUM EQUIPMENT CHANNELS EQUIPMENT CHANNELS INSTRUMENT NUMBER OPERABLE NUMBER OPERABLE

1. SRV SID Temp.;Lep.PecjTep. 1C61-R506 1 1C61-R512 1

2. SRV 51C Temp. 1 1C61-R513 SRV 51G Temp.,,LppF=ITeco.1C61-R507 1

3. Lpp. Pu)T,41C61-R508 1 1C61-R514 1

4. Supp. Pool. Lvl 1C61-R504 1 1C61-R511 1

. 5. RPV Lvl IC61dR010 1 1C61-R509 1

6. RPV Press. IC61-R011 1 1C61-R510 1

7. Upper DW Temp. IC61-R502 1 NA

8. Lower DW Temp. . 1C61-R501 1 NA

9. SX Strnr. Osch. Press.' 1C61-R503 -1 NA -

10. RCIC Cond. Tnk Lyl. 1C61-R505 1 NA

11. RHR Loop A Flow IC61-R005 1 NA

12. RCIC Turb. Speed 1C61-R003 1 NA

. 13. RCIC Flow 1C61-R001 1 NA 14.

RCIC Turb. Flow Cnti IC61-R001 1 hA

.. .y,.. . .. . . - . . . , . . . - , . ,_ ~ m CLINTON - UNIT 1 3/4 3-73 AUG 2 91985

.- g . we.- .e, --ea.m,o- g.e. e w .m e e m e . . o w.- e. .e e. +-e.. . . 4 % .--..-.--,.-es. --ew---. e . e- ..-g

--.n

i . , I

\ .

l

~ .

i TABLE 4.3.7.5-1 n .

h i C 1; ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Ei i{

E -l 8

APPLICABLE '!

I g l CHANNEL CHANNEL OPERATIONAL q INSTRUMENT CHECK CALIBRATION CONDITIONS

}

l H 1. Reactor Vessel Pressure M R 1, 2

2. Reactor Vessel Water Level l H. R 1, 2

3. Suppression Pool Water Level ,

H , R 1, 2, 3 e

.. 4. Suppression Pool Water Temperature M R 1,2,3

i t i H

5. , Drywell Pressure le 4 M R 1, 2 ?

6. Drywell Air Temperature , H R 1, 2

7. Drywell Hydrogen and Oxygen Concentration Analyzer M* 1, 2 Q*

, i y and Monitor

! [ 8. Containment Pressure M R 1, 2 h 9. Containment Temperature H R 1, 2

10. Containment Hydrogen and Oxygen Concentration Analyzer M* Q* 1, 2 -

l and Monitor i 11. Safety / Relief Valve Acoustic Monitor NA R 1, 2 I 12. Containment /Drywell High Range Gross Gamma Radiation H R** I

' 1,2,3 -

Monitors 1

13. HVAC Stack High Range Radioactivity Manitorf M 1,2,3 'C::3 I l 14. SGTS Exhaust High Range Radioactivity Monitor # O 1

M R 1,2,3 -vt l -

i

Accomplishedautomaticallyusinganintegralsarplegassupplycontaining(a)3.2vol.%hyhog'eh,""

l i

1.0 vol.% helium, 21 vol.% oxygen, 0.9 vol.% argon and 73.9 vol.% nitrogen. -

- The CHANNEL CALIBRATION shall consist of an electronic calibration of the channel, not including the

)go detector, for range decades above 10 R/hr and a one point calibration check of the detector below 10 R/hr with an installed or portable gamma source.

Q -

, ce #High range noble gas monitors. ,

t 40 YD

$ s g

I -

r

. (,

a

TABLE 3.3.7.9-1 FIRE DETECTION INSTRUMENTATION TOTAL NUMBER INSTRUMENT LOCATION OF INSTRUMENTS

HEAT FLAME SMOKE (x/y) (x/y) (x/y)

AUXILIARY BUILDING FIRE DETECTION ZONE A-la 7/0 A-lb 11/0 A-lc A-ld A-le

'A-2a 3/0 A-2b 3/0 A-2c 3/0 A-2d A-2e A-2f A-2g A-2h A-21 .

A-2j A-2k . 10/0 e A-21

's ,. A-2m .

1/0 _ _ .

A-2n~ ~

33/0 A-20 5/0 -

3/0

~

A 3a A-3b 2/0 A-3c A-3d . 10/0 A-3e 1/0

'A-3f 63/0 A-3g 5/0

. A-4 1/0 A-5 1/0 CONTAINMENT BUILDING FIRE DETECTION ZONE C-1 4/0 C-2 1/0

(x/y): x is number of Function A (early warning fire detection and notification only) instruments .

y is number of Function B (actuation of fire suppression '

systems and early warning fire detection and -

notification) instruments. ..

The fire detection instruments located within the containment -

are not required to be OPERABLE during the performance of Type A .

Containment Leakage Rate Tests. -

CLINTON - UNIT I 3/4 3-85 y yy, . wr y 7 % --- ----4 -y ,m-- 7 - - - + - g w-.-- .-

I a l TABLE 3.3. 7.9-1 (Continued)

FIRE DETECTION INSTRUMENTATION TOTAL NUMBER OF INSTRUMENTS *

"- ~~~

INSTRUMENT LOCATION HEAT FLAME SMOKE (x/y) (x/y) (x/y)

!! FUEL BUILDING FIRE DETECTION ZONE

. F-la F-lb 3/0 F-lc.

F-ld F-le F-lf F-lg i F-lh F-li

, F-lj F-lk F-Im ,

14/0 F-In F-lo

,~. F-1p 131/0 i 4.; . . _ . _. .

i -.

DIESEL GENERATOR BUILDING FIRE DETECTION ZONE D-1 D-2

.D-3 D-4a 0/5 D-4b D-5a 0/5 D-5b D-6a 0/5 D-6b D-7 D-8 To be supplied later D-9 D-10 24/0

ee M

6

.CLINTON - UNIT I 3/4 3-86

N *.**e%N =~ w .,.., +w, o ww , . . . . . , , . . , , . _ , , , _ , ,, ,

-vw

TABLE 3.3.7.9-1 (Continued)

FIRE DETECTION INSTRUMENTATION TOTAL FUMBER INSTRUMENT LOCATION OF INSTRUMENTS

HEAT FLAME SMOKE

~~-

- (x/y) (x/y)-(x/y)

CONTROL BUILDING FIRE DETECTION

, ZONE CB-la CB-lb -

CB-1c CB-ld 16/0 CB-le .

36/0 CB-lf 39/0 CB-lg 2/0 CB-lh CB-li 94/0 CB-2 5/0 CB-3a 6/0

- CB-3b 1/0 CB-3c 1/0 CB-3d

~

1/0 CB-3e 1/0 CB-3f 1/0

f. ~' CB-3g 1/0

's CB-4 5/0 CB-Sa 3/0 CB-Sb CB-Sc CB-6a 110/0 CB-6b 6/0 CB-6c 17/0 CB-6d 2/0 50/0 CB-7 6/0 PGC C Panels and Floor Sections 0/225 309/0

=

CLINTON - UNIT I 3/4 3-86a

;;i i

i j !!

9 i g TABLE 3.3.7.11-1 -

ll}

o e

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION i<

E y MINIMUM i-g CHANNELS INSTRUMENT OPERABLE ACTION

1. GROSS RADI0 ACTIVITY MONITORS PROVIDING

ALARM AND AUTOMATIC TERMINATION OF RELEASE i

! a. Liquid Radwaste Discharge Process Radiation Monitor 1- 110

2. GROSS BETA OR GANIA RADI0 ACTIVITY MONITORS i PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC w TERMINATION OF RELEASE '

I h .>

w a. Plant Service Water Effluent Process Radiation Monitor 1 Ill i m

b. Shutdown Service Water Effluent Process Radiation Monitor 1/ Division p 11.1 I

c. Fuel Pool lleat Exchanger Service Water Radiation Monitor 1 111

3. FLOW RATE MEASUREMENT DEVICES l .

a. Liquid Radwaste Effluent Line l 1 112 m ,,

b. Plant Service Water Effluent' Line 1 112

. c. Plant Circulating Water Effluent Line 1 ,. 112 y m ;

3=

!8:

@

CHANNEL si g CHANNEL SOURCE CHANNEL FUNCTIONAL '

INSTRUMENT CHECK CHECK CALIBRATION TEST

1. GROSS RADI0 ACTIVITY MONITORS PROVIDING
b ALARM AND AUTOMATIC TERMINATION OF RELEASE' U
' I
, a. Liquid Radwaste Discharge Process !
i
'i Radiation Monitor Effluent Line D; P R(3) Q(1) ,
, L!
w 2. GROSS BETA OR GAMMMA RADI0 ACTIVITY MONITORS l
' ~
) PROVIDING ALARM BUT NOT PROVIDING AUTOMATIC g m TERMINATION OF' RELEASE d>
N a.
Plant Service Water Effluent Process-I
Radiation Monitor j D M R(3) Q(2) !
i
\
b. Shutdown Service Water Effluent Process '
Radiation Monitor D M R(3) Q(2) -
c. Fuel Pool lleat Exchanger Service Water
Radiation Monitor D M R(3) Q(2) o
.m
3. FLOW RATE MEASUREMENT DEVICES go i'
.. ;all3
a. Liqpi.d Radwaste Effluent Line '
D(4) NA R 'Q
,.,,, g
b. Plant Service Water Effluent Line D(4) NA R Q
, g _ ,.
on+c<c..u,4;u m sma,,t w,; nm ::,. ;; ;
. m M
M I!
. ' . 8; i:
.1 .
. 5
, TABLE 4.3.7.12-1 j P
g RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,
p{
e E '
CHANNEL MODE IN WilICil 3
CHANNEL SOURCE CHANNEL FUNCTIONAL- SURVEILLA! ICE i E INSTRUMENT CHECK CHECK ~ CALIBRATION TEST REQUIRED I.I,
1. POST-TREATMENT AIR EJECTOR l H '
i OFF-GAS PRM l a. High Range Noble Gas Activity D D -
R(2) Q(1) ..
Monitor Providing Alarm and Automatic
i Termination of Release f b. Effluent System Flow-Rate Monitor D[ NA RW Q# - *
! c. PRM Flow-Rate Monitor 0] NA RF Q -NA"- .
2. STATION HVAC EXHAUST PRM , ,'
l R a. High-Range Noble Gas Activity D,' M R(2) Q(1)
, . Monitor ,
b. Low-Range Noble Gas Activity 0 'M R(2) Q(1) i I
I 8 Monitor .
i t c. Iodine Sampler W NA NA NA
l d. Particulate Sampler Wl NA NA NA * '
l e. PRM Flow Rate Monitor D' NA R Q
f. . Effluent System Flow Rate Monitor D NA R Q
, 3. STANDBY GAS TREATMENT SYSTEM EXilAUST PRM- *10 J
a. High-Range Noble Gas Activity D M R(2) Q(1)
C2 l Q
' Monitor j
b. M'edium-RangeNobleGasActivityMonitor D M R(2) Q(1)
Kao , . ,
, c. Low-Range Noble Gas Activity Monitor D M R(2) Q(1)
* ~!
s
d. High-Range Iodine Sampler W NA NA NA
Q
-~
6 r
e. Low-Range Iodine Sampler W NA NA NA I
I, g f. Particulate Sampler W NA NA NA
g g.
h.
PRM Flow-Rate Monitor Effluent System Flow-Rate Mo.ii'.or D
D.
NA NA R Q l
R Q 2 . 1, e
--,c.._.-----= _ 3 ~ : -- - x - - - - - - - -- - _3.-
N00F & REV!EW COPY TABLE 4.3.7.12-1 (Continued)
TABLE NOTATION ,p ,
At all times. - -
a During main condenser off gas system operation.
t During operation of the main condenser air efector. --
(1) The CHANNEL FUNCTIONAL TEST shall also demonstrate that control room alarm annunciation occurs if any of the following conditions exist;
1. Instrument indicates measured levels above the alann setpoint.
2. Circuit fa'ilure.
3. Instrument indicates a downscale failure.
4. Instrument controls not set in operate mode.
(2) The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained ftpm suppliers that participate in measure-
ment assurance activities with NBS. These standards shall permit calibrat-ing the system over its intended range of energy and measurement range.
. Subsequent CHANNEL CALIBRATION shall be performed using the initial radio-active standards or other standards of equivalent quality or radioactive sources that have been related to the initial calibration.
(3) The CHANNEL CALIBRATION shall include the use of standard samples contain-ing a nominal:
1. 1.0 vol. % hydrogen, balance nitrogen, and
2. b vol. % hydrogen, balance nitrogen.
~
~
CLINTON-UNIT 1 3/4 3-102 AUG 2 91985 m u- a a _
0 4
l TABLE 3.3.9-1 ,
l 0 PLANT SYSTEMS ACTUATION INSTRUMENTATION i i E
'{i o TRIP FUNCTION
[er CeedW4Me/p/ APPLICA8LE . ' ! !
l
_1. CONTAINMENT SPRAY SYSTEM
c gfT 2,#! MINIMUM N PERABLE CHANNELS OPERATIONAL CONDITIONS
lI e a. Drywell Pressure-High 2 1,2,3 n i
b. Containment Pressure-High 2 1,2,3 i
c. Reactor Vessel Water Level-Low Low Low, Level 1 2 1,2,3 i d. Timers
)'li (1) B (so Min d , 1 1,2,3 ;
(2) Loop Loop BA,enly Loop (Te sec.) 1 1,2,3 ,
f y e. Manual Initiation 1 1,2,3 Mialmewei Oya6lc
{ 2. FEE 0 WATER SYSTEM / MAIN TUR8INE TRIP SYSTEM C6mnacM 4
Reactor Vessel Water Level-High, Level 8
!] O a. 3 1
3. SUPPRESSION POOL MAKEUP SYSTEM ACTION
{j
!' a. Drywell Pressure-High 2 1,2,3 50 9
Reactor Vessel Water Level-Low Low Low, !
j b. .
- Level 1 2 1,2,3 50 6]
cs j
2* 51 9
c. Suppression Pool Water Level-Low Low 1, 2, 3 i M l l
' d. Suppression Pool Makeup Time 1 1,2,3 51 :: s
'i
e. SPMS Manual Initiation 2* 1,2,3 51 2
ij
' l'
> 1,2,3 51 g f. SPMS Mode Switch Permissive 1 n c:3 l'
*Two trip systems with two-out-of-two logic. ', M i !
.u.
i
n 't ,
, i, a
s
-: i .
1 I
( TABLE 3.3.'9-2 -
3 ;
PLANT SYSTEMS ACTUATION INSTRUMENTATION SETPOINTS ,
E
- ALLOWABLE. .
TRIP FUNCTION TRIP SETPOINT VALUE
c . I
1. CONTAINMENT SPRAY SYSTEM i i
j a. Drywell Pressure-High 5 1.68 psig < 1.88 psig.
b. Containment Pressure-High , 5 23.0 p.sia 5 23.5 psja '
Reactor Vessel Water Level-Low Low Low, Level 1
c. ->-145.5 in.* ~~>-147.7 in.
l d. Timers l 1. $ 10.17 min. > 10.10 5 10.23 min.
2. Loop Loop B Aenly , Loop qosec.) (B (,m min.) -
5 90 sec. 5 90.6 sec.
2. FEEDWATER SYSTEN/ MAIN TURBINE TRIP SYSTEM
a. Reactor Vessel Water Level-liigh, Level 8 s 5 52.0 in.* 5 52.6 in! -
D 3. SUPPRESSION. POOL MAKEUP SYSTEM I
a. Drywell Pressure-High f 5 1.68 psig 5 1.88 psig'
n tu .
j b. Reactor Vessel Water Level-Low Low Low, .
l 1
Level 1 > -145.5 inches * > -147.7 inches t
l. '
c. Suppression Pool Water Level-Low Low d > El. 730'-1 9/16" > 729'-0" .___.-
d. Suppression Pool Makeup Timer
> 25 minutes 30 minutes' E c3
e. SPMS Manual Initiation NA NA $ 4
f. SPMS Mode Switch Permissive NA NA
z
, 4
} *See Bases Figure B 3/4 3-1.
' b ! .
k '
g'
' P 0 ,
8
i 8
}' ;
J =
I-TABLE 4.3.9.1-1 .
P g PLANT SYSTEMS ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E CilANNEL OPERATIONAL .
8 l CHANNEL FUNCTIONAL CHANNEL CONDITIONS'IN WHICH
l
, E TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED ti 1
, 1. CONTAllWENT SPRAY SYSTEM r,
a. Drywell Pressure-liigh 5 H 1, 2, 3 l
b. ~ Containment Pressure-High S H Rfa R* 1,2,3
, c. Reactor Vessiel Water Level-Low -
Low Low, Level 1 S H Rg ,) 1, 2, 3 l
d. Timers HA H R 1, 2, 3

g
e. Manual Initiation NA , H_ NA 1, 2;. 3 8
~~
2. FEEDWATER SYSTEH/ MAIN TURBINE TRIP SYSTEM l l w -
1 Reactor Vessel Water Level-High, i ) a.
i w Level 8 S H R 1 , ,
i e l h 3. SUPPRESSION POOL HAKEUP
-[ a. Drywell Pressure-High S H R(a) 1, 2',' 3 i
b. Reactor Vessel Water Level - = ,
RI ")
~ ~ ' ~
Low Low Low, Level 1 S H 1, 2i 3 .I l o
c. Suppression Pool Water Level-Low Low S I $
H RIJ 1, 2', 3 c2 i
r
d. Suppression Pool Hakeup Timer * ' '
NA H Q 1,2,3 R* ,
, e. SPHS Hanual Initiation NA R NA J .* .' ,1,. *2,, 3 s f.
SPHS Mode Switch Paraissive NA R NA' 1, 2,' 3 q
i
! i
==
(a) Calibrate the analog trip module at least once every 31 days. O 3 .
8?
(b) Calibnda 4ke 4 rip knit (Act4) a+ (cad once: every 31 clays
~
t 4
t $
. _ .. .- -..;._..a..
PROC & REVIEW COPY REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION .
3.4.3.1 The following reactor coolant system leakage detection systems shall be OPERABLE:
a. The drywell atmosphere particulate radioactivity monitoring system,
b. The drywell sump flow monitoring system, and
c. Either the drywell atmosphere gaseous radioactivity monitoring system or the drywell air coolers condensate flow rate monitoring system.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and,3.
ACTION: .
With only two of the above required leak' age detection systems CPERABLE, opera-tion may continue for up to 30 days provided grab samples of the drywell atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactive monitoring system is inoperable; otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTOOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
. SURVEILLANCE REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demon-strated OPERABLE by: .
a. Drywell atmosphere particulate and gaseous monitoring systems performance
,, of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBP.ATION at least once per 18 months.
b. Drywell sump flow monitoring system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION TEST at least once per 18 months.
c. Drywell air cooler condensate flow rate monitoring system performance of a CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBPATION at least once per 18 months. -
d. Flow-testing the drywell floor drain sump inlet piping for blockage at least once every 18 months during shutdown. -
M (guire,1 ok ?dve.r hfd or drartd 6alf CLINTON - UNIT 1 3/4 4-9 0 w w. - - - -
l- .
ll a
i s j TABLE 4.4.6.1.3-1 '
'd t n C
t:
REACTOR VESSEL MATERIAL SUWEILLANCE PROGRAM-WITHDRAWAL SCHEDULE r-j I
o e z . i.
CAPSULE VESSEL LEAD LEAD k WITHDRAWAL. TIME j' g NUMBER LOCATION FA et.Es) FACTOR at T (EFPY) ,, I-
[ 1. Capsule 1 3" , c,7 0.89 M (, '!L l 2. Capsule 2 177* 67 0.89 , M 15 i
3. Capsule 3 183* .(r7 0.89 - M Ecg_ ,.
, t I
4 1
1 !
M
, c' l
~
. l, 1
7s.. 3 CS: ,
c.' j ,
"T1 * .
{ f m.
]
1- :e. ' . , n, m
,9 i 2:= A ;
4
o C
k a' :
l g -
CD i-i
. " ca ,
it. .
<ts-: j.'.i
! $ j
m 4
i e l, I
... : L .. .
~
..- . . . . ~ .~. : . . a c. . a..
~^ '
~ ~ ~~ ^ ~
... ~~~~~. . .
. PROOF & f3 EW COPY L .._
REACTOR COOLANT SYSTEM 3/4.4.9 RESIDUAL HEAT REMOVAL '
HOT SHUTDOWN _y
~
LIMITING CONDITION FOR OPERATION l
~
. 3.4.9.1 Two# shutdown cooling mode loops of the residual heat removal (RHR) system shall be OPERABLE and, unless at least one recirculation pump is in operation, at least one shutdown cooling mode loop shall be in operation *'##
with each loop consisting of at least:
a. One OPERABLE RHR pump, and
b. One OPERABLE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITION 3, with reactor vessel pressure less than the RHR cut-in permissive setpoint.
ACTION: -
~
a. With less than the abcVe required RHR shutdown cooling mode loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible. Within one hour and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, demonstrate the operability of at least one _ _ _ .
alternate method capable of decay-heat removal for each inoperable RHR shutdown cooling made loop. Be in at least COLD SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .**
b .' With no RHR shutdown cooling mode loop or recirculation pump in operation, immediately initiate corrective action to return at least one RHR shut-down cooling mode loop or recirculation pump to operation as soon as possible. Within one hour establish reactor coolant circulation by an alternate method and monitor. reactor coolant temperature and pressure at least once per hour.
SURVEILLANCE REQUIREMENTS
'~
4.4.9.1 At least one shutdown cooling made loop of the residual heat removal system, one recirculation pump, or alternate method 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 shutdown cooling mode 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 surveillance testing provided the other loop is OPERABLE and in operation.
The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE. .
.##The RHR shutdown cooling mode loop may be removed from operation'during -
hydrostatic testing.
Whenever two or more RHR subsystems are inoperable, if unable to attain COLD [.
SHUTDOWN as required by this ACTION, maintain reactor coolant temperature as .
low as practical by use of alternate heat removal methods'.
CLINTON - UNIT 1 3/4 4-26 AUG 231965
. PROOF & BLEW COPY 4
REACTOR COOLANT SYSTEM COLO SHUTDOWN LIMITING CONDITION FOR OPERATION .
3.4.9.2 Two# shutdown cooling mode loops of the residual heat removal (RHR)
~
-- , system-shall be- OPERABLE unless one recirculation pump is in operation, the
htleastoAeshutdowncoolingmodeloopshallbeinoperation**##witheach loop con.usting of at least

~
a. Ona OPERABLE RHR pump, and
b. One OPERABLE RHR heat exchanger.
APPLICABILITY: OPERATIONAL CONDITION 4.
ACTION:
^
a. With less than the above required RHR shutdown cooling mode loops OPERABLE, within one hour and at least once pr.r 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter demonstrate the operability of at least one alternate method capable of decay heat removal for each inoperable RHR shutdown cooling made loop.
. b. With no RHR shutdown cooling mode loop or recirculation pump in operation, within one hour establish reactor coolant circulation by an alternate method and monitor reactor coolant , temperature and pressure at least once per hour. - --
SURVEILLANCE REOUIREMENTS-4.4.9.2 At least one shutdown cooling mode loop of the residual heat removal system, recirculation pump, or alternate method 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 shutdown cooling made 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
_ surveillance testing provided the other loop is OPERABLE and in operation.
The shutdown cooling pump may be removed from operation for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period provided the other loop is OPERABLE. '
The shutdown cooling mode loop may be removed from operation during -
hydrostatic testing. ,
e CLINTON - UNIT 1 3/4 4-27 AUG 2 3 ESS
. _ , . . _ , _ _ . _ . . ;.. ---- ._. 1 ::2 -. .
.1- i 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - OPERATING -
LIMITING CONDITION FOR OPERATION'
}' _
3.5.1 ECCS divisions I, II and III shall be OPER.18LE with:
a. ECCS division I consisting of:
(
~ 1.
The OPERABLE low pressure core spray (LPCS) system with a flow path .
capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel.
2. The OPERABLE low pressure coolant injection (LPCI) subsystem "A" of the RHR system with a flow path capable of taking suction from the -
suppression pool and transferring the water to the reactor vessel. *
3. Seven OPERABLE ADS valves.
. b. ECCS division II consisting of:
1. T'a OPERABLE low pressure coolant injection (LPCI) subsystems "B" and "C" of the RHR system, each with a flow path capable of taking suction from the suppression pool and transferring the water to the reactor vessel.
'_ 2. ,
Seven OPERABLE ADS valves. ,
, c. ECCS division III consisting of the OPERABLE high pressure core-spray (HPCS) !
system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor or RcIc s4%. hnk APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2*'# and 3*[*
] RS-Ti i
l "The A05 is not required to be OPERASLE when reactor steam dome pressure is less than or equal to 100 psig. ,
See Special Test Exception 3.10.5. '
49 Ons LPCI sub.sysfem of Sr. RNA system may bc <// ned' f in 4he AuNewn coo /I l'ene/ar vessel
.. y made 3 a . . .when ~ , ,. . ..... <,. a . .
l C.LINTON - UNIT 1 3/4 5-1 AUG 2 91985 e ,- .-_,.----.e - , . - -
PROOF & REVIEW COPY l
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS 4.5.1 ECCS divisions I, II and III shall be demonstrated OPERABLE by:
a.
l
At least once per 31 days for the LPCS, LPCI and HPCS systems

! ' 1.
i Verifying by venting at the high point v'ents that the system piping from the pump discharge valve to the system isolation valve is filled i wit!) water.
1 2.
Verifying that each valve (manual, power operated 'or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct
position.
b. Verifying that when tested pursuant to Specification 4.0.5 each:
1. LPCS pump develops a flow of at least 5010 gpm against a test line i
pressure greater than or equal to (119) psid.
2. LPCI pump develops a flow of at least 5050 gpm against a test line pressure greater than or equal;to (119) psid.
3. HPCS pump develops a flow of at least 5010 gpm against a test line pressure greater than or equal to (490) psid.
c.
For the LPCS, LP'CI and HPCS systems, at least once per 18 months performing a system functional test which includes simulated automatic actuation of the system throughout its emergency operating sequence and verifying that each automatic valve in the flow path actuates to its correct position.
Actual injection of coolant into the reactor vesseT may.be excluded from this test. -
d. For the HPCS system, at least once per 18 months, verifying that the suction is automatically transferred from the RCIC storage tank to the suppression pool on a RCIC storage tank low water level signal and on a suppression pool high water level signal. i
!' e. or the ADS by:
1- A* N -+ :. em Si uop,.,,_,'2r ,
.e
e. . o.. : . ; C"" enu m gua., 7;;7 y7 7 g m,, "e,_
p - _ . _. 3 , u u ,. 6, ,,, -
, _ , , y, , ,,,
v +--
ld At least once Der 18 monthsberforming a system functional test which I
l includes simulated automacic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
Except that an automatic valve capable of automatic return to its ECCS posi- -
tion when an ECCS signal is present may be in position for another made of "
operation. -
CLINTON - UNIT 1 3/4 5-4 AUG 2 91985
. . . = - . = . : .
a- + - . --
_--- _=---
. PR00F & REvi&Y CDPY l
EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2, [ Manually opening each ADS valve when the reactor' steam dome pressure is greater than or equal to 100 psig* and observing that:
~
. a. .
The control valve or bypass valve position responds' accordingly,
. or -
b. There is a corresponding change in the measured stream flow, and
c. The acoustic tail pipe monitor alarms.
3 [ Performing a CHANNEL CALIBRATION of the accumulator low pressure alarm system and verifying an alarm setpoint of > 140 psig on decreasing pressure.
/ .- -
"The provisions of Specificatio'n 4.0.4 are not applicable provided the surveil-lance is performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure is adequate -
to perform the test. _.-
CLINTON - UNIT 1 3/4 5-5 AUG 2 91985-e - , . - . .
~ ~ ~ ~ - -~ ~ ~ ~~ ~
-.- . . . ~' d 2" O ....;.t.._. .._ . . . __ i-~~ T_~~ .
I im00F & REVIEW COPY EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of 'the following shall be OPERABLE: 1-
, a. The low pressure core spray (LPCS) system with a flow path capable of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel. i l
b. Low pressure coolant injectior) (LPCI) subsystem "A" of the RHR system with 1
a flow path capable of taking suction from the suppression pool and trans-ferring the water to the reactor vessel. '
1 l
c. Low pressure coolant injection (LPCI) subsystem "B" of the RHR sys' tem with a flow path capable of taking suction from the suppression pool and trans-l ferring the water.to the reactor vessel.
d. Low pressure coolant injection (LPCI) subsystem "C" of the RHR system with a flow path capable of taking suction from the suppression pool and trans-ferring the water to the reactor vessel.
e. The high pressure core spray (HPCS) system with a flow path capable of taking suction from one of the following water sources and transferring the water through the spray sparger to the reactor vessel:
1. From the suppression pool, or
2. , When the suppression pool level is less than'the' limit or is drained, from the RCIC storage tank containing -at least 125,000 available gallons of water, equivalent to a level of 95L -
. APPLICABILITY: OPERATIONAL CONDITIONS 4 and 5*.
. ACTION:
a. With one of the above required subsystems / systems inoperable, restore at least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or suspend all operations that have a potential for draining the reactor vessel.
b. With both of the above required subsystems / systems inoperable, suspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel. Restore at least one subsystem / system to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or establish PRIMARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .
[ :.
"The ECCS is not required to be OPERABLE provided that the reactor vessel head -
is removed, the cavity is flooded, the reactor cavity to steam dryer pool ---
gate is m uoves and water level'in these upper containment pools is maintained -
within theilimits of Specification 3.9.8 and 3.9.9.
4* open CLINTON - UNIT 1 3/4 5-6 M 2S E
. _ , _,- .. .. __ -. . . . . - - - - - - - - - - - - - - - - - - - - - - ~ -
T , --, e - - - - , , , - - ,y-----,------wyy w ---y --- g-g---m--wy-ggc%,_w-g---g , g- w- *. ,y--w-, - - - , ,y --.wg mpy m y. -,.,y-w-,g---w-ey-+.me*mw-myy.--yve.w+++,rvow ww
- = . - -- -- - - - -.
n
! PM0F & R&iEW COl i
EMERGENCY CORE COOLING SYSTEMS l
3/4.5.3 SUPPRESSION POOL.
LIMITING CONDITION FOR OPERATION -
~~
\
LS 3 The suppression pool shall be OPERABLE:
8 n
a. In OPERATIONAL CONDITIONS 1, 2 and 3 with a contained water volume of at !
least 146,400 ft 3, equivalent to a level of 18'11".
b.
'In OPERATIONAL 3 CONDITIONS 4 and 5* with a contained water volume of at least 98,700 ft , equivalent to a level of 12'8", except that the suppres-
' sfon pool level may be less than the limit or may be drained provided that:
1. No operations are performed that have a potential for draining the
, reactor vessel,
2. The reactor inode switch is locked in the Shutdown or Refuel position, !
3. The RCIC storage tank contains at least 125,000 available gallons
~% of water, equivalent to a level of 95%, and
4. The HPCS system is OPERABLE per Specification 3.5.2 with an OPERABLE flow path capable of taking suction from the RCIC storage tank and-transferring the water through the spray sparger to the reactor vessel. -
APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2, 3, 4 and 5".
. ACTION:
. a. .In OPERATIONAL CONDIT[0N 1, 2 or 3 with the suppression pool water level
~
less than the above limit, restore the water level to within the limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and
in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
~
b. In OPERATIONAL CONDITION 4 or 5* with the suppression pool water level less than the above limit or drained and the above required conditions ;
not satisfied, suspend CORE ALTERATIONS and aril operations that have a 1
, potential for draining the reactor vessel and lock the reactor mode switch ;
in the Shutdown position. Establish SECONDARY CONTAINMENT INTEGRITY with- ..
1
. in 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . '
l "The suppression pool is not required to be OPERABLE provided that the reactor vessel head is removed. the cavity is flooded, the reactor cavity- to steam ,
dryer pool gate istremoved) and the water level in these upper containment -
pools is maintained witninyhe limits of Specifications 3.9.8 and 3.9.9.
l I'open T CLINTON - UNIT 1 3/4 5-8 AUG 2 91585
. . = _ . .
., m m . . _ .. . _ . _
PR00F & REVH COPY EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS
.. _ _. , 4.5.3._1___The suppression pool shall be determined OPERABLE by verifying the i water level to be greater than or equal to ;
j
a. 18'11" at l' east once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , in OPERATIONAL CONDITIO 'i,2or3..
{ l
b. 12'8" at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , in OPERATIONAL CONDITIONS 4 and 5.
4.5.3.2 With the suppression pool level less than the above' limit or drained in OPERATIONAL CONDITION 4 or 5*, at least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s:
a. Verify the required conditions of Specification 3.5.3.b to b,e satisfied, or
b. Verify footnote conditions
to be satisfied.
"The suppression pool is not required to be OPERABLE provided that the reactor. '
vessel head is removed, the cavity is flooded, the reactor cavity to steam dryer pool gate is e movecq and the water level is maintained within the limits -
of Specifications 3.s.e and $.9.9. "
l'Open AUG 2 91985 CLINTON - UNIT l' 3/4 5-9
-,,-,v. -,w.-,. -
P!!00F & ElH COPY CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE LIMITING CONDITION r0R OPERATION r
3. 6.1. 2 Containment leakage rates shall be limited.to: -
. a. An overall integrated leakage rate of less than or equal to':#
~
i 1. La,.0.65 percent by weight of the containment air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at
. Pa, 9.0 psig.
b. A combined leakage rate 'of less than or equal to 0.60 La, for all penetra-tions and all valves subject to Type B and C tests when pressurized to Pa,
.0 psig.
ss than or equal to 28 scf per hour for any one main steam line through the isolation valves when tested at Pa, 9.0 psig.
d. .A combined leakage rate of less than or equal to 0.08 La, for all penetra-tions shown in Table 3.6.4-1 of Specification 3.6.4 as secondary contain-ment bypass leakage paths when pressurized to Pa 9.0'psig.
e. A combined leakage rate of less than or equal to 1 gpm times the total number of ECCS and RCIC containment. isolation valves in hydrostatically
. tested lines which penetrate the primary containment, when tested at
. 1.10 Pa, 9.9 psig.
APPLICABILITY: OPERATIONAL CCNDITIONS 2*Aa d 3. _ _ .
ACTION: .
With:
a. The measured overall integrated containment leakage rate exceeding .
0.75 La, or
b. The measured combined leakage rate for all penetrations and all valves subject to Type B and C tests exceeding 0.60 La, or
c. The measured leakage rate exceeding 28 scf per hour for all four main steam lines through the isolation valves, or
d. The combined leakage rate for all penetrations shown in Table 3.6.4-1 as secondary containment bypass leakage paths exceeding 0.08 La; or
e. The measured combined leakage rate for all ECCS and RCIC containment isolation valves in hydrostatically tested lines which penetrate the pri-mary containment exceeding 1 gpa times the total number of such valves,
~
~
%1EumpUon fs A ppankt< J o$ 10C.fR 50 (. S S E R ~2 , P} G ~ G ,7 )
fr"See Special Test Exception 3.10.1. ]
CLINTON - UNIT 1 3/4 6-2 AUG 2 91985 a
- y, -- . -..gw y.-.w. . . , , - - - , , . , - , - . , - - - , , _ _ . , . - , - - - , -9.,.-%.v.---. - - - w.yr,.,.yyv.y-e--y, ,--- ,--r,,ww,.--,.-y-,.---_ , .,,..,,.e w ---., w
.=.- - -- - - -
.! PRODQRa!N C
. I CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS ..:
, LIMITING CONDITION FOR OPERATION T
.: 3.6.1.3 Each containment air lock shall be OPERABLE with:
a. Both. doors closed except when the air lock is being used for normal tran-
sit entry and exit through the containment, then at least one air lock door shall be closed, and
b. An overall air lock leakage rate at Pa, 9.0 psig:
1. For the personnel air lock, elevation 823'-3", of less than or equal to 0.02 La.
2. For the personnel air lock, elevation 741'-0", of less than or equal to 0.05 La. - ,
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2* and 3.
ACTION:
a. With one containment ~ air lock dooi inoperable:
1. Maintain at least the OPERABLE air lock door closed and either restore the inoperable air lock door to OPERABLE status within
. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or lock the OPERABLE air lock door closed.
2. Operation may them continue until performance of the next required overall air lock leakage test provided that the OPERABLE air lock-door is verified to be locked closed at least once per 31 days.
3. Othemise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and
, in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
4. The prov'isions of Specification 3.0.4 ar,e not applicable.
b. With the containment air lock inoperable, except as a result of an inoper-able air lock door, maintain at least one air lock door closed; restore 2 the inoperable air lock to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the followina 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
. With one containment air lock door inflatable seal system air; flask pressure instrument channel inoperable, restore the inoperable channel
to OPERABLE status within 7 days or verify air flask pressures to be
> 90 psig at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
. "See Special Test Exception 3.10.1.
CLINTON - UNIT 1 3/4 6-5 AUG 2 91995
. . . . . - . . . . . . . . . _ . - . . .. .- - ~ - - - - - -
l
. PP,00F & HEW COPY l 1
CCNTAINMENT SYSTEMS SURVEILLANCE REOUIREMENTS
_ . _ . .4.6.1.3. .Each containment air lock shall be demonstrated OPERABLEi-=
a. Within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months following each closing, except when the air lock is being
used for multiple entries, then at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ,"by verifying seal leakage rate less than or equal to 5 scf per hour when the gap between the door seals is pressurized to Pa, 9.0 psig.
^
b. By conducting an overall air lock leakage test at Pa, 9.0 psig, and verifying that the overall air lock leakage rate is within its limit:
1. At least once per 6 months # ,
2. Prior to establishing PRIMARY CONTAINMENT INTEGRITY when maintenance
'~
has been performed on the air lock that could affect the air lock sealing capability.
c. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.
f*%
((, _ _ . . . . . . . _ . . ..
~
The provisions of Specification 4.0.2 are not applicable. .
CLINTON - UNIT 1 3/4 6-6 AUG 2 S 1985
__ . . . _ . . . _ - - - - - ~ ~
CONTAINMENT SYSTEMS l DRYWELL STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION 3.6.2.4 The structural integrity of the drywell shall be maintained at a level
I consistent with the acceptance criteria in Specification 4.6.2.4.
I APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. 1 ACTION:
With the structural integrity of the drywell not conforming to the above require-ments, restore the. structural integrity to within the limits within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in within theatfollowing least HOT 24 SHUTDOWN hours. within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and.in COLD SHUTOOWN SURVEILLANCE REQUIREMENTS 4.6.2.4.1 The structural integrity of the exposed accessible interior and exterior surfaces of the drywell shall be determined during the shutdown for each Type A containment leakage rate test by a visual inspection of those sur-faces. This inspection shall be p formed prior to the Type A containment leakage rate test to verify no appu ent changes in appearance or other abnormal degradation.
4.6.2.4.2 Reports Any abnormal degradation of the drywell structure detected l during the above required inspections shall be reported to the Commission pur-suant to Specification 6.9.2. This report shall include a description of the condition of the concrete, the inspection procedure, the tolerances on cracking, and the corrective actions taken.
/Nce P l* <-
? TOse. r% alt *mest 7 M a. Tarvis (la nt.4 h g,h d'.n a t,ktM .
z o + pac + 09.RST5 CLINTON - UNIT 1 3/4 6-18 AUG 2 91985
- . . . . . . - - - .- -- - ---- - - - - - - - - ~ ~ - * ~ ~ ^
-p,,. . - _ ,--
__c.__ .--7-- ...y-, _ - - ~ - - - - - - - - , - - - - - - -
i I
'. n TABLE 3.6.'4-1
?
CONTAINHENT AND ORYWELL ISOLATION VALVES o i
i *
' ' HAXIMUM SECONDARY 8 APPLICABLE ISOLATION CONTAINHENT TEST ,
j E VALVE PENETRATION ISOLATION OPERATIONAL TIME BYPASS PATil PRESSURE ;]
NUMBER NUMBER SIGNAlt , CONDITIONS (Seconds) (Yes/No) (psig) i l 1. Automatic Isolation Valves
a. Primary Containment j
i U
1) Hain Steam Line C 5 1, 2, 3 No 9.0 i 1821-F022C C,D,E,F,G,H,J,U 3-5' 1821-F02BC C,D,E,F,G,H,J,U 3-5
, 1821-F067C C , D , E , F , G , 11, J , U 21 .
, w 2) Hain Steam Line A 6 1,2,3 No 9.0 h l ) 1821-F022A C,D,E,F,G,H,J U 3-5 '
. 1821-F028A C, D, E, F, G,'H, J, U 3-5 1 J, 1821-F067A C,D,E,F,G,H,J,U 21
. O
} 3) Hain Steam Line D 7 1,2,3 No 9.0 1821-F0220 -
C,D,E,F,G,H,J,U 3-5 1821-F0280 C,D,E,F,G,H J,U 3-5 '
, 1821-F067D .C,D,E,F,G,H,J,U '21
~
4) Hain Steam Line B 8 1,2,3 No 9.0 1821-F0228 1821-F028B C,D,E,F,G,H,J,U C,D,E,F,G,H,J,U 3-5 3-5 3
C3 IB21-F0678 . C,D,E,F,G,H.J,U 21 Q i
5) feedwater/RI,lR. Line A 9 1,2,3 l[ ..,,. 9.0
j 1821-F032A B, L 0.5 Ye,s 3D 1E12-F053A A,S,T,X M 39 N
M 3
E
" 6) Feedwater/RilR Line B 10 .
1,2,3 $
1821-F0328 B, L 0.5 Yes 9.0 c3 lC 1E12-F0538 A, S, T, X. AA A- C3 mco
g 3
m -< p i - un '
t:
. t, e
l
,q
s. . . '
TA,BLE 3.6.4-1 (Continued) ~
CONTAINMENT AND DRYWELL ISOLATION VALVES E -
MAXIMUM . SECONDARY
c. VALVE APPLICABLE
PENETRATION ISOLATION ISOLATION CONTAINMENT TEST NUMBER OPERATIONAL TIME 4 NUMBER SIGNAtt CONDITIONS BYPASS PATil PRESSURE *
(Seconds) _(YES/NO) (psig) l.
% utomatic Isolation Valves (Continued) 6.
Primary Containment (Continued) ,i
30) Instrument A'ir Supply 57 '
IIA 005 1; 2, 3 Yes U 9.0 I IIA 006 S 0
. 5 i , !$1) Instrument Air Bottles i i';* IIA 0128 58 L, 8 I!
. 1, 2, 3 Yes 9.0 [i l 'i' IIA 012A X 14 t'
14
'!y32) Servic'e Air Supply 59 t
, ISA030 , 1,2,3 Yes B. L '1 9.0 1SA029 5 B , L.
5 '
33) RWCU Suction Line 60 L+
IG33-F001 1, 2, 3 No
. B,F,N,1,2,E,X 9. 0 1G33-F004 15 B,F,N,1,2,E,X 15
' 34) RWCU Return to Filter 61 '
IG33-F053 1, 2, 3 No
! IG33-F054 B, F. H 1, 2, ,,X, 9. 0 - - ~
15 B,F,N,1,2, ,X '
15 u '
-c ;
35) tysrogen Recombiner Supply 62 C;3 4 1,2,3
{ GHD08 B, L '
30
.Yes i
9.0 avs j l
, 36) RWCUToRNR/Ef'
.,..n p j IG33-F040 64 ki 1, 2, 3 No 9.0 IG33-F039 B,F,N,1,2,E,X 15 B,F.N,1,2 E,X m !.j!
ij37) RWCU Transfer To Radwaste
'IWX019 65 5
1, 2, 3, & #
15 g '
j
,,' B, L Yes 9.0 ca IWX020 'i 2 O B, L '
l 1
2 M q o '
~ .._ .. . . . . . .- ...
f t
TABLE 3.6.4-1 (Continued) l l1 H if CONTAINMENT AND DRYWELL ISOLATION VALVES o
q l.
z i HAXIMUM SECONDARY APPLICABLE ISOLATION CONTAINHENT TEST il g VALVE PENETRATION ISOLATION OPERATIONAL TIME BYPASS PATH PRESSURE il y NUMBER NUMBER SIGNAlt CONDITIONS (Seconds) (YES/NO) (psig)
Automatic Isolation Valves (Continued)
Primary Containment (Continued) -
1
b. Drywell 4
i ! 1) Plant Chilled Water Supply 53 1, 2, 3 NA No NA m IWO551A L', U 3 g IWO551B L, U ,
JJ) Plant Chilled Water Return 53 1,2,3 NA No NA j j m IWO552A , L, U
(,
l IWO552B -
L, U 6
3) Drywell HVAC Supply 101 A '
l 1,2,3 No NA IVQ001A L, B, H, Z, 5 IVQ001B L, 8, M, Z, 5 J*
4) Drywell HVAC Exhaust 102 1,2,3 No NA IVQ002 L, B, M, Z, 5 6
IVQ005 L B, H, Z, 5 6 - - - - - - -
! i IVQ003 L. B, M, Z, 5 6 c l ! --J .
2. Manual Isolation Valves C3

c2
,. . -et
a. Primary, Containment 1. . .., go '
' ' ~
1) RHR/LPCI A Injection 15 4 !
At All Times (a)
NA g-
{ NA No 9. 0 ,i
,- IE12-F044A -
l
c f* RHR/LPCI B Injection 16 - NA At All Times (a) 9.0 Q
l j. 2) 1E12-F0448
~
NA No Q ;!
i i 9, "
u .... ,! -< I I
i t, .
. I i !
t
i p TABLE 3.6.4-1 (Continued) {
t:! s S CONTAINMENT AND ORYWELL ISOLATION VALVES
. =
HAXIMUM SECONDARY E VALVE APPLICABLE ISOLATION' CONTAINMENT 1EST PENETRATION ISOLATION !
q NUMBER NUMBER OPERATIONAL TIME BYPASS PATil IRESSURd SIGNAtt CONDITIONS (Seconds) (YES/N0) {psig)
Test Connections, Vents and Drains (Continued)
, Primary Containment (continued) ~
26) llead Spray l 42 NA 1E51-F034 NtAllTimes(a) NA No 9. 0 ,
i 1E51-F035 IE51-F390 1E51-F391 ,
1E12-F061 R
+
1E12-F062 1
? 27) RCIC Turb Steam Supply 43 NA t IE51-F399 e At All Times (a) NA No 9. 0
, IE51-F072 IE51-F401
28) RCIC Turb Vacuum Breaker 44 NA 4
1E51-F080 At All Times (a) NA No 9. 0 i 1E51-F082 i 1E51-F345-
~
IESI-F375 K~s ;
! 1E51-F376 $$
1E51-F083 c:s
29) Main Stream, Drain Line 45 NA i- Qo 1821-F017 At All Times (a) NA ' Noi 9' 9. 0 :::c2
30) CCW Supply 46 NA k
ICC164 At All Times (a) NA 9.0 !
g 10C266 No P Yes ss to E
i I
w p-= eo
a..".Al 'm 14-\ ^
t 4 2 PROOF & REVIEW COPY j CONTAINMENT SYSTEMS 3/4.6.7 ATMOSPHERE CONTROL TNIS PAGE OPEN PENDING REC CONTAINMENT HYDROGEN RECOMBINER SYSTEMS
,.[i INFORMATION FROM TdE APPUCA
$l ' LIMITING CONDITION FOR OPERATION I
i 3.6.7.1 Two independent containment hydrogen recombiner systems shall be
'NI OPERABLE.
x~
$ APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
45 ACTION: ,
d. With one containment cd/r hy fhydrogen recombiner system inoperable, d restore the inoperable system to OPERABLE status within 30 days or be in at X.1 least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
or -
.ny
'.% SURVEILLANCE REQUIREMENTS
.h
.T 4.6.7.1 Each containment hydrogen recombiner system shall be demonstrated Pir OPERABLE:
.$ - a.
c"]_
At least once per 6 months by verifying during a recombiner system func-
g -
tional test that the heater sheath temperature increases to greater than or equal to 600*F within 60 minutes.
6;: ;
'f.j
b. At least once per 18 months by: -
i?
~ ~
'l 1. Performing a CHANNEL CALIBRATION of all recombiner operating instru-cl mentation and control circuits.
il -
J . 2. Verifying through a visual examination that t'here'is no evidence of
abnormal conditions within the recombiner enclosure; i.e, loose
.j wiring or structural connections, deposits nf fo' reign material.s, etc.
.I,h. 3. Verify 1ng the integrity of all heater electrical circuits by
~.; \a performing a resistance to ground test following the above required 5 \ functional test. The resistance to ground for any heater phase i}.e .
g shall be greater than or equal to 10,000 ohms.
4 4. Verifying during a recombiner system functip,nal $ast that the reac-tion chamber temperatu increase to pe > 6 150)"F within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and
} is maintained between 177fF and 6223}2F for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
~
. l l
=
b l
CLINTON - UNIT 1 3/4 6-66
AUG 2 9 IS85 i
. - . . =- - __ _-- . _ - _ _
n.
.' [
.. m . . -
.~. T h D ' A
W M h W M A M i F '~* % N ' W M #.
. , , +: _
v+
.i l
l 4.kf
,hhh t i
CONTAINMENT SYSTEMS i
CONTAINMENT AND ORWELL HYDROGEN IGNITION SYSTEM
.4 i g LIMITING CONDITION FOR OPERATION '!
& )
J
' .y'i w 3.6.TMF~c5n' tainment and drywell hydrogen ignition system consisting of:
j, a. TWi Tndependent containment and drywell hydrogen ignition iIu'bsystems each G consisting of six circuits as listed in Table 3.6.7.3-1 with no more than ,
$ two igniter assembifes inoperable per circuit and no more than five l q igniter assemblies inoperable per subsystem, and 2 !
'S b. At least two igniter assemblies in each enclosed area specified.in '
.;c 5
Table 3.6.7.3-2 shall be OPERABLE. I Q APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2 0
.9 .
ACTION:
5.3 ...
. . . a _ With one containment and drywell hydrogen ignition subsystem inoperable,_-
Tj restore the inoperable subsystem to OPERA 8LE status within 30 days or be W
.3 in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .
3 %- .
)
ij N- b. With less than two igniter assemblies OPERA 8LE in any enclosed area specified in Table 3.6.7.3-1, restore at least two igniter assemblies in 1
2. ., --
each enclosed area to OPERABLE status within 30 days or be in at least l HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ' t h! . . i SURVEILLANCE REQUIREMENTS -
/.1 1 jd;a' - 4.6.7.3 The containment and drywell hydrogen ignition system shall be id demonstrated OPERABLE:
1 55 ' - .
l
~t .a. At least once i
. r ays by energizing the supply Dreakers and performing i ig ...
a current surement of n:h circuit. . .
f;i . . b. At less onca per 18 months by energizing the supply breakers and ve'rify-d~ ~
3 i ng___
urrent measurement sufficient current draw to develop 1700*F tem-paraT e for those igniter assembifes in high, radiation areas and verifying
)"j a sur ace temperature of at least 1700*F for each of the remaining igniters.
7
5 %
,~
D g ,~n R'tWS CSS h \ A -
l: NN N ' '
(
CLINTON - UNIT 1 3/4 6-69 AM.13 E i l 1
l
_ _ . _ .s.___._. . _ _ _ _ .
i
~
., . .g S i a 5%: ou -i:
' Guba
PR00F & REREW COPY Table 3.6.7.3-1 Hydrogen Igniter Circuits Division I l I
CfrduitT Circuit 2 Circuit 3 Circuit 4 Circuit 5' ~ Circuit 6
-] IHG12EN - 1HG03EB 1HG07EB 1HG06ED 1HG07EM . 1HG08EA
IHGHEB 1HG03ED 1HG12EA IHG06EE 1HG09ED "' 1HG08EC l i 1EJ11ED 1HG03EF 1HG12EC IHG06EG 1HG09EF 1HG08EE l
{ 1HGHEF 1HGOSEB 1HG12EE 1HG06EJ 1HG09EH 1HG08EF l
! IHGHEH ING05EC 1HG12EG IHG06EL 1HG09EN 1HG08EG ;
.i 1HGUEK ING05EF 1HG13EB 1HG07ED 1HG09EM IHG08EJ i 1HGUEM 1HG13ED 1HG07EF 1HG09EQ 1HG10EE l
l 1HG12EJ 1HG13EF 1HG07Ett 1HG10EA 1HG10EG 1 1HG12EL IHG13EH 1HG07EK 1HG10EC 1HG10EJ
1HG13EK 1HG09EB .....
1HG10EK i 1HG13EM 1HG13EP .
l
) ,
Division II ,
i Circuit 1 Circuit 2 Circuit 3 ' Circuit 4 Circuit 5 Circuit 6 lr -- - -
1HG06EA 1HG06EC 1HG03EC 1HG07EN 1HG08EB 1HG06EH-1HGHEA 1HG06EH ING06EB
j (' 1HG06EM 1HG03EE 1HG08ED 1HG11EC 1HG11EA 1HG11EC 1HG06EK 1HG08EH i 1HG06EF 1HG05ED 1HG09EA 1HG11EE 1HG11EE ING08EK" i 1HG07EA 1HG05EE .. 1HG09EC 1HG11EG 1HGHEG 1HG08EM j 1HG07EC IHGOSEG 1HG09EE 1HG11EJ- 1HGUEJ 1HG08EN
.) 1HG07EE 1HG09EG 1HGUEL 1HG11EL 1HG10EB 2- 1HG07EG 1HG09EJ 1HGHEN 1HGHEN 1HG10ED j~ 1HG07EJ 1HG09EL 1HG12EH 1HG12EH 1HG10EF si .. 1HG07EL 1HG09EP 1HG12EK 1HG12EK 1HG10EH
1 1HG12EM IHG12EM 1HG10EL
/
. 1HG10EM I -
i
z. .
1 Pg d
1, .
'}
1 0
f ,
D.
4 6
CLINTON - UNIT 1 3/4 6-70 AUG 2 s 1985
i em . w. ~. v + 1 sumamsessewah '
y . .'
1 PR00F & REVIEW COPY A
jf PLANT SYSTEMS
}h SURVEILLANCE REQUIREMENTS (Continued) - -
lb M 4. Verifying that each fire protection pump starts sequentially to gm , . .- --
maintain the fire protection water system pressure greater than or equal to 65 psig.
.1
d. ~At least once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.
M . .
{ 4.7.6.1.2 Each diesel driven fire protection pump shall be demonstrated -
,N OPERA 8LE: _ . t
(;.i 4
3, a. At least once per 31 days by: -
4to
$f
1.- Verifying that the fuel day tank contains at least allons of -
. fuel.
S
' $w
2. Starting the pump from ambient conditions and operating for greater than or equal to 30 minutes on recirculation flow.
i $
,y
% b. At least once per 92 days by verifying that a sample of diesel fuel from'
@g the fuel storage tank, obtained.in accordance with ASTM-D270-75, is within
4 the acceptable limits specified in Table 1 of ASTM D975-77 when checked j.h for viscosity, water and sediment.
%g,' c. At least once per 18 months, during shutdown, by subjecting the diesel to M te an inspection in accordance with procedures prepared in conjunction with h
n its manufacturer's recommendations for the class of service, S~ 4.7.6.1.3. Each diesel dri wn fire pump starting 24-volt battery bank and
~
9 y, charger shall be demonstrated OPERABLE:
I
a. At least once per 7 days by verifying that:
(n , ,
, a
-) '1. The electrolyta level of each pilot cell is above the plates',
[j 2. The pilot cell specific gravity, corrected to 77'F and full 4 electrolyte level, is greater than o,r. equal to. 1.250 , and.
k E 3. The overall battery voltage is greater than or equal'to 24' volts. .,
$ 4
$ b. At least once per 92 days by verifying that the specific gravity is
-, appropriate for continued service of the battery.
M c. At least once per 18 months by verifying that:
l@
?. 1. The batteries,' cell plates and battery racks show no visual indica- -
f) tion of physical damage or abnormal deterioration, and i n$ 2. Battery-to-battery and~ terminal connections are clean, tight, free e E, of corrosion and coated with anti-corrosion material.
e, 4h '
i 4 i
g CLINTON - UNIT 1 3/4 7-17 g gg g
. 1 A ... _
. -- y , ;-. r ,= _ x..s - - - - - - - - ---.-
. . . , -* t** _
q . s .- - ,,
. D - - _
f-'?ik0fb~mg & N CQfY TABLE 3.7.6.5-1 (Continued)
.. , FIRE HOSE STATIONS
. . __ y . .
!. . Z._I ; . i i .=-.. .____
- -- ~nMEWHOSE" RACK (1)
6 s1 . . LOCATION AN0' ELEVATION 4.l;._ . . ., m . t yOfNTIFICATION t- ..r ze %;w . . .__ . , . r- .m .
3=~ =-q p 7
' ng;3 (qMControl:BuildingcElevation 781' - .
.
E7y '-i ii..
. ..2.,
- $ $g f%w.e _ . . *.
,m 1..
4-Q-. -
. - ~ Y*Stai'rwell' sodthwest corner-r . . --- b ' '=. 'AA' ' ' ~125 c
. _s - 2. Outside Div.;.3; Battery Room . _. __.J. : 7. Z: PZ . . . AA . .130,
3. -Div.jlCableSp' reading! Area 11. ' 9 ; .Y ~128
4. Passage outside Div.~1 Inverter Room ' -~
Y - 130
5. Outside Div. 4' Inverter Room . .. 'V - 124
6. Pass' age outside Div."2 Inverter Room- V - 130
~
7. Div. 2 Cable Spreading Area T - 128
8. North side of wall V - 130
9. Northeast quadrant T - 133
10. East side of wall Y - 130
11. East from battery charger AA - 130
- 12. South side. - Y - 133
13. Southeast' corner AA - 135
]4. East side new door V __202 r._. Control Buildino, Elevation 800' -
- 1. Northwest quadrant T - 124
2. Southwest corner AA - 124
3. Southwest quadrant between doors AC - 128
4. Outside shift sup, office . AC - 130 I 5. Outside planning and sched. room AC - 130
6. Southeast quadrant AC - 133
7. Southeast corner , AA - 135
.-. s. . Control Building, Elevation 825'
1. West side on wall AA - 130
'. _ 2. Near 480 Volt substation K Y - 125
' ., i' " .fE 3.
'~
Southwest quadrant near. door . .m-- .
AA 124
4. c Southwest quadrant .
AA ~133 l~25. ~ West: side of wall *
.= 4 - G-' -
V - 135 l
6. ~ Southeast corner _ .
AA ..
135 I
~ ~ ~ ~
t. 4 Control Buildinc, Elevation 847' _ . . _ . _ ZJ. _ -
. w . . ...n.--..-
-w,- - - .
. --..= .....=.. .
C=="* --
I.= South side of wall - - - - - - - - - - -
'AA - 124 L j
2. South side of wall near door A C-133 L m ;
u. Screen House Elevation 699'
1. On west side of missile wall 3-2
2. Between plant service water pumps B-6 ..
3. Diesel fire water pump room B -11 em CLINTON - UNIT 1 3/4 7-27 AUG 9 loo 5
-. -,,.-,, , . - . . - . , , , . , . .- ,_ - . , . ~ , . . . , - , . , - - , n ,, -.-----,-
a ,f:.: :. .n.;::. e a...%,,.. 2yy ., x- w.p.A%.w ;.~ . -a, ,7., 4, , - --- -.._ .,- --fyg;, .. g , ;,
. ,... - .. - 1..
43 . ,
. u. ; ,
i. ,
I I , !
-i A ,
TABLE 3.7.6.6-1 l 4 !
A YARD FIRE HYDRANTS AND' ASSOCIATED HYDRANT HOSE HOUSES S., .
i; --
e H0SE HOUSE -
@. . , HYDRANT NUMBER NUMBER LOCATION
,a 9
M OFP 112 29 "r9 + 23 N 3
g 5 + 38 E -
? 0FP 113 30 ~ 8 + 04 N h.; i .
8 + 33 E
.D, OFP 128 4 -
4 + 20 S l
':1 e 8 + 95 E_
T.8 .
a
- 25 0FP 131 5 1 + 24 S
% 8 P+ 93 E *
&a : - OFP 132 . 23 0 + 79 N -
55! -
4 + 73 E OFP 133 24 0 + 48.5 N ii) 2 + 77 E
^
91 ..
+3 ~ .. . . . .
'JJ *
' 0FP 134 25 1 + 56 N .
~] -
/ 0 + 05 E 2 '
. .i 0FP 135 10 5 +.05 N T4 8 + 94 E a
i, -
i;41 ,
OFP 136 8 2 + 28 N 6 ' 8TE i.4, . -
}j -
0FP 168 27 7 + 46 N 0 + 05 E
]i~3'
k . - .
OFP 169 26 4 + 66'N d' -
! 0 + 05 E-a g 0FP 171 28 9 + 25 N 4..
1 + 74 E
.e/
V-2
.#,J
.c .
. :n ~
.x Q ,.
l}
?f
,s_1, .
. CLINTON - UNIT 1 3/4 7-30 AUG 2 91985
,,y-+. , , - - - - - - - - , - - , ., -, - - , , _ _ , - - .n._,,.n .-,,,-. ,, , ,
.w.m :
~
~ s k s #G & & Listk
.a .- + .-
hy w - -
a el d
1 s PM0F & REY!EW COPY ELECTRICAL POWER SYSTEMS
-}
LIMITING CONDITION FOR OPERATION (Continued) -
N[.]
.i. ACTION (Confinued) fl c. With one offsite circuit of the above-required A.C. sources and diesel h ~
. generator 1A or 18 of the above required A.C. electrical power sources -
p -
inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and a@l at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If a diesel generator became E inoperable due to any cause other than preplanned preventive mainte-74 3 nance or testing, demonstrate the OPERABILITY of the remaining OPERA-y '
BLE diesel generators, separately, by performing Surveillance Require-l ments 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months *. Restore at i j least one of the inoperable A.C. sources to OPERABLE status within
(
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and 1 2 in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore at least two .
l p.3 offsite circuits and diesel generators 1A and 1B to OPERABLE status l J.7 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from time of initial loss or be in at least HOT SHUT-76 DOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in COLD SHUTDOWN within the follow- ;
}*.} ing 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
d. With diesel generator 1C of the above required A.C. electrical power Nl sources inoperable, demonstrate the OPERASILITY of the offsite A.C. -
() sources by performing Surveillance Requirement 4.8.1.1.1.a within
..j 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter. If the diesel gen-erator became inoperable as a result of any cause other than preplanned
@y preventive maintenance or testing, demonstrate the OPERABILITY of the W remaining OPERABLE diesel generators, separately, by perfo.rming Sur-Q veillance Requirements 4.8.1.1.2.a.4 $nd 4.8.1.1.2.a.5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months *.
3: Restore diesel generator 1C to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or Q declare the HPCS system inoperable and take the ACTION required by Specifications 3.5.1 and 3.7.1.1.
J C.i
..] e. With diesel generator 1A or 18 of the above required A.C. electrical'e
.H 9.3 power sources inoperable, in addition to taking ACTION b or c, as appifcable, verify within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months that all required systems, subsys-g-
) tems, trains, components and devices that depend on the remaining P.L OPERABLE diesel generator as a source of emergency power are also I OPERA 8LE, and that the appropriate shutdown service water (SX) pump d, is OPERABLE if diesel generator IB is inoperable; otherwise, be in at fj least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and'in COLD SHUTDOWN g within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . -
g h f. With both of the above required offsite circuits inoperable, demon-V strate the OPERABILITY of three diesel generators, separately, by 7 performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5
$j within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months unless the diesel generators are already operating.
Restore at least one of the above required offsite circuits to OPERABLE
$e '
status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT SHUTDOWN within the next .
12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . With only one offsite circuit restored to OPERABLE status, Oi ..
'I *This test is required to be completed regardless of when the inoperable diesel "
3 generator is restored to OPERABILITY.
n .
'i CLINTON - UNIT 1 3/4 8-2 AUG 29 1E FN y.:
4 9 .. . . . , . -. .. . . . . . +
. . . _ - _ m , _ . , . - - _ _ _ _ _ . . _
.__..______._,__-.,_ .___-__----._.__~__...._---__-_-_-_..-,..
, . 3 ;; . Nub ! W hii & & iSf95.89&#.'.'$ANSdf9&E$,'f}N:-'.$...'2_lQb .. - .
PROOF & REVH COPY j.
U.4n i
'd '
ELECTRICAL POWER SYSTEMS
'N h{ SURVEILLANCE REQUIREMENTS 2
Q 4.8.1.1.1 Each of the above required independent circuits between the offsite
E transmission network and the onsite Class 1E distribution system shall be:
A
' d~ -
al' Determined OPERA 8LE at least once per 7 days by verifying correct breaker j alignments and indicated power availability, and
~M ' b. . Demonstrated OPERA 8LE at least once per 18 months during shutdown by
'hp ,
transferring, manually and automatically, unit power supply from the r
normal circuit to the alternate circuit.
2 4.8.1.1.2 Each of the above required diesel g.nerators e shall be demonstrated 92 OPERA 8LE:
,d
.M
a. In accordance with the frequency specified in Table 4.8.1.1.2-1 on a
!p STAGGERED TEST. BASIS by: .
h
1. Verifying the fuel level in the day fuel tank.
2. Verifying the fuel level in the fuel storage tank.
4 h('
D,;
3. Verifying the fuel transfer pump starts and transfers fuel from the storage system to the day fuel tank.
4 4. Verifying the diesel starts
from ambient condition and accelerates
@q to at least 870 rps in less than or equal to 10 seconds. The genera-tor IA and IB voltage and frequency shall. be 4160 t 420, - O volts and
d. 60 + 6,~ - 0 Hz and the generator 1C voltage and frequericy shall be E;i 4160 i 420 and 60 2 1.2 Hz within 10 seconds after the start signal.
The diesel generator shall be started for this test by using one of g$ - the following signals:
3 a) Manual.
'l.j b) Simulated loss of offsite power by itself.
y c) Simulated loss of offsite power in conjunction with an ESF
A actuation test signal.
j d) AnECCSactuationtagsygnalbyitself. .
g ge 5. Verifying the diese enerator is synchronized, loaded o greater d than or equal t for diesel generator IA for diesel 5 generator 18 an 2200 kW for diesel generator 1C in less than or d equal to 90 seconds,* and operates with this load for at least
60 minutes.
M 6. . Verifying the diesel generator is aligned to provide standby power 4 >[j to the associated emergency busses. .
W s
h "All diesel generator starts for the purpose of this surveillance test may be d] preceded by an engine prelube period. Further, all surveillance tests, with- ~
the exception of once per 184 days, may also be preceded by warmup procedures' M
and may also include gradual loading as recommended by the manufacturer so .
d. -- that the mechanical stress and wear on the diesel engine is minimized. .
!b .
'A
'n
'as
.i CLINTON - UNIT 1 3/4 0-4 gg?,3 g 1
4 .

._,,.---,.,,.--,.,.-y,--rm.m.~.-w-,--,,__

.,_.,_,.-__%.,.ym-,,..___..,-_
-_...-.---.---......--,-.,-r. . . . . . - - . . . . . . . -
> W :k h h$.t;&s & hpHj'R$.h k5l=4;%5im.:f.
3:3.. _
A.
,y) '
j IsAtJ V i O i i d N N I h .-
N1 ELECTRICAL POWER SYSTEMS y'C }
y SURVEILLANCE REQUIREMENTS (Continued) .THIS PAGE OPEN PENDING r RECElPT O IM..en. ....,,v D..AA A,Titu,,t1 cmmo m 4 m, , u ,A,r r ,
nnemauT
_n m r
y
7. Verifying the pressure in all diesel generator air start receivers to W be greater than or equal to 200 psig. ~ ' - ~ ~ -~
@. b. At least once per 31 days and after each operation of the'dielel'where" - "
.@ the period of operation was greater than or equal to I hour'by' checking ' " " '
g for and rooving accumulated water from the day fuel tanks. - -
r dj c. At least once per 92 days by removing accumulated water. from the fuel
. storage tanks.
A*
u .
p d. At least once per 92 days and from new fuel oil prior to addition to the h
g storage tanks, by obtaining a sample in accordance with ASTM-D270-1975, and by verifying that the sample meets the following minimum requirements y and is tested within the specified time limits: .
M
^g 1. As soon as sample is taken from new fuel or prior to addition to
3 the storage tank, as applicable, verify in accordance with the. tests sf specified in ASTM-0975-77 that the sample has

i:a y a) A water and sediment conthnt of less than or equal to 0.05 -
W volume percent.
D 3.3 b) A kinematic viscosity @ 40*C of greater than or equal to h9 1.9 centistokes, but less than or equal to 4.1 centistokes. -
w h .(p $ (c) A specific gravity as specified by the manufacturer @~ 60/60*F l g
l Yf of greater than or equal to but less than or equal to R E.} or- an API gravity @ 60*F of greater than or equal to
$j . degrees but less than or equal to degrees.)
3 ([lF h! d) An impurity level of less than 2 mg of insolubles per 100 m1.
- (2) when tested in accordance with ASTM-D2274-70; analysis shall be R] completed within 7 days after obtaining the sample but may be j sampled and analyzed after the addition of new fuel oil.
N (3,) 2. Within two weeks after obtaining the sample, verify that the other
$ properties specified in Table 1 of ASTM-0975-77 and Regulatory d^g Guide 1.137 Position 2.a. are met when tested in accordance'with a
ASTM-0975-77..
e. At least once per 18 months,# during shutdown, by:
h' 1. Subjecting the diesel to an inspection in accordance with procedures e prepared in conjunction with its manufacturer's recommendations for l
$.:a this class of standby service.
(
~
fj #For any start of a diesel, the diesel must be operated with a load in accord-p, ance with the manufacturer's recommendations. ~(
p -
{ CLINTON - UNIT 1 -
3/4 8-5 ma AUG 2 91985
, N
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n x z
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. ~ w. . :: ,. :
a 3 .
1
f

P!!DOF & REM COPY 8
ELECTRICAL POWER SYSTEMS y SURVEILLANCE REQUIREMENTS (Continued) a j *
. b. At least once per 92 days and within 7 days after a battery discharge with
$,7*f- battery terminal voltage below 110-volts, or battery overcharge with
..g battery teminal voltage above 150-volts, by verifying that: -- ' - "
- fj
[ 1. The parameters in Table 4.8.2.1-1 meet the Category 8 limits,'
~
] ; - ]'
' TN 2. There is no visible corrosion at either terminals or connectors, " - -
t or the connection resistance of these items is less than
.$b 150 x 10.e ohns, and
- IF
3. The average electrolyte temperature the pilot cells of connected
'1 cells-is above 65*F.
%,2t ,
!vi ,
c. At least once per 18 months by verifying that:
i.W g -
1. The cells, cell plates and battery racks show no visual indication h
M.
of physical damage or abnormal deterioration,
.M 2. The call-to-cell'and terminal connections are clean, tight, free of d corrosion and coated with anti _-corrosion material, n -
@ 3. The resistance of each cell-to-cell and terminal connection is less
I'$ than or equal to 150 x 10 s ohms, and v
"i 4. '- The battery charger will supply at least'300 amperes for Divisions I ps and II and 100 amperes for Division III at a minimum. of 10,5 volts
.j for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . -
y d. At least once per 18 months, during shutdown, by verifying that either:
21
1. The battery capacity is adequate to supply and maintain in OPERABLE
@j
,e status all of the actual emergency loads for the-design duty cycle
. when the battery is subjected to a battery service test, or
' .3 iO
! .ii 2.. The battery capacity is adequate to supply a dummy load of the follow-l {d "l ing profile while maintaining the battery terminal volt, age greater _
l than or equal to 105 volts.
a) Ofvision I '
.Al/ j
$ > 549 amperes for the first 60 seconds l
$ I 227 amperes for the next 59 minutes
{
.ft i147amperesforthenext180 minutes st b) Division 2 IM 'l
> 404 amperes for the first 60 seconds ,
a > 274 amperes for the next 59 minutes q > 86 amperes for the next 180 minutes 1 -
?.
, f4 CLINTON - UNIT 1 3/4 8-13
! Aus z s isas 4
i e.I
_ __ _ r _ - . - - ^ - -
. " -- . ,m.
. .g . . ... _,4_. - . . e .sy.y .
_ .M r =_r = q a
- o q
PH00F& HEV!EW COPY
~!
4 TABLE 4.8.2.1-1
.,4
+ -
BATTERYSURVEILLANCEREQUIRENFE
.A.
.a ~
h CATEGORY A(1) CATEGORY B(2)
?$
$W . Parameter Limits for each Limits for each A11owable(3) designated pilot connected cell value for each cell connected cell
.@k
i. Electrolyte > Minimum level > Minimum level Above top of y Level indication mark, indication mark, plates, w -
and < h" above and <-%" above and~not maximum level maximum level overflowing
{Q."gt indication mark indication mark i
Float Voltage l'2.13 volts 1 2.13 volts (c) > 2.07 volts
~
h .Specifi{,) 1 1.200(U) g 1.195 Not more than 16 Gravity .020 below the h # average of all
> 1.19 connected cells k
.' ' Average of all Average of all y connected cells connected cells g 1.205 (D)
} 1 1.195
.:e
>3. (a) Corrected for electrolyte temperature and level'.
4 (b) Or battery charging current is less than (2) ' amperes when on float charge.
ff ~
(c) May be corrected for average electrolyte temperature.
-(1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months all the Category 8
' ;i d measurements are taken and found to be within their allowable values, and
).j provided all Category A and 8 parameter (s) are restor.ed to within limits a within the next 6 days.
Q (2) For any Category 8 parameter (s) outside the limit (s) shown, the battery
$ may be considered OPERABLE provided that the Category 8 parameters are 6 within their allowable values and provided the Category 8 parameter (s)
[j are restored to within limits within 7 days.
9 (3) Any Category B parameter not within its allowable value indicates an f inoperable battery.
/s) u Q .
l 5 l& .
is
'~
I
M. - . .'
..'s4
d CLINTON - UNIT 1 3/4 8-15 a
a AUG 2 91985
.L- - _ . _ . . _ . _ . _ . _ . . . . - . - ^
- _ - . - - - - - - . - _ ___ _ - . - -_ -__~-_ ~-
f
. _ _ ,. , ny. '=
wk-..:6 -%d%% % 4. cam %=% W !$$6
^
PROOF & REV!EW COPY ELECTRICAL POWER SYSTEMS .
3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS DISTRIBUTION - OPERATING
! LIMITING CORDITION FOR OPERATION 3.8.3.1 The following power distribution system divisions shall be energized:
l l,, _ ,a. A.C., power distribution:
~ -~
j 1.~~Dfvision I, consisting of:
l; a) 4160 volt AA/ Bus 1A1 1 b) 480 volt Unit Substations A and'1A i c) 480 volt AC MCC's- l
', 1) Aux. Bldg. MCC's 1A1, IA2, IA3, IA4 l i 2) SSW MCC 1A 1
3) DG MCC 1A afrel l 4
i / 4) 9* -' n . Idg. MCC's El, E2, and G
, / d) 0 volt A.C. di ribution panels in 480 volt Auxiliary Building .
/ MCC 1A1 and Control Building MCC E2.
' - I e,) 120 volt AC uninterruptible distribution panels energized from I k 001A, which is fed from 480V Auxiliary Building MCC,1Al e d.
SV DC MCC 10C13E.
i \ .5') Domp<c Mcc A MI %Was Mcc E2 and frwm
, 2. Division II, consisting of:
!7 a) 4160 volt A[C[ Bus 181'
]' '
b) 480 volt Unit Substations B and IB _.
i c) 480 volt AC MCC's
' Aux. Bldg. MCC's 181, 182, 183, 184
~
'1) i 2) SSW MCC 18 -
1 j 3) DG MCC 18. M/*I -
J- 4) N"'---- Idg. MCC's F1, F2, and H
} d) 120 volt AIC/di ribution panels in 480 volt Auxiliary Building
~
2 MCC 181 and Control Building MCC F2.
l e) .120 volt AC uninterruptible distribution panels energized from
j. 1C71-S0018, which is fed from OV Auxiliary Building MCC 181 %eQ
! tomg,12>41ctcc1oc145- w u u.s a cc o .,a s.,
j 3. Division III, consisting.of: -
l a) 4160 volt 4[C[BusICI b) 480 volt A/C/ Aux. Bldg. MCC IC and IC1 and SSW MCC IC
} c) 120 volt Altf distribution panels in 480 velt Aux. Bldg.'MCC IC
.i and IC1. -
d) 120 volt AE, uninterruptible distribution panels energized from IC71-500 W, wh is f f 480V Auxiliary Building MCC IC l C, g and 125V D g
d D's4eib.4.6 '
-; 4. Reactor Protection System (RPS) 120V AC Tdlenoid Buses A an B from j their associated inverters. - .-
b. D. C. power distribution
1. Division I, consisting of 125 volt D[C[ Battery 1A,125 volt Battery
.' Charger 1A,125 volt 07C/MCC 1A and Of stribution Panel.
i .
CLINTON - UNIT 1 3/4 8-17 l AUG 2 91%S
- - - - _ - - ~ - - _ - - - , , - , - - - - , - , - - - , - - _ _ _ _ - - -
.. .u. - -:. . . . . . . . . . . . . .. -a -~:-- --~- - " " ~ ~ "
PRDOF & Rt1EW COPY ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION'(Continued)
2. Division II, consisting f125voltD!C Battery 18, 125 volt Battery Charger 18, 125 volt D/C NCC 18, and Distribution Panel. l 3 N ivision III, consistin f 125 volt C[8atteryIC,125voltBattery Charger IC, 125 volt O " '"n istribution Panel.
APPLICA8ILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
i
a. For A.C. power distribution: *
1. With either Division I or Division II of the above required AM distribution system not energized, re-energize the division within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and
in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . s .
2. With Division III of the above required A distribution system not energized, declare the HPCS system inoperable and take the ACTION
! required by Specification 3.5.1.
l
3. For inoperable RPS Solenoid Bus inverters:
r -
a. With an RPS Solenoid Bus inverter inoperable transfer the bus tc
( . . - . the alternate power source provided the other RPS Solenoid Bus --
is not being supplied from the alternate source.
i b. With both RPS Solenoid Bus invert.ers inoperable'de-en~ergize one RPS Solenoid Bus.
c. Withthefrequencyofthe120VA[CfsupplytotheRPSSolenoid buses A or B < 57 Hz, demonstrate the OPERABILITY of all equip-ment which could have been subjected to the abnormal frequency for all class 1E loads connected to the associated buses, by performance of a CHANNEL FUNCTIONAL TEST or CHANNEL CALIBRATION, as required, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
b. For 0.C. power distribution:
1. With either Division I or Division II of the above required DIC[
distribution system not energized, re energize the division within i 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and
. in COLD SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
, l 2. With Division III of the above required 0/C/ distribution system i not energized, declare the HPCS system inoperable and take the ACTION j required by Specification 3.5.1. ,
\
4 o
CLINTON - UNIT 1 3/4 8-18 M 2 91985 1
1
NY' g., . , sQ.% ~'i.M_UMblA$ W hkbN S U
"" ~
4 1 PRDOF & REVIEW COPY j
j ELECTRICAL POWER SYSTEMS
.i 4 DISTRIBUTION - SHUTOOWN
?
4 LIMITING CONDITION FOR OPERATION 3
y 3.8.3.2 As a minimum, the following power distribution system divisions shall j
- be energized: ,
y a. For A.C. power distribution, Division I or Division II,~ and when the HPCS
,j system is required to be OPERABLE, Division III, with:
t 1. Division I consig, igg of: -
3,J a) 4160 volt A/C/ Bus 1A1.
'o b) 480 volt Unit Substations A and 1A.
p.4 c) 480 volta 7ChtCC's d 1) Aux. Bldg. MCC's IA1, 1A2,~1A3, 1A4.
g 2j gsW g my c,ofy i h 4) on Idg. MCC's El, E2, and di .
d 120 A.C. distribution panels in 480 volt Auxiliary Building
?gl MCC 1A1 and Control Building MCC E2. .
f I e) 120 volt AC uninterruptible distribution panels energized.from q IC71-S001A, which is fed from 480V Auxiliary Building MCC 1A1 n rougk
2. iv s on ng a) 4160 volta)C/' Bus 181.
6]p',
{; - b) 480 volt Unit Substations.B and 18.
!$ c) 480 volt A/C/MCC's
% 1) Aux. Bldg. MCC's 181, 182, 183,'184 h -) SSW MCC IB fg
gl -
M 3) D MCC IB 4'! m 4) ont . MCC's F1, F2, and H
, M d) 7120 vo A.C. distribution panels in 480 volt Auxiliary Building i.?. and Control Building MCC F2.
[1 e) 120 volt AC uninterruptible distribution panels energized from
jj 1 0018, which is fed from 480V Auxiliary Building MCC IBl%rog M- 5 Cedral Odlj
MCC F2 el h 3.
59 c.Nc MC Divisi Mg10C14E.
II conhisting of:
Ed di a) 4160 volt A/C.78us qgo
.k) 480 volt A/C/,AB MCCIC1.
IC anb)d Aux.yAc Tgmf.,e,es-Bldg. MCC IC1, and SSW MCC 1C.
%;;p g .r) 120 volt A/C(distribution panels in 480 volt Aux. Bldg. '
MCC IC and Aux. Bldg. MCC IC1.
i M
i.s g #) 120 volt AC uninterruptible distribution panels energized from IC71-S001A, wh is fed 480V Auxiliary Building MCC IC and 125V O 22- -*
j -5eoic
~/j BMcWufien Panel
't .
b. For D.C. power distribution, Division I or Division II, and when the l ', .
HPCS system is required to be OPERABLE, Division III, with:
! .T '
l ih 1. Olvision I consisting ofJ25 volt 0.C. Batteries IA,125 volt Battery .
S Charger 1A,125 volt 0/Cf MCC-IA, and Ofstribution Panel.
'M l .f; 2.
Division Charger 18, II consisting 125 volt 0 /t/ MCC-1B, and Distribution Panel.of,125 volt 0.C.
l t f CLINTON - UNIT 1 3/4 8-20 l AUG 2 91985 l j, .
l-_--- '
d3%N.% DIE 4MNNIt'$.S$$EdM@$@'dd$$1'R - 'M ELECTRICAL POWER SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) i Division III consisting of 125 volt
_ 3.--Charger IC,125 volt D.C. *rl", .4C. BatteriesPanel.
IC,125 valt Battery j -- - - - a Distribution
~
APPLICA8ILITY: OPERATIGNAL. CONDITIONS, 4 5 and 8
-l ACTION: -
?
a. For A.C. power distribution:
1. With both Division I and Division II of the above required A.C. dis-tribution system not energized, suspend CORE ALTERATIONS, handling of irradiated fuel in the secondary containment and operations with a potential for draining the reactor vessel.
~
, 2. With Division III of the above required A.C. distribution system not energized, declare the HPCS system inoperable and take the ACTION I required by Specifications 3.5.2 and 3.5.3. '
, b. For D.C. power distribution:
?
1. With both Division I and Div.ision II of the above required D.C. dis-r tribution system not energized, suspend CORE ALTERATIONS, handling of Jrradiated fuel in the Auxiliary Building and Enclosure Building and operations with a potential for draining the reactor vessel.
2. With Division III of the above requir,ed D C. distribution system not energized, declare the HPCS system inoperable and take the ACTION f required by Specifications 3.5.2 and 3.5.3.
j c. The provisions of Specification 3.0.3 are not applicable.
1 i .
4 SURVEILLANCE REQUIREMENTS
'l l- 4.8.3.2 At least the above required power distribution system divisions shall
? -
De determined energized at least once per 7 days by verifying correct . breaker ~
alignment and voltaga en the busses /MCCs/ panels. *
?
i
When handling irradiated fuel in the secondary containment.
1 -
i CLINTON - UNIT 1 3/4 8-21 g .4 3 g 4
$. .. q.:. . : J.? , , ' -T,n
_.. .7. - <._. :. . .
Q . _
, ---...~.
hff N TABLE 3'.8.4.1-1
) , (continued)
$ Auxiliary Building MCC 1A2 (1AP73E)
.j j
. Location 121, V (R C); EL 781 FT.
f! .jj.. Each Compartment listed below has two (2) identical -
circuit breakers in series 4
'21
,- CIR BKR PENETRATION EQUIPMENT CABLE PENETRATION SYSTEMS D4 COMPT TRIP CABLE SIZE SERVICE rd NUMBER NUHBER AFFECTED i .1 IB 15 #6 RHR Valve Resid. Ht.
.f,~'j IE12-F037A 1RH63A 1EE09E Removal
'];;1 E; 13C 15 #6 IIA 0128 - IA02A 1EE09E Instrument Air Y.i
$5 .
?fi
$2
..n u
A 7.03 .
' ~ '
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El
.4 3
91 .
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.'6 T,
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7
. CLINTON - UNIT 1 3/4 8-30 AUG 2 9145 I
b
~ .s m m&wsam+&w w -
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! PRDOF& REVIEW CDPY l '
. .i TABLE 3'.8.4.1-1
..;)
i (continued) 1 g Auxiliary Butiding MCC 182 (IAP76E)
.' . Location 106, V (R,C); EL 781 FT S . Each Compartment listed below has two (2) identical
j circuit breakers in series. .
. ,g..
CIR BKR PENETRATION EQUIPMENT CA8LE PENETRATION SYSTEMS g COMPT TRIP CABLE SIZE SERVICE NUMBER NUMBER AFFECTED M ~. ~~ 11C _ 15 #6 Isol. Valve Standby 9 IC41-F001B 15C06A 1EE10E Liq. Cont, i:]t y 28 15 #6 Inlet Valve ICC08A 1EE10E Component Q. :.
ICC068 Cool Water 18 15 #6 Inlet Valve Component k 1CC065 1CC080 1EE10E Cool Water
$ 2C 15 #6 Outlet Valve Component 7i ICC070 1CC09A 1EE10E Cool Water
.1 a
3 2A 15 #6 Outlet Valve Component 11 1CC067 1CC090 1EE10E Cool Water
$1 ij . 10C 15 #6 Sup Pool Viv Resid. Ht.
,j - - - -
-1E12-F073A . .1RH42A .1EE11E Removal -
a G 118 15 #6 Isol Valve -Shutdown
..M ISX0958 ~1SX57A 1EE11E Serv. Water
.?-
3 -
10A 15 #6 Suct. Valve fj - 1HG0098 1HG06A 1EE11E H2 Recomb.
q:,
~
11A 15 #6 Sup. Pool Viv Resid. Ht.
.0 1E12-F0738 1RH43A 1EE11E Removal R .
(
. 14R 15 #6 Spray Valve Resid. Ht.
g 1E12-F0288 1RH62A 1EE11E Removal c.3
a 108 15 #6 Upper Pool Resid Ht.
!.'J Univ.
.j IE12-F0378 1RH64A 1EE11E Removal 21
ij 61 4
.J d
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-i AUG 2 3 ixi
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f PROOF &. REVIEW TABLE 3.8.4.1-1 '
(continued) -l Li M Auxiliary Building MCC 183 (IAP77E)
Iy . Location 106, V (R.C); EL 781 FT m
Each Compartment listed below has two (2) identical
}5 circuit breakers in series. .
6 CIR BKR PENETRATION EQUIPMENT CABLE PENETRATION SYSTEMS .
COMPT TRIP CABLE SIZE SERVICE NUMBER AFFECTED l 3] ,, !iUMBER ..
$4 2A 15 #6 Isol. Valve Component y ICC050 ICC12A 1EE10E Cool Water q
Ai 28 15 #6 Isol. Valve Component S 1CC053 1CC120 1EE10E Cool Water 77
v. 38 15 #6 Isol. Valve Component 1CC071 1CC13A 1EE10E Cool Water hd 3C 15 #6 Isol. Valve Component ICC074 1CC130 1EE10E Coot Water M
..q
$ 3A 15 #6 Isol. Valve Component 4 1CC060 ICC160 1EE10E Cool Water p
ff 4A 15 #6 Isol. Valve ' Component 0- 1CC127 1CC16L 1EE10E Cool Water 3
$ 4C 15 #6 Isol. Valve .
-Cycled
,d ICY 017- '1CYO6A 1EE10E condensate SA 15 #6 Isol. Valve Cycled g ICYO20, ICYO6F 1EE10E Condensate m
d SB 15 #6 Isol. Valve Fuel Pool C! IFC007 1FC05A 1EE10E Cooling .
hi g SC 15 #6 Isol. Valve Fuel Pool j 1FC037 1FC20A' 1EE10E' Cooling 1
h 10A 15 #6 Isol. Valve . Reac.
$ 1E51-F063 1RIO2A 1EE11E Inject.
M d 14A 15 #6 RCIC Valve Reac.
71 1E51-F076 1RI15A 1EE11E Inject.
?i 108 '15 #6 Isol. Valve Reac. Water h 1G33-F001 1RI15A 1EE11E Cleanup k =-
1!
4 ,
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s
'y CLINTON - UNIT 1 3/4 6-36 .gg n
61 v . - . _. - . , , . - . . .
L
.mc s : mr m . 4cm 3.;e.m d:3 PROOF.& RtTEW COPY '
1 d TABLE 3~.8.4.1-1 S.} (continued)
?
..a Auxiliary Building MCC 1H (1AP95E)
. . Location 119, I (R, C); Ei. 762 Ft v}..
El Each Compartment listed below has two (2) identical fd circuit breakers in series.
Z
,g, CIR BKR PENETRATION EQUIPMENT CABLE PENETRATION SYSTEMS g COMPT TRIP CABLE SIZE SERVICE NUMBER ' NUMBER AFFECTED - -
k 70 80 350 MCM Welding h IEWO2E ,
l1EWOIA 1EE03E Welding
@M
~2C 15 #6 Supp. Pool -
M Fill Valve Supp. Pool i-M, 1SM004 ISM 05A 1EE05E Make-up
..C 15 #6 RWCU Reac. Waste 9l 1WX01PA IWXO6A 1EE05E Cleanup n
^I
! 2A 15 #6 RWCU c'$ 1G33-F107 1RT33A 1EEOSE Hoists A
E4 3A 15 #6 RWCU Reac. Water
!$ 1G36-C001A 1RT43A 1EE05E Cleanup M
F[j 78 30 #2 Monorafl .
lfg 1821-E300 1HC13E 1EE05E Hoists .
~
. 7F 15 #6 Hatch Shield-y Door 1HC68G 1HC65A 1EE07E Hoists
.- SA 20 #6 Circuit 7 1FH06Y 1EE07E - Fuel Handling
, ', IF42-E001 68 15 #6 Refuel Plat . Fuel
. '. 1F15-E003 1FH11E 1EE05E Handling c '
1, '
4A 15 #6 Air Hand Fan Chilled
~N
IW605SF 1W625G 1EE07E Water
-)? 48 40 #6 Air Hand Fan Chilled 1W605SH 1WO99A 1EE07E Water
-?.d?
fi] 40 15 #6 Air Hand Fan Chilled F-j 1W605SM 1W625U 1EE07E Water
?.i
,1 4C 15 #6 Air Hand Fan Chilled 3 1W605SK IW627A 1EE07E Water n! '1
.e
^?!
U
'; CLINTON - UNIT 1 3/4 8-40 AUG 2 91985 i
_, .. ... -_ _v.- -e-.=_ ..---- - -
- , - , - - - - , - . ,, - - - - , - - -n-m,.,,,.,-n,. .m. 7,, , ,-,,w - - - - - - - - , - , , w- .,,.------r-- - ,,-+ -
~
dw;._ - ;a. . __
u-
. . - ; n.
if -
3 M008h N E i is
] TABLE 3.8.4.1-1 (continued)
'd
] Auxiliary Building MCC 1H (IAP95E) (Continued)
CIR'BKR PENETRATION EQUIPMENT CABLE PENETRATION SYSTEMS b
COMPT TRIP CABLE SIZE SERVICE NUMBER NUMBER AFFECTED
30 40 #2 Air Hand Fan Chilled
,j 1Wil05SB IW[05A 1EEOSE Water h 6A 100 350 MCM Oil Pump Reac. -- -
M 1833-0003A 1RR19A 1EE36E Recire.
53 2B 100 350 MCM Mixing Htr. Standby Q IC41-0003 - ISC03A 1EE36E Liq. Control ja
3 38 30 350 MCM Tnk Htr. Standby ij
.n IC41-D002 ISC04A 1EE36E Liq. Control ,
M1 7A 15 #6 Fan Mtr.
7J
1833-0003A 1RR21A 1EE36E Reac. Recire.
@g 7C 15 #6 Area Coolers 1WO34C 1EE07E Area Coolers 1 IWO340 1EE07E Area Coolers
(( 1WO34E 1EE07E Area Coolers
{j . 1WO34F 1EE07E Area Cool.ers f; . . _ . .
-1WO34P 1EE07E - Aiea~ Coolers 9.?
1WO34Q 1EE07E Area Coolers di IWO34R 1EE07E' Area Coolers h -
IWO345 1EE07E. -Area Coolers
.3 -
jj 18 15 #6 IVP090A 1VP37A 1EE05E - Chilled Water 1.s IC 15 #6 1VP091A IVP38A 1EE05E Chilled Water s
,[ 1D 20 #6 1F15-E003EC 1FH13C '1EE07E Fuel Handling 7D 20 #6 1F15-E003EA 1FH13A 1Et07E Fuel Handling 2.1 Q
V4 ,
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.bi CLINTON - UNIT 1 3/4 8-41
% AUG 2 91985 di
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PRDOF &.RfI!EW COPY TABLE 3'.8.4.1-1 0
} CONTAINMENT PENETRATION CONDUCTOR
-; OVERCURRENT PROTECTIVE DEVICES
) .
@ DEVIC E SYSTEM (S)
,,Q , AND LOCATION AFFECTED
.,gs
,aa -
9 2. Type Switchgear
.pl
'M Polar Crane - Penetration IEE03E 2-350MCM per i
.p. ,
i.. Unit Substation 1Al Compt. 78 l[f sf (R C); EL 781 FT
. ?! Primary Protection
.S B8E Solid State Trip Device Type S514 -
@s' 3
"6
,- Current Sensor 600A . -
L.T. Setting 1.1 X TAP ST Setting 10 X TAP
.i
$}4..
}r, Secondary Protection
. ., 5 .g
'Mris Westinghouse Type CO-8 Relay
.9 ly$ _
ce:
bR *
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_:.9
.2..
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-e
.y ;
.' 21 y
E sb
i -
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1 3;3 w
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{ (" List all primary and backup breakers.) -
I CLINTON - UNIT 1 3/4 8-44 AUG 2 91985 t
L l
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PRODF & REWEW COPY
_{ TABLE 3.8.4.2-1 f MOTOR GDERATED VALVES THERMdL OVERLOAD PROTECTION i
jf Valve No. . Bray Direction System (s) Affected d
1821-F016 Continuous close Q
G 1821-F019 Continuous Close Nuclear Boiler Nuclear Boiler S 1821-F065A Continuous Open/Close Nuclear Boiler A 1821-F065B Continuous Open/Close Nuclear Boiler M 1821-F067A Continuous Close Nuclear Boiler T 1821-F067B Continuous Close - Nuclear Boiler
'E 1821-F067C Continuous Close Nuclear Boiler 2 1821-F067D Continuous Close Nuclear Boiler E 1821-F068 Continuous Close Nuclear Boiler Y 1821-FC98A Continuous Close Nuclear Boiler
.M 1821-F0988 Continuous Close Nuclear Boiler 1B21-F098C Continuous Close Nuclear Boiler
@% 1821-F0980 . Continuous Close Nuclear Boiler .
p 1CC049 Continuous Close Component Cool Water j ICC050 Continuous Close Component Cool. Water q ICC053 Continuous Close Component Cool Water Qi ICC054 Continuous Close Component Cool Water 1CC057 Continuous Close Component Cool Water -
?j
,i- ICC060 Continuous Close Component Cool Water 4 1CC065 Continuous - Close Component Cool Water
? '
.CC068 Continuous Close Component Cool Water 1 .
.. CC071 Cont-inuous Open/Close Component Cool Water 1 1CC072 Continuous Open Component Cool Water li ICC073 Continuous Open . Component' Cool Water
?;1 1CC074 Continuous Open/Close Component Cool Water
'M 1 Continuous Close Component Cool Water gge075B Continuous Close, Component Cool Water h]Vi d
- lauz6A ICC076B ICC127 Continuous Continuous Continuous Close Close Close Component Cool Water Component Cool Water Component Cool Water fh ICC128 Continuous Close -Component Cool Water r,m 1;;;,,
~ . _ . . Cm. ==,c=,=.y_
C=M =x 01 . C;...p . .. . . :. 0 - '. k' 6 1CY016 Continuous Close Cycled Condensate 1 1CY017 Continuous Close Cycled Conde'nsate 5 '1CY020 Continuous Close Cycled Condensate n
{ .1CY021 Continuous Close Cycled Condensate O IC41-F001A Continuous Open Standby Liquid Control 23, IC41-F0018 Continuous Open Standby Liquid Control "i
ss y
Con %uoM
# CompomVCool %d
~
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..~ -.. _
..._.,.m,._
- , , . - - ..n-. , . . , - - , . ,-.,--.n,_,-_.-- -----,-,n..-,- . . . - - - - . . - , , - - - . - - - - . . - - - - - - - -
s- "
j .
TABLE 3.8.4.2-1(Continued) d, MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION i
{l Valve No. ,
Bypass Direction System (s) Affected
)n, IE12-F003A IE12-F0038 Continuous Continuous Open Open Residual Heat Removal Residual Heat Removal fj 1E12-F004A Continuous Open/Close Residual Heat Removal a 1E12-F0048 Continuous Open/Close Residual Heat Removal Q 1E12-F006A Continuous Close Residual Heat Removal 4 1E12-F006B Continuous Close Residual Heat Removal t,1 1E12-F008 Continuous Open/Close Residual Heat Removal y 1E12-F009 Continuous Close Residual Heat Removal 23 1E12-F011A Continuous Close Residual Heat Removal 1 1E12-F011B Continuous Close Residual Heat Removal
'ij 1E12-F014A Continuous Open/Close - Residual Heat Reinoval n 1E12-F014B Continuous Open/Close Residual Heat Removal l q IE12-F021 Continuous Close . Residual Heat Removal D .
1E12-F023 Continuous Op'en/Close Residual Heat Removal 1E12-F024A Continuous Open/Close Residual Heat Removal f
9] 1E12-F024B Continuous Open/Close Residual Heat Removal
% 1E12-F026A Continuous Close Residual Heat Removal di 1E12-F0268 Continuous Close Residual Heat Removal 7.} 1E12-F027A Continuous Open/Close Residual Heat Removal 1E12-F0278 Continuous Open/Close Residual Heat Removal
.] 1E12-F028A Continuous Open/Close Residual Heat Removal N 1E12-F028B Continuous Open/Close Residual Heat Removal i 1E12-F037A Continuous Open/Close Residual Heat Removal 1E12-F0378 Continuous Open/Close Residual Heat Removal
? 1E12-F040 Continuous Close Residual Heat Removal li 1E12-F042A Continuous Open/Close . . Residual Heat'Rsmoval M 1E12-F0428 Continuous Open/Close Residual Heat Removal 1E12-F042C Continuous Open/Close Residual Heat Removal
]h IE12-F047A Continuous Open Residual Heat Removal
.} 1E12-F047B Continuous Open Residual Heat Removal U 1E12-F048A Continuous Open/Close Residual Heat Removal 3 .1E12-F0488 Continuous Open/Close Residual Heat Removal
l. 1E12-F049 Continuous Close Residual Heat Removal
-9
, 1E12-F052A Continuous Close Residual Heat Removal 4^'
1E12-F0528 Continuous Close Residual Heat Removal 1E12-F053A Continuous Open/Close Residual Heat Removal n 1E12-F0538 Continuous Open/Close Residual Heat Removal 1 1E12-F064A Continuous Open/Close Residual Heat Removal Ej 1E12-F0648 p Continuous Open/Close Residual Heat Removal 1E12-F064FC Continuous Open/Close Residual Heat Removal 9] 1E12-F068A Continuous Open Residual Heat Removal
,j 1E12-F0688 Continuous Open Residual- Heat Removal 94 IE12-F073A Continuous Open/Close Residual Heat Removal j 1E12-F0738 Continuous Open/Close Residual Heat Removal Y
y 5 .
a y - -
f.1
.9 CLINTON - UNIT 1 3/4 8-47 L:
AUG 2 91985
. ' . ~ s [ ,
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P200F & REVIEW COPY ,
,? i 4 TABLE 3.8.4.2-1(Continued) v.t _
[1 MOTOR OPERATED VALVES THERMAL OVERLOAD PROTECTION '
h . Valve No. Bypass Direction a ,
hstem(s)Affected M 1E51-F059 Continuous Open/Close Reac. Core Isol. Cool 3 1E51-F063 Continuous - Open/Close y- ~
1E51-F064 -
1E51-F068 Continuous Continuous Open/Close Open/Close Reac. Core Isol. Cool Reac. Core Isol. Cool
%q Reac. Core Isol. Cool '
1E51-F076 Continuous Open/Close Reac. Core Isol. Cool 3
1E51-F077 Continuous Open/Close Reac. Core Isol. Cool 21 1E51-F078 Continuous O Reac. Core Isol. Cool
? DE58-F095 Continu*es C> pen /Closenn/ % Class
c. Core 1.sel. Cool
., IFC007 Continuous Close Fuel Pool Cool & Clean
". 1FC008 Continuous Close Fuel Pool Cool &-Clean 1FC011A Continuous Open/Close Fuel Pool Cool & Clean 1FC011B Continuous Open/Close Fuel Pool Cool & Clean 1FC015A ' Continuous Open/Close Fuel Pool Cool & Clean 0 1FC015B Continuous Open/Close Fuel Pool Cool & Clean 4 1FC016A Continuous Close Fuel Pool Cool & Clean 1FC0168~ Continuous Close Fuel Pool Cool & Clean 1FC024A Continuous Close Fuel Pool Cool & Clean 1FC0248 Continuous Close Fuel P:o1 Cool & Clean
>i 1FC026A Continuous Open/Close Fuel Pool Cool & Clean n 1FC026B Continuous Open/Close Fuel Pool Cool & Clean
% 1FC036 Continuous Close Fuel Pool Cool & Clean la;t IFC037 Continuous Close Fuel Pool Cool & Clean- -- -
k 1FP050 Continuous Close Fire Protection N IFP051 Continuous Close Fire Protection -
5 1FP052 Continuous Close
' Fire Protection d 1FP053 Continuous Close Fire Protection
d. 1FP054 Continuous Close Fire Protection M 1FP078 Continuous Close Fire Protection d! *1FP079 Continuous Close Fire Protection ilo 1FP092 Continuous close Fire'Pfotection k 1G33-F001 1G33-F004 Continuous Close React.'Wtr. Clean Up
$2 Continuous Close React. Wtr. Clean Up 7 1G33-F028 Continuous Close React. Wtr. Clean Up 3 1G33-F034 Continuous Close React. Wtr. Clean Up d 1G33-F039 Continuous Close React. Wtr. Clean Up s 1G33-F040 Continuous Close Rea-t. Wtr. Clean.Up
1G33-F053 Continuous Close React. Wtr. Clean Up
!$ 1G33-F054 Continuous Close React. Wtr. Clean Up ,
4-
% 1HG001 Continuous Open H2 Recombining 9 1HG004- Continuous Open/Close H2 Recombining N 1HG005 Continuous Open/Close H2 Recombining M 1HG068 Continuous Open/Close H2 Recombining
f. .
.3 3 -- -
3m 9
4 :
CLINTON - UNIT 1 3/4 8-49 "3 AUG 2 91985~
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i h
i ELECTRICAL POWER SYSTEMS
3/4.8.4.5 REDUNDANT FAULT PROTECTION FOR PGCC FIRE PROTECTION, COMMUNICATION,-
j RPS AND MSIV CIRCUITS ilh LIMITING CORDITION FOR OPERATION 3.8.4.5 All over-current devices shown in' Table 3.8.4.5-1 shall be OPERA 8LE.
~
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3. - - - - - -
ACTION:
, a. With one or more of the above required conductor overcurrent devices shown in Table 3.8.4.5-1 and/or fuses tested pursuant to Specification 4.8.4.5.a.2 j inoperable:
1. Restore the protective device (s) to OPERABLE status or deenergize the circuit (s) by tripping, racking out, or removing inoperable device '
1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and I -
j
2. Verify at least once per 7 days thereafter the inoperable device is
.; tripped, racked out, or removed.
3 b. The provisions of Specification 3.0.4 are not applicable to overcurrent 4 devices which have the inoperable device racked out or'~r esioved or, which j (~ ..
have the alternate device tripped, racked out,.or removed.
a .. - -
@ SURVEILLANCE REQUIREMENTS .
'.4 d 4.8.4.5 Each over-current protective device shown in Table .4.5-1 shall be .
]
demonstrated OPERABLE: J -
a. At least once per 18 months.
) 1. By selecting and functionally testing a representative sample of at 9 least 10% of each type of circuit breakers. Circuit breakers. selected
[ for functional testing shall be selected on a rotating basis. Testing
of these circuit breakers shall follow manufacturer's instructions and shall test the long time, and instantaneous elements for pickup and
[J time delay, where applicable. Circuit breakers found inoperable j during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable
)? . during these functional tests, art additional representative sample of f at least 10% of all the circuit breakers of the inoperable type shall l also be functionally tested bntil no more failures are found or all
,; circuit breakers of that type have been functionally tested.
M .
~'.
]
.!I
f. CLINTON - UNIT 1 3/4 8-59
-] AUG 2 91985
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PH00F & REVIEW COPY ' -
3 REFUELING OPERATIONS
Y}i REFUELING PLATFORM
'k,. LIMITING CONDITION FOR OPERATION ~ '
M 3.9.6.2 The refueling platform shall be OPERA 8LE and used for handling fuel
=
assemblies or control rods within the reactor pressure vessel.
W .
it APPLICABILITY: During handling of fuel assemblies or control r6(s within the
{4 reactor pressure vessel.
ACTION:
$ With the requirements for refueling platform OPERABILITY not satisfied, suspend i
E use of any inoperable refueling platform equipinent from operations involving E the handling of control rods and fuel assemblies within the reactor pressure
& vessel after placing the load in a safe condition.
M. '
SURVEILLANCE REQUIREMENTS 31 b 4.9.6.2 Each refueling platform crane or hoist used for handling of control rods 74 or fuel assemblies within the reactor pressure vessel shall be demonstrated 7:i OPERABLE within 7 days prior to the start of such operations with that crane
.E .
or hoist by:
y .
E a.
E Demonstrating operation of the overload cutoff on the main hoist when the load exceeds 1200 2 50 pounds. ~
f.
9 b. Demonstrating operation of the overload cutoff 'on the frame mounted and 8 monorail hoists when the load exceeds 500
50 pounds.
$' t f c. Demonstrating operation of the uptravel mechanical stop on the frame d1 mou and monorai1~ hoists when 'uptravel brings the top of the grapple
'y to eet below the p TWorm: r:eUs.
ct "-' " ' '
i
d. Demonstrating operation of the downtravel mechanical ' cutoff on the main si hoist when grapple hook down travel reaches 2-4 inches below fuel assembly g
handle.
..a
'j e. Demonstrating operation of the slack cable cutoff on the main hoist when y the load is less than 50 2 10 pounds.
3 f. Demonstrating operation of thie loaded interlock on the main hoist when 9 the load exceeds 485 2 50 pounds. -
7
'O g. Demonstrating operation of the redundant loaded interlock on the main g hoist when the load exceeds 550
50 pounds. ,
3 h. Demonstrating operation of 'the main hoist raise power cutoff when the ..
%. .. refueling platfom area radiation monitor dose rate exceeds 10 mR/hr.
g .
A W
N e
T CLINTON - UNIT 1 3/4 9-10 g^ AUG !91985
p .
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pn00F & RBs 00PY REFi!ELING OPERATIONS j
! AUX 1LlARY PLATFORM LIMITING CONDITION FOR OPERATION
=
3.9.6.3 The auxil.tary platform shall be OPERA 8LE. '
APPLICA8ILITY: During handling of" control rods with the auxiliary platform.
ACTION:
'dith the requirements for auxiliary platform OPERABILITY'not satisfied, suspend use of the auxiliary platform after placing the load in a safe
' condition. -
SURVEILLANCE REQUIREMENTS 1
3 4.9.6.3 The auxiliary platform hoist shall be demonstrated OPERABLE within 7 -
g_ days prior to the handling of control rods by:
] a. Demonstratin'g operation of the overload cutoff when the load j- exceeds 500 pounds. .
)~.
9
b. Demonstrating operation of the ' auxiliary platform ggist uptravel-
. stops when the grapple is lower than or equal top feet below g
the platform rails.
p -
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! i CLINTON - UNIT 1. 3/4 9-11 1 AUG 2 91985
't
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a .
REFilELING OPERATIONS PRDOF & REVH COPY 2
'j MuG1PLE CONTROL ROD REM 0WA .
d LIMITING CONDITION FOR OPERATION b ^~~ F
% N 3.9.10.2
'f Any number of control rods and/or control rod drive mechanisms may
<; be removed from the core and/or reactor pressure vessel provided that at least the following requirements are satisfied until all control rods and control
, g rod drive mechanisms are reinstalled and all control rods are inserted in the g core.
g g a.
The reactor mode switch is OPERA 8LE and locked in the Shutdown position
fr. or in the Refuel position per Specification 3.9.1, except that the Refuel d position "one-rod-out" interlock may be bypassed, as required, for those i$ control rods nnd/or control rod drive mechanisms to be removed, after the bl fuel assemblias have been removed as specified below.
d, *
%j b. .
N -
The source range monitors (SRM) are OPERA 8LE per Specification 3.9.2.
$, c. The SHUTDOWN MARGIN requirements of Specification 3.1.1 are satisfied.
Si
$ d.
All other control rods are either inserted or have the surrounding four
$13 .
fuel assemblies removed from the core cell.
J
,M -
e.
The four fuel assemblies surrounding each control rod or control rod drive
}$
d mechanism to be removed from the core and/or reactor vessel are removed y f...fromthecorecell.
"~ '
/
ih ---- > APPLICABILITY: OPERATIONAL CONGTION 5.
~
. 21 .
ACTION:
. yj. With the requirements of the above specification not satisfied, suspend removal M
~d of control roc.s and/or control rod drive mechanisms from the. core and/or reactor pressure vessel and initiate action to satisfy the above requirements.
.it . .
y# -
m h CSS / M Mc ~ S k "I'* ~' "'*'1'P a -
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'b W .~.,.
v. , .
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s
.J) CLINTON - UNIT 1 3/4 9-17 AUG 2 91985
,-------m.,-, -
-,,me+, w,.,-,...w, , , - . . - ---.------,-,--e ---r-- - - - -- -, -- -r ,- - - , , - . _%,
_,-r,.--,.u,- , - -
. . ; , 2. . .;c. .G il dIdilHiMinN2fi.G#"'S . :
1 '* 4AGElsifNSfd215EilkMSEddESP2dMMTYDEEWA$#'#MM(g;. '
i .,
3 p,-
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I TABLE 3.12-1 NE P .
t, RADIOLOGICAL ENVIRONMENTAL HONITORING PROGRAM !y-2 .
Number of 7 I
E Representative , y 4 Exposure Pathway Samples and Sampling and Type and Frequency />
! s and/or Sample Sample Locations, Collection Frequency of Analysis ',
.i 1. JIRECT RADIATION b
40 routina monitoring stations Quarterly Gamma dose quarterly. >
. ;j DR1-DR4Deither with two or lt more costmeters or with one j instrument for measuring and k' recording dose rate continuously, i placed as follows: j
i (1) an inner ring of stations, j
, one in each meteorological
{ 4:' sector in the general area of
i the SITE BOUNDARYJ
!* M ii' J. (2) an outer ring.of stations, one *
! in each meteorological sector in jl the 6 to 8 km range f, rom the site; i
^i i
(3) the balance of the stations I .to be placed in
l special interest areas such as '
j, population centers, nearby resi- .
i dences, sch'ols, o and in'1 or 2 I c:3 .;
j areas to serve as control sta- N
[
j! , tions. 98 '
! ::x:s -,
't i<
til ct E Eo i. I.
J $ .
M (J
i $ * (-
i#5 *
. E l
- .. (..:. .u W...:k:. A&uhR:51IEdM?.?bi1OG .a. ?LC2GT42n$%IMAGEWiiMEMNEWIBhtWY" "WhbMMRE; 7.o M
{ TABLE 3.12-1 (Continued) k j
RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM h P Number of '
i Representative
% Exposure Pathway Samples and Sampling and Type and Frequency i
,l-o y and/or Sample Sample Locations a Collection Frequency of Analysis y@
jl g 2. AIR 8ORNE l
!i 4 Radiciodine and Samples from 5 locations: Continuous sampler Radiciodine Canni' ster:
e Particulates operation with sample I-131 analysis weekly. -
l. a. -3 samples " ' from close collection weekly, or b..:
to the 3, SITE BOUNDARY loca- more frequently if i* tions in different sectors of required by dust Particulate Sampler: ..
i the highest calculated annual loading. Gross beta radioactivity C
) average ground-level D/Q. analysis folloging -
t
- filter change; 1 b. I sample ^ from the vicin- Gamma isotopic analysis' lty of a community having the of composite (by l highes calculated annual aver- location) quarterly.
j l , w age ground-level D/Q.
! c. I sample from a control . ,
~
! M location, as for example 1
.A 15-30 km distant and in the .i
! least prevalent wind direction.c l 3. WATERBORNE .-
i a. Surface f I sample upstream Compositesampfeover Gamma isotopic analysis" i 1 sample downstream 1-month period monthly. Composite for '
tritium analysis quarterly.
b. Ground Samples from 1 or 2 sources Quarterly Gamma isotopic
and tritis --
J.
)
)
affected,.
,onlp if likely to be analysis quarterly. g g
l c. Drinking 1 sample of each of 1 to 3 Composite sample over 2-week period8 I-131 analysis on each composite when the dose S ([
{ e of the nearest water I, i ; supplies that could be a when 1-131 analysis . calculated for the consump- # ,
'affected by its discharge, is performed, monthly tion of the water is greate N .
composite otherwise than 1 ares per year.I Con k i
5 ee 1 sample from a control location, positeforgrossbetaaHd gamma isotopic analyses E lii-in i
monthly. Composite for tritium analysis quarterly.
Q {f ;
g ,
y si F 4
6.
e.y- a n.m msaw ww ~- -
~
e .
5:e PROOF & RSH COPY h INSTDUMENTATION M.1 5
$ BASES y :
~
. E .- . . REACTOR PROTECTION SYSTEM INSTRUMENTATION (Continued) -
r -
!ik pe '.# .. .
ThE system meets the intent of IEEE 279 for nuclear ~ power plant 'ppotection sys-tems. The bases for the trip settings of the RPS are discussed in the bases foi Specification 2.2-1.
y '
The measurement of response time at the specified frequencies provides assurance
F that the protective functions associated with each channel are completed within C the time limit assumed in the accident analysis No credit was taken for those
u channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or 1
y$' total channel test measurement, provided such tests demonstrate the total -
N -
channel response time as defined. Sensor response time verification may be
$. demonstrated by either.(1) inplace, onsite or offsite test measurements, or
}j (2; utilizing replacement sensors with certified response times.
7 Because the trip logic of the solid state reactor protection system results in p$ a trip of all four divisions and full reactor scram if the logic is satisfied IG - for the coincident logic reactor trip functions or the non-coincident NMS reac-
? tar trip function, the REACTOR PROTECTION SYSTEM RESPONSE TIME tests of the
, 8 various reactor trip functions can only be performed during shutdown. All four G divisional logic response times are therefore checked for every response test of M two RPS channels of each function. Each function has four logic trains through E, two out of four coincident logic circuits located one in each division. There 54 are four coincident logic circuits associated with each reactor trip function each of which will cause the trip of one RPS division logic.
h.- j
~
5.f 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION I
g b .st/AILl)A l d* Thisspecificationensurestheeffectivenessoftheinstrumentationdsedto Cj' mitigate the consequences of accidents by prescribing the OPERABILITY requirement PJ . trip setpoints and response times for isolation of the reactor systems. When hd necessary, one channel may be inoperable for brief intervals to conduct required ,
P surveillance. Some of the trip settings may have tolerances explicitly stated M where both the high and low values are critical and may have a substantial l
F,f effect on safety. The setpoints of other instrumentation,' where only .the high M
0 or low and of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of a
h WI the systems involved.
(j The 2-out-of-4 logic for the MSIV isolation functions is identical to the logic y of the RPS and the instrumentation response time is demonstrated,jn an identi-fij cal manner.
R '
f Except for the MSIVs, the safety analysis do'es not address individual sensor l@ . response times or the response times of the logic systems to which the sensors .
733 are connected. (For 0.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves,'it is assumed that the A.C.
n W5 CLINTON - UNIT 1 B 3/4 3-2 AUG 2 9 2
e. m w mm m.w ey. w n ~;e w w.s easess 3
I JR00F &RMM COPY l f i
i CONTAINMENT SYSTEMS b
ff BASES 8-
4.
b '
$ - --. d4.6;2.'3 ORWELL AIR LOCKSC
~ -
', , sip
. uq,;.; .
p& , . --. . .The limitations on closure for the drywell air locks are required to meet the R--1 4
-. ' restrictions-on DRWELL INTEGRITY and the drywell leakage rate given in Speci- .
fications.3.6.2.1:and 3.6.2.2.- The specification makes allowances for the-fact -
3.y. - that there'may be long periods of time when the air locks will be in a closed and secured position during reactor operation. Only one closed door in each
,l.
air lock is required to maintain the integrity drywell.
_ 3/4.6.2.4 ORWELL STRUCTURAL INTEGRITY
.~/:
5 'This limitation ensures that the structural integrity of the drywell will be .
2 _ , . maintained comparable to the original design specification for the life of the ~ -
9 unit. A visual inspection in conjunction.with Type A leakage. tests is suffi-y ~ cient to^ demonstrate -this capability. e , 4
. - S.w .Q,- >
. :. t . .u --
a. . ..; , _
c.
.i; 3/4.6.2'.5' 'DRWEl( INTERNAL' PRESSURE
~
,a U. The, limitations on drywell-to-containment differential pressure ensure that
' is - the drywell peak pressure of 18.9 psig does not exceed the design pressure of- -
p.} 30.0 psig and that the containment peak pressure of 9.0 psig does not exceed
f. the-design pressure of ~15.0 psig during steam-line break conditions.~ 'The maxi-E*; mum external drywell pressure differential is limited to 0.1 psid, well h below the (2.3) psid at which suppression pool . water will be forced "over the 91 wier wall and into the drywell. The limit of 1.0 psid for initial positive 7~
~
drywell to containment pressure will limit the drywell pressure to 18.9 psid
. which is less'than'the design pressure and is consistent with the safety analysis to limit drywell internal pressure.
f.*.
m . .
ij~ 3/4.6.2.6 DRWELL AVERAGE AIR TEMPERATURE
.J ..
, 11 . The limitation on' drywell average air temperature ensures that peak drywell Wc temperature does not. exceed the design . temperature of 330*F during LOCA condi-E tions and is consistent with the safety analysis.
YT 3/4.6.2.7 DRWELL VENT AND PURGE -
~
b 3L
<d 3r
. The.drywell. purge system mustbe normally maintained closed to eliminate a potential challenge to containment structural integrity due to a steam bypass
%~ of the suppression pool Intermittent venting of the drywell is allowed for y pressure control during OPERATIONAL CONDITIONS 1, 2 and 3, but the cumulative
- time of venting is limited to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months per fuel cycle. Venting of the drywell fj is prohibited when the 12-inch continuous containment purge systee or the 3 ~
36-inch containment building ventilation system supply or exhaust valves are .
M open. This eliminates any resultant direct leakage path from the drywell to =-
j the environment. ,,
I!- In OPERATIONAL CONDITIONS 1, 2 and 3, the drywell purge 24-inch exhaust valve
a can be opened only if it is blocked so as not to open more than 50*. This 4 assures that the valve would be' able to close against cr" PM. ressure .
buildup resulting from a LOCA. depe,d CLINTON - UNIT 1 B 3/4 6-5 .
MG 2 S 25
~
.y"nm;.fus. , .q, } } y;,yg ,;.. . :-
, , ; . i .:
~- . f.X Q * . -
.. E E. ...
- - - - - $Y $
j PEDF & EEW COPY si '
) CONTAINMENT SYSTEMS 3
p_._. ----- BASES -- - ;
10
,W6 DRWELL VENT AND PURGE (Continued).
a _-
ki - Operation of the drywell vent and purge 24-inch supply and exhaust valves during plant operational conditions 4 and 5 is unrestricted; the-50* blocks
O may be removed to allow full opening of the valves, and the cumulative time
.ij for v,ent and purge operation is unlimited.
-). 3/4.6.3 DEPRESSURIZATION SYSTEMS _ .
1 The specifications of this section ensure that the drywell and containment pres- ,
4* sure will not exceed the design pressure of 30 psig and 15 psig, respectively,
,j<: during primary system blowdown from full operating pressure.
5.1 . The suppression pool water volume must absorb the associated decay and structural -
.:' sensible heat released during a reactor blowdown from 1040 psia. Using conser-vative parameter inputs, the maximum calculated containment pressure during and j following a design basis accident is below the containment design pressure of
g. 15 psig. Similarly the drywell pressure remains balow the design pressure of
.j 30 psig.- The maximum and minimum water volumes for the suppression pool are m

c 150,230 cubic feet and 146,400 cubic feet, respectively. These values include
,9 the water volume of the containment pool, horizontal vents, and weir annulus.
q'. Testing in the Mark III Pressure Suppression Test Facility and analysis have -
G assured that the suppression pool-temperature will not rise above 185*F for the
$ full range of break sizes. - -
h~
3!
. Should it be necessary to make the suppression pool in' operable, this shall only be done as specified.in Specification 3.5.3. -
16
] Experimental data indicates that effective steam condensation without exco.ssive q load on the containment pool walls will occur with a quencher device and pool 9 temperature below 200*F during relief valve operation. Specifications have been.
M
placed on the envelope of reactor operating conditions to assure the bulk pool 4 temperature does not rise above 185*F in compliance with the containment struc-h tural design criteria.
4~.
$ In' addition to the limits on temperature of the suppression pool water', operat-ing procedures define the action to be taken in the event a safety-relief valve ni S inadvertently opens or sticks open. As a minimum this action shall include:
P (1) use of all available means to close the valve, (2) initiate suppression pool
. water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief
'9'd valves are used to depressurize the' reactor, their discharge shall be separated
,q from that of the stuck-open safety relief valve to assure mixing and uniformity
.3 of energy insertion to the pool. "
y The containment spray system consists of two 100% capacity trains, each with
.g y three4 pray rings located at different elevatioas about the inside circumference of the containment. RHR A pump supplies one train and RHR pump B supplies the 3
-) other. RHR pump C cannot supply the spray system. Dispersion of the flow of
.0 water is effected by 251 nozzles in each train, enhancing the condensaton of
M l
} CLINTON - UNIT 1 8 3/4 6-6 -
t d .. AUG 2 S 1985
g ,. ..
g
.x x . . m. pe.wxmpaa i .- .
1 PM0F & REYlE# COPY I
I.
! CONTAINMENT SYSTEMS BASES 3/4.6.5 ORYWELL POST-LOCA VACUUM RELIEF VALVES .
.The post-LOCA drywell vacuum relief valve system is provided to . relieve the vacuum j in the drywell due to steam condensation following blow-down. Containment air i j is drawn through the vacuum relief valve check valves in the two branches of f separate post-LOCA vacuum relief line and in a branch of each drywell puroe j :ompressor discharge line. Vacuu'n relief initiates at a differential pres .ur, j of one psi. This vacuum relief, in conjunction with the rest of the dryweil purge system, is necessary to insure that the post-LOCA drywell H2 concentra-tion does not exceed 4% by volume.
1 l Following vacuum relief, the drywell purge system pressurizes the'drywell, for
.j - ing noncondensibles through the horizontal vents and into the containment at a rate designed to maintain the H2 concentration below the flammable limits.
L There are two 100% vacuum relief systems .so that the plant may continue opera j tion with one system out of service for a limited period of time.
i 1 -
Four vacuum breaker lines, with two valves in series in each line are provided.
[ Any (three) vacuum relief valve lines can provide full vacuum relief capability 3/4.6.6 SECCNDARY CONTAINMENT The secondary containment completely encloses the primary containmen't, except
,; for the upper personnel hatc.h. It consists of the fuel building, gas control
?;, boundary, and portions of the auxiliary building enclosed by th'. extension of -
1 .the gas control boundary and the ECCS cubicles. The standby r;as treatment .
] system (SGTS) is designed to achieve and maintain a negative 1/4" W.G. pressure
within the secondary containment following a design bas ~is' accident. This 4- design provides for the capture within the secondary containment of the radio-q active releases from the primary containment, and their filtration before 3 . release to the atmosphere.
-4 f' l Establishing and maintaining a vacuum in the secondary containment with the I standby gas treatment system once per 18 months, along with the survei.llance of the doors, hatches, dampers, and valves, is adequate to erisure that there are
~
v no violations of the integrity of the secondary containment.
3 "
The OPERABILITY of the standby gas treatment systems ensures that sufficient j iodine removal capability will be available in the event of a LOCA. The reduc-
) tion in containment iodine inventory reduces the resulting site boundary radia-4 tion doses associated with contai.iment leakage. The operation of this system q g and resultant iodine removal capacity are consistent with the asstimptions used .
in the LOCA analyses. Continuous operation of the system with the heaters ij, , M , OPERABLE for 10 hcurs during each 31 day period is sufficient to reduce the -
j(Q uildup of moisture on the absorbers and HEPA filters. ._
$ Drywen We.uum re. lief valves are provided on 4hcdepuell fc Pass
$Y' %4Clea+
4a preeny quan4an excess(Nesnega+we.
of' gas -9eom %e.horn pressure. con ke
%wevft en topin3toi n%e dcy w%e j% 4
'j k CLINTON - UNIT 1 8 3/4 6-8 AUG 2 31985
k l M- i d b + W * % & U I&:$ .~ N "- T - :
w. :. = - a -
1 J
_ PROOF & Ras COPY l i
'i l
i'~- ADMINISTRATIVE CONTROLS .
i 3 ,
6.4 TRAINING.
j 6.4.1 A retraining and replacement training program for the unit ~ staff shall be maintained under the direction of the Director-Nuclear Training, shall meet
{
, ) or exceed the requirements and recommendations o Section 5.5 of ANSI Standard d 1i N18.1-1971 and Appendix "A" of 10 CFR Part 55 an i the supplemental requirements specified in Sections A and C of Enclosure 1 of .he March 28, 1980 NRC letter
.J,j to all licensees, and shall include familiarization with relevant industry -
y operational experience.
A y.;/
6.5 REVIEW AND AUDIT . ..
S .
6.5.1 FACILITY REVIEW GROUP (FRG) . . . . _ .;..._
.q .
.3 FUNCTION el
'l4 6.5.1.1 The FRG 'shall function to advise'the Power Plant Manager on all 3 matters related to nuclear safety.
1
}b p., COMPOSITION y . . ..
.i .. 6.5.1.2 The FRG shall be composed.of.the: . .
'. Chairman: - Assistant Pow 6r Plant Manager-Operations -
Member: D/necVor -f.;;i:trt ":n. '"xt ":r-iFMaintenance
.'N- .
Member: Arese- L.;.;c.i; . T k 8 0perationi
.N'; Member: c Nar S W Technical Member: Supervisofr-I -
f.i Member: F,ce s e St;;;-;i; . -Radiation Protection ,
. Member: Supervisor _Huclear Member: fr;r"i;;c" Quality Assurance Refe.esenhuis 1
d . ALTERNATES. . .
-T.
.] 6. 5.1. 3 All alternate members shall be appointed in writing by the FRG t1 Chairman to serve on a temporary basis; however, no more than two alternates Q shall participate as voting members in FRG activities at^any one. time..
, 9 3g - MEETING FREQUENCY -
Ij 6.5.1.4 The FRG shall meet at least once per calendar month and as convened
-: by the FRG Chairman or his designated alternate. -
r,
$ QUORUM '-
.1 -
Jj . . 6.5.1.5 The quorum of the FRG necessary for the performance of the FRG -
i responsibility and authority provisions of these Technical Specifications
.: shall consist of the Chairman or his designated alternate and four members including alternates.
i ;,
j CLINTON - UNIT 1 6-7 AUG :S 1985
.___,_._____.,.__,_.7_____m__ _ , _ _ _ , , , ,..,,,,.__.,-_,,-,,,,,,,,_.--y ,__.,,_,_,_y--____ __..y,, _ _ _ , _ _ -. -,,p , , _ - - _v.--..

U-600275, Forwards marked-up Proof & Review Tech Specs,Per 850904 Request,Indicating Typos & Corrections to Info Submitted in 840928 & s

Text

Subject:

Dear Mr. Butler:

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U-600275, Forwards marked-up Proof & Review Tech Specs,Per 850904 Request,Indicating Typos & Corrections to Info Submitted in 840928 & s

Text

Subject:

Dear Mr. Butler:

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