Text

Forwards Proposed Amends to Draft Tech Specs W/Discussion of Justification
	ML20092L140
Person / Time
Site:	Catawba
Issue date:	06/25/1984
From:	Tucker H; DUKE POWER CO.
To:	Adensam E, Harold Denton; Office of Nuclear Reactor Regulation
References
	NUDOCS 8406290256
	Download: ML20092L140 (74)
	v • d • e

f .- t rt DUKE POWER GoxPAxy P.O. BOX 33180 OllAHLOTTE, N.C. 28242 HALB. TUCKER reternoxe vers casanomwr (704) 073-4531

.mu. --- June 25, 1984 Mr. Harold R. Denton, Director t Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 r Re: Catawba Nuclear Station, Unit 1 ,

Docket No. 50-413 !

Draft Technical Specifications

Dear Mr. Denton:

Attachments 1-5 of this letter contain proposed amendments to the Draft Technical Specifications for Catawba Unit 1. Each attachment contains the proposed changes and a discussion of the justification.

Very truly yours, haJB.TA Hal B. Tucker RWO/rbs Attachments cc: Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region il 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 NRC Resident inspector Catawba Nuclear Station Mr. Robert Guild, Esq.

Attorney-at-Law P. O. Box 12097 Charleston, South Carolina 29412 i Mr. Jesse L. Riley Carolin} Environmental Study Group 854.Henley Place Charlotte, North Carolina 28207 ()O [ '

Palmetto Alliance g6 6 040625

.2135i Devine Street A 05000413 Columbia, South Carolina 29205 PDR

s t Attachment 1 Page 1 Table 4.3-2, page 3/4 3-46 Item 8.d needs to be revised to include a note (3) under TRIP ACTUATING DEVICE OPERATIONAL TEST. Note (3) would read: "(3) Monthly testing shW1 consist of relay testing exluding final actuation of the pumps or valves."

This is needed in order to avoid having to operate the Auxiliary Feedwater pumps and valves more often than is necessary. Specification 4.7.1.1.1.b calls for verification of these valves and pumps every 18 months. This is adequate to ensure operability of the system. '

Item 8.f needs to be revised as shown since the installed instrumentation ;

are pressure switches and do not contain an analog channel.

Page 3/4 3-47 items 10.a and 10.b need to be revised to include a note (2) under TRIP ACTUATING DEVICE OPERATION TEST. Note (2) should read: -

"(2) Monthly testing shall consist of voltage sensor relay testing excluding actuation of the load shedding diesel start and time delay timers."

This is needed in order to avoid having to operate the diesel generators from cold conditions. Monthly cold starting of the diesels will have an adverse impact on the diesels' components by causing unnecessary excessive wear.

Item ll.b should be revised as shown in order to provide consistency between item 11.b and Specification 4.7.6.e.2.

Page 3/4 3-48, Item 14.c needs to be revised as shown for the same reasons cited above. Specification 4.7.4.b calls for verification of the operability of the Nuclear Service Water System pumps and valves. Item 14.g needs to be revised as shown for the same reasons cited above for Item 8.f.

Page 3/4 4-49, Item 15.c needs to be revised as shown for the same reasons cited for Items 10.a and 10.b. Item 16.c should be deleted since there is realignment of the Auxiliary Building Filtered Ventilation Exhaust System upon receipt of a Loss-of-Offsite Power signal. Table 4.3-8, page 3/4 3-80, Items 2, 3.a and 3.b should be revised as shown for the same reason cited for Item 8.f on Table 4.3-2.

~

TABLE 4.3-2 (Continued) -

hh

> ENG_INEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

~

2-SURVEILLANCE REQUIREMENTS TRIP g, ANALOG ACTUATING MODES

q CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE j FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST 2S REQUIRED

8. Auxiliary Feedwater

a. Automatic Actuation N.A. N.A N.A. N.A. M(1) M(1) Q 1,2,3 Logic and Actuation Relays

b. Steam Generator S R H N.A. N.A. N.A N.A 1,2,3 Water Level-Low-Low g; c. Safety Injection See Item 1. above for all Safety njection Surveillance Requirements.

(( d. Los s-of-Of fsi te N.A. R N.A. M(3$) N.A. N.A. N.A 1,2,3

Power

e. Trip of All Main N.A. N.A. N.A. R N.A. N.A. N. A 1, 2 Feedwater Pumps

f. Auxiliary Feedwater Suction Pressure- g4 Low J' gA JI )[ /4 S/I. 8 N.A. N.A.. N. A. 1,2,3

9. Containment Sump Recirculation

a. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Logic and Actuation Relays

b. Refueling Water S R H N.A. N.A. N.A. N.A. 1, 2, 3, 4 Storage Tank Level -

Low-Low Coincident With >

Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

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TABLE 4.3-2 (Continued) '

5?

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION s2 SURVEILLANCE REQUIREMENTS

$r TRIP

. ANALOG ACTUATING c CHANNEL DEVICE MODES l-2 CHANNEL MASTER SLAVE FOR WHICH -

-4 CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY FUNCTIONAL UNIT CHECK CALIBRATION TEST RELAY SURVEILLANCE!

s TEST LOGIC TEST TEST TEST IS REQUIRED .

10. Loss of Power

a. 4 kV Bus N.A. R N.A. N.A. N.A. N.A.

Undervoltage-Loss M(2 1,2,3,4 '

of Voltage

b. 4 kV Bus N.A. R N.A. N.A. N.A.

Undervoltage-Grid M(2)) N.A. 1,2,3,4 Degraded Voltage w 11. Control Room Area Ventilation Operation

,p,

); a. Automatic Actuation N.A. N.A. N.A. N.A.

!%C3 M(1) M(1) Q All [E==

w Logic and Actuation ""T' A Relays w H

b. Loss-of-Of fsite Poser N.A. R N.A. )( N.A. N.A. N.A. 1, 2, 3

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

, 12. Containment Air Return and Hydrogen Skimmer Operation

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4

$ b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) 1,2,3,4 Q

Actuation Logic and Actuation Relays

c. Containment S R M N.A. N.A. N.A. N.A. 1,2,3 2-Pressure-High-High j[((.

mn l 13. Annulus Ventilation v2 5I' Operation E-e

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4 -.

- - . - . _ - --. . . .- --- - -.-- _ .-= -

i ,

' TABLE 4.3-2 (Continued) 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

$ SURVEILLANCE REQUIREMENTS I

5 2,

TRIP e' ANALOG ACTUATING MODES c- CHANNEL DEVICE MASTER SLAVE FOR WHICH i5 CHANNEL CHANNEL CHANNEL- OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST

] TEST IS REQUIRED

.13. Annulus Ventilation Operation (Continued)

b. Automatic ,N.A. N.A. N.A. N.A. M(1) M(1) 1,2,3,4 Q

Actuation Logic and Actuation Relays

c. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

g 14. Nuclear Service Water Operation y a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4 s b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Actuation Logic

and Actuation Relays

c. Loss-of-Offsite N.A. R N.A. M(3) N.A. N.A. N.A. 1,2,3 Power

d. Containment Spray See Item 2. above for all Containment Spray Surveillance Requirements.

e. Phase "B" Isolation See Item 3.b. above for all Phase "B" Isolation Surveillance Requirements.

f. Safety Injection Sje-I em 1. above for all Safety Injec

n> Surveillance Requirements. .T

g. Suction Transfer- '

Low Pit Level /g )( gA [gA jf. N.A. N.A. N.A. 1,2,3 ,

15. Emergency Diesel R

Generator Operation 80 (Diesel Building . %9 Ventilation Operation, a$

Nuclear Service Water 5 Operation) *

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4

. , - . ~ r -- ,- r--, -----

e

TABLE 4.3-2 (Continued) 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION c SURVEILLANCE REQUIREMENTS

$ TRIP -

. ANALOG ACTUATING MODES c- CHANNEL DEVICE MASTER SLAVE FOR WHICH 5

CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST w IS REQUIRED .

t

15. Emergency Diesel j Generator Operation '

(Diesel Building Ventilation Operation Nuclear Service Water Operation) (Continued)

b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Actuation Logic w and Actuation

) Relays Y c. Loss-o f-Of fsi te N.A. R N . ,A. M(Z N.A. N.A. N.A. 1,2,3

$ Power

d. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

16. Auxiliary Building M Filtered Ventilation T Exhaust Operation 9 i

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1 2,3,4 g

b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 E i

Actuation Logic 9:

'. and Actuation Re1 y s

5 L :: "-off: t& w -*- *-A- -*- e -*-A. -M-*r 2, 2, 2 -

k Pe"e- w o

d. fety Injection See Item 1. above foNafety-Injectinn _Su_rveillance Requirements. '@

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

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-TABLE 4.3-2 (Continued) 9 .

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

'$ SURVEILLANCE REQUIREMENTS !

5 TRIP l

ANALOG ACTUATING MODES c CHANNEL DEVICE MASTER SLAVE FOR WHICH E CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE

[ FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC' TEST TEST TEST IS REQUIRED :

17. Diesel Building '

Ventilation Operation i

, a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4

b. Automatic N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 Actuation Logic ;

and Actuation i

Relays O !

y c. Emergency Diesel See Item 15. above for all Emergency Diesel Generator Operation Surveillance !

Generator Operation Requirements w 7 t E 18. Engineered Safety Features ~

Actuation System Interlocks I

a. Pressurizer N.A. R H N.A. N.A. N.A. N.A. 1,2,3 ,b Pressure P-11 j?

b. Pressurizer N.A. R M N.A. N.A. N.A. N.A 1,2,3 !

, Pressure, not P-11 i

c. Low-Low T,yg, P-12 N.A. R M N.A. N.A. N.A. N.A. 1, 2, 3 ,!

a ,

d. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 3

e. Steam Generator S H N.A.

j Water Level, P-14 R M(1) M(1) Q 1,2,3 hl g.!

i

5,f TABLE NOTATION i

(1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. u'

(I) MonEly feshng .sh4([ consis4 of Voda scosoc ce fes}/n ewc.luding qe/uedon o[ Me' /oc

, ShedE9, d.sese.l skr4 ud Nme ele y bmers.

a v . g usti,s w w 1., m a , u u a ev . - ._

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j . TABLE 4.3-8. -

t i! O RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

,{ f ANALOG

, CHANNEL 2

CHANNEL. SOURCE CHANNEL OPERATIONAL

) E INSTRUMENT CHECK CHECK CALIBRATION TEST H

H 1. Radioactivity Monitors Providing /

Alarm and Automatic Termination

of Release

,l a. Waste Liquid Discharge Monitor (Low Range - D P R(2) Q(1) f EMF-49)

'lI' b. Turbine Building Sump Monitor (Low Range - D M R(2) Q(1)

EMF-31)

'i j c. Steam Generator Water Sample Monitor (EMF-34) D M R(2) -Q(1) y 2. Continuous Composite Samplers and Sampler g Flow Monitor -

Conventional Waste Water Treatment Line D N.A. R

/MA

3. Flow Rate Measurement Devices

a. Waste Liquid Effluent Line D(3) N.A. R

! b. Conventional Waste Water Treatment Line D(3) N.A. R ,

J

c. Low Pressure Service Water Minimum Flow D(3) N.A. R Q i Interlock #

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i Attachment 2 Proposed Amendment to Catawba Unit 1 Draft Technical Specification 4.5.2.d.1.b Concerning ECCS Subsystems - Tavg 3 350 F

.. . Attachment 2 Page 1

. o The proposed revision would change the setting for the interlocks which cause the valves to autor,!atically close from "less than or equal to 600 psig" to "less than or equal to 660 psig." The change is required to provide an allowance for possible drift and instrument error.

-Station procedures specify a setting of 600 psig. Increasing the setting in the Technical Specifications will allow cperating flexibility and avoid a Technical Specification violation if instrumentation drifts.

The setting will remain at 600 psig per the station procedures. The Surveillance Requirements to verify the setpoint will also remain the same.

, , Attachment 2 Page 2 EMERGENCY CORE COOLING SYSTEMS DE y..-

SURVEILLANCE REQUIREMENTS (Continued) A b) l fofoO With a simulated or actual Reactor oolant System pressure signal less than or equal to sig the interlocks will cause the valves to autom ly close.

j 2) A visual inspection of the containment sump and verifying that the subsystem suction inlets are not restricted by debris and that the sump corrponents (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

e. At least once per 18 months, during shutdown, by:

1) Verifying that each automatic valve in the flow path actuates to its correct position on Safety Injection and Containment s

Sump Recirculation test signals, and

} 2) Verifying that each of the following pumps start automatically upon receipt of a Safety Injection test signal:

a) Centrifugal charging pump, b) Safety Injection pump, and c) Residual heat removal pump.

f. By verifying that each of the following pumps develops the indicated differential pressure when tested pursuant to Specification 4.0.5:

1) Centrifugal charging pump 1 2380 psid,

2) Safety Injection pump > 1430 psid, and

3) Residual heat removal pump 1 165 psid.

g. By verifying the correct position of each electrical and/or mechanical stop for the following ECCS throttle valves:

1) Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS subsystems are required to be OPERABLE, and

, _ _ _ _ At least once per 18 months.

Injection Throttle Safety Injection Throttle -

Cerdd. Valve Number Valve Number d4g4g i

[.

p"E 1141-14 (I-16 1NI-164 1NI-166 I-18 1NI-168 NI-20 INI-170

CATAWBA - UNIT 1 3/4 5-7

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Attachment 3 l t

Proposed Amendment to Catawba Unit 1 Draft .

Technical Specification 4.6.1.3.b Concerning l

- Containment Airlock Surveillance

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Attachment 3 Page 1 Catawba Unit 1 Draft Technical Specification 4.6.1.3.b currently requires overall containment airlock leakage tests to be performed "...if opened when CONTAINMENT INTEGRITY was not required..." The proposed amendments would change this to require the overall airlock leakage test to be performed

...when maintenance has been performed on the air lock that could affect the air lock sealing capability." This proposed change constitutes an exemption to Appendix J of 10 CFR 50.

The proposed amendments are justified for several reasons:

(1) Opening the airlock, i.e., opening both doors simultaneously, is no different in terms of capability to reseal than opening one door at a time during normal entries.

(2) Test data taken on McGuire Unit 1 (Catawba and McGuire have the same air locks) on sixteen different occasions since June 1981 have not indicated any tendency of the airlock leakage rate to increase after opening both airlock doors simultaneously. (See the attached t'.ble.)

(3) The overall airlock leakage rate will be measured at least once per 6 months regardless of the airlock operating or maintenance history. Also, the test would be performed after maintenance activities potentially affecting the airlock sealing capability.

(4) A seal integrity test is performed prior to establishing containment integrity and once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months per Specification 4.6.1.3.a. This is a more meaningful and more conservative test for detecting seal problems than the overall airlock leakage rate test because it verifies the integrity of each seal on each door. Because the overall airlock leakage test involves pressurizing between the doors, this test only verifies that at least one of the two seals on each door is sealed. (The airlocks have four seals between containment and ouside.) The overall airlock leakage rate test might detect potential problems with the airlock not related to the door seals; however, such problems would not occur as a result of opening both doors simultaneously.

(5) The current requirement poses a significant burden. The airlocks will usually be opened during outages to facilitate equipment transport into and out of containment. Then just prior to entry into Mode 4, the overall airlock leakage test must be performed.

Installing strongbacks, performing the test, and removing strongbacks will require at least 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months per airlock during which access through the airlock is prohibited. Any access and egress to lower containment during testing of the lower airlock will involve climbing through the emergency hatch between upper and lower containment. This will result in more contamination in upper containment which will usually be cleaner than lower containment. Similarly access to upper

.. ..- Attachment 3 -Page 2 containment while testing the upper airlock will require passing through lower containment where radiation levels will be higher, thus increasing radiation exposure to personnel and increasing

. contamination in upper containment. The proposed changes would allow better scheduling of the overall airlock leakage test during periods when the need for access to containment is minimal.

The proposed amendments would remove the requirement to test the overall airlock '

leakage after each opening of the airlock and add a requirement to test whenever the airlock sealing capability might have been affected by maintenance. Because '

any effect on the airlock sealing capability potentially caused by opening the l doors would be detected by another required test and because the overall airlock ;

leakage test will be performed every 6 months and after maintenance, the proposed changes are insignificant to safety. ;

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r Attachment 3 Page 3 McGuire Unit 1 Containment Air Lock Leakage Data The following test data from June 1981 to April 1983 show the containment I air lock leakage rates measured after the air locks had been opened. Note that all tests met the acceptance criterion of. 0.05 La (4530 sccm).

t Upper Airlock Lower Airlock :

Date- Leakage (sccm) Date Leakage (sccm) l 04/23/83 580 04/25/83 655 11/18/82 412 11/20/82 410 07/13/82 290 07/14/82 115 03/12/82 865 03/13/82 225

, l12/29/81 .193 12/28/81 417 11/20/81 258 11/22/81 751 06/06/81 45 10/02/81 492 08/03/81 951 06/11/81 259 Y

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Attachment 3 Page 4 CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS 4.6.1.3 Each containment air lock shall be demonstrated OPERABLE:

a.

Within 72 hours8.333333e-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 that the seal leakage is less than 0.01 L, as determined by precision flow measurements when measured for at least 30 seconds with the volume between the seals at a constant pressure of 14.7 psig; b.

By conducting overall air lock leakage tests at not less than P,,

14.7 psig, and verifying the overall air lock leakage rate is within its limit:

1) At least once per 6 months,# and

2) Prior to establishing CONTAINMENT INTEGRITY 4f g reed t .,

a-CCNT".IPST D'TEC"ITY :: =t rgid c.

At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

d.

'At least once per 6 months by conducting a pressure test at not less than P , 14.7 psig, to verify door seal integrity, with a measured leak r$te of less than 15 sccm per door seal.

when ma iden<nce. has b8cn fe. cme d on O c air lock da[ 6uN dfed % a;e lo&. seating c.qd;l:Iy.##

d The provisions of Speci'fication 4.0.2 are not applicable.

h'This conkl es an egemfico / lo Affew/iy gT d ]d C FA 50.

CATAWBA - UNIT 1 3/4 6-9 '

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Attachment 4 Proposed Amendment to Catawba Unit 1 Draft Technical Specifications Concerning Radiation Monitoring Instrumentation

Attachment 4 Page 1 The proposed change would allow radiation monitor EMF-39 to not be operable prior to initial criticality.

In Inspection Report 50-413/84-50, a Region II inspector identified an Inspector Follow-up Item 84-50-01 concerning unsatisfactory installation of EMFs 38, 39, and 40. In order to resolve this item a Nuclear Station Modification (NSM) has been written. The work required will not be able to be completed until prior to initial criticality.

EMF-39 is a high gaseous radioactivity containment atmosphere monitor. Since there will not be any appreciable radioactivity in the containment prior to initial criticality, an exemption in the form of the proposed Technical Specification change is requested.

~

TABLE 3.3-6 9

g E RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS 8

c MINIMUM

{

FUNCTIONAL UNIT CHANNELS TO TRIP / ALARM CHANNELS OPERABLE APPLICABLE MODES ALARM / TRIP.

SETPOINT ACTION

1. Containment '

a. Containment Atmosphere - High 1 1 All ***

Gaseous Radioactivity (Low 30 Range - EMF-39)

b. Reactor Coolant System Leakage Detection i

1) Particulate Radioactivity y (Low Range - EMF-38) N.A. 1 1,

2) Gaseous Radioactivity Q4 '

H.A. ;

33 (Low Range - EMF-39) N.A. 1 m '

4 N.A. 33

2. Fuel Storage Pool Areas _

a. High Gaseous Radioactivity (Low Range - EMF-42) 1 1 **

5 1.7x10 4 pCi/ml 34

b. Criticality-Radiation Level (Fuel Bridge - Low Range -

EMF-15) 1 1 *

5 15 mR/h 32

3. Control Room >

N Air Intake-Radiation Level - g 1/ietehe /Ntd:

! High Gaseous Radioactivity 54aha /ie All ~< 1.7x10 4 pCi/ml 31 5

( p (Low Range - EMF-43 A & B) k 7 4. Auxiliary Building Ventilation ,1 1 All 5 1.70x10 4 pCi/mi 35 a

High Gaseous Radioactivity u h

(Low Range - EMF-41)

S. "

l Component Cooling Water System (EMF-46 A&B) 1 1 All i 1x10 2 pC1/ml 36

,. , Attachment 4 Page 3" !

TA8LE 3.3-6 (Continued)

. TABLE NOTATIONS '

i With fuel in the fuel storage pool areas. '

With ifradiated fuel in the fQel storage pool areas.

Trip Setpoint concentration value (pci/ml) is to be established such that t the actual submersion dose rate would not exceed 2 mR/h in the containment i building. The Setpoint value may be increased up to the equivalent limits of Specification 3.11.2.1.in accordance with the methodology and parameters i in the 00CM during containment _ purge or vent provided the Setpoint value doespxceed tg the maximum _ concentyation activity in the containment l determined'by the sample analysis perf6rmed for to each release in accordance with Table 4.11-2. >

.jf A of

- l cg{icaN rVor /c inika/ cHkeaCh. .

f ACTION STAT

(

ACTION 30 - With less than the Minimum Channels OPERABLE requirement, j operation may continue provided the containment purge and l exhaust valves are maintained closed. ,

i ACTION 31 - With the number of operable channels one less than the Minimum

Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate the affected ;

, Control Room Ventilation System intake from outside air with -

(- recirculating flow through the HEPA filters and charcoal adsorbers.

ACTION 32 - With less than the Minimum Channels OPERABLE requirement, opera- i tion may continue for up to 30 days provided an appropriate !

portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable +

monitors to OPERABLE status within 30 days or suspend all >

operations involving fuel movement in the fuel building. -

ACTION 33 - Must satisfy the ACTION requirement for Specification 3.4.6.1.' ;

j ACTION 34 - With the number of OPERABLE channels less than the Minimum ,

t Channels OPERABLE requirement, operation may continue provided !

i the Fuel Handling Ventilation Exhaust System is operating and

discharging through the HEPA filters and charcoal adsorbers. {

Otherwise, suspend all operations involving fuel movement in j the fuel building.

l ACTION 35 - With the number of OPERABLE channels less than the Minimum Channels '

OPERA 8LE requirement, operation may continue provided the Auxiliary Building Filtered Ventilation Exhaust System is operating and l ,

discharging through the HEPA filter and charcoal adsorbers. !

ACTION 36 - With the number of OPERABLE channels less than the Minimum Channels ;

OPERABLE requirement, operation may continue for up to 30 days j

{ provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are i

( collected and analyzed for radioactivity (gross gamma) at a lower ifmit of detection of no more than 10 7 pCi/ml.

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TABLE 4.3-3 n .

H RADIATION MONITORING INSTRUMENTATION FOR PLANT h

> OPERATIONS SilRVEILLANCE REQUIREMENTS s '

ANALOG g CHANNEL MODES FOR WHICH ,

CHANNEL CHANNEL OPERATIONAL SURVEILLANCE

[ FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED -

1. Containment -

a. Containment Atmosphere - High i

Gaseous Radioactivity (Low f

/ g D Range - EMF-39) S R _ M All

! b. Reactor Coolant System Leakage Detection S R M 1,2f3f4 (Low Range - EMF-38 and Low Range - EHF-39) w

, D 2. Fuel Storage Pool Areas -

Y a. High Gaseous Radioactivity S R M **

$ (Low Range - EMF-42) '

b. Criticality-Radiation Level S R M *

(Fuel Bridge - Low Range -

EMF-15)

3. Control Room 4

Air Intake Radiation Level - S R M -

All

! High Gaseous Radioactivity -

! (Low Range - EMF-43 A & B)

4. "o.

Auxiliary Building Ventilation g.

c High Gaseous Radioactivity (Low Range - EMF-41)

S R H All e

t C l\ 4 I *

5. Component Cooling Water System S R M All

$ (EHF-46 A&B) E E

h. TABLE NOTATIONS a

With fuel in the fuel storage pool area. ,

- With irradiated fuel in the fuel storage pool areas.

Alof agliceWe frloe b MibW cn/iuff.

p <.

)

Attachment 5 Proposed Amendment to Catawba Unit 1 Draft Technical Specifications Supplying Infor.'ation and Correcting Errors ex-

s b

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS .

3/4.5.1 ACCUMULATORS ,

t Cold Leg Injection.......................................

3/4 5-1 Upper Head Injection.....................................

3/4 5-3 3/4.5.2 ECCS SUBSYSTEMS - T,yg > 350 F........................... 3/4 5-5 3/4.5.3 ECCS S UB S YS TEMS - T,y g < 3 5 0 F . . . . . . . . . . . . . . . . . . . . . . . . . . .3/4 5-9 3/4.5.4 REFUELING WATER STORAGE TANK............................. 3/4 5-11 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity.................................... 3/4 6-1 Containment Leakage...................................... 3/4 6-2 TABLE 3.6-1 PATHS............

SECONDARY CONTAINMENT BYPASS LEAKAGE 3/4 6-5 Containment Air Locks.................................... 3/4 6-8 Internal Pressure........................................ 3/4 6-10 Air Temperature.......................................... 3/4 6-11 Containment Vessel Structural Integrity.................. 3/4 6-12 Reactor Building Structural Integrity.................... 3/4 6-13 Annulus Ventilation System............................... 3/4 6-14 Containment '!: y s t em A . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-16 3/4.6.2 DEPRESSURIZATIO aHD $LINGSYSTEMS Containment Spray System................................. 3/4 6-18 t

3/4.6.3 CONTAINNENT ISOLATION VALVES............................. 3/4 6-20 TABLE 3.6-2 CONTAINMENT ISOLATION VALVES.......................... 3/4 6-22 3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Monitors........................................ 3/4 6-30 Electric Hydrogen Recombiners. . . . . . . . . . . . . . . . . . . . . . . . . . . . 3/4 6-31 Hydrogen Mitigation System............................... 3/4 6-32 3/4.6.5 ICE CONDEPSER Ice Bed ................................................ 3/4 6-33 Ice Bed Temperature Monitoring System.................... 3/4 6-35 Ice Condenser Doors...................................... 3/4 6-36 Inlet Door Position Monitoring System.................... 3/4 6-38 i CATAWBA - UNIT 1 VIII JUN 8 UMW

r -,

s 3

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS t SECTION PAGE Divider Barrier Personnel Access Doors and

~

4 Equipment Hatches...................................... 3/4 6-39 Containment Air Return and Hydrogen Skimmer Systems...... 3/4 6-40 Floor Drains............................................. 3/4 6-42 Refueling Canal Drains'................................... 3/4 6-43 Divider Barrier Seal..................................... 3/4 6-44 TABLE 3.6-3 DIVIDER BARRIER SEAL ACCEPTABLE PHYSICAL PROPERTIES... 3/4 6-45 3/4.6.6 CONTAINMENT VALVE INJECTION WATER SYSTEM ................ 3/4 6-46 ;

3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE !

Safety Va1ves............................................ 3/4 7-1 i TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH. ;

..SETPOINT..WI.TH. INOPERABLE STEAM LINE SAFETY VALVES DURING '

FOUR LOOP 0PERATION...................................... 3/4 7-2 i TABLE 3.7-2 STEAM LINE SAFETY VALVES PER L00P..................... 3/4 7-3 Auxiliary Feedwater System............................... 3/4 7-4 Specific Activity........................................ 3/4 7-6 1 ;

TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PR0 GRAM..................................... 3/4 7-7 Main Steam Line Isolation Valves......................... 3/4 7-8 i 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION.......... 3/4 7-9 ,

3/4.7.3 COMPONENT COOLING WATER SYSTEM........................... 3/4'7-10 :

-3/4.7.4 NUCLEAR SERVICE WATER SYSTEM............................. 3/4 7-11 5 3/4.7.5 STANDBY NUCLEAR SERVICE WATER P0ND....................... 3/4 7-12 3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM.. ................. 3/4 7-13 3/4.7.7 AUXILIARY BUILDING FILTER D L.,..L",T = EXHAUST SYSTEM... 3/4 7-16 3/4.7.6~ SNUBBERS................... ................... 3/4 7-18 FIGURE 4.7-1 5 j SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST........... 3/4 7-23 3/4.7.9 SEALED SOURCE CONTAMINATION.............................. 3/4 7-24 3/4.7.10 FIRE SUPPRESSION SYSTEMS

-Fire Suppression Water System............................ 3/4 7-26 Spray and/or Sprinkler Systems........................... 3/4 7-28 l CO 2 Systems.............................................. 3/4 7.30 I Fire Hose Stations....................................... 3/4 7-32 [

CATAWBA - UNIT 1 IX , ;

JUN 819K ;

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

__.-.m.

..~,__,.__--.w.._.

.. .- l

\~

BASES ~

t SECTION -- - - -

PAGE i i

t

, TABLE B,3/4.4-1 _REAC_ TOR _ VESSEL _ TOUGHNESS _........._................. B 3/4 4-9

~ ~ ' ~ ~

FIGURE B 3/4.4-l'. FAST' NEUTRON FOJENCE {E>1MeV)"AS'A FUN'CTION 0F~ ~

FULL POWER SERVICE LIFE.................................. B 3/4 4-10 FIGURE B 3/4.4-2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT I 0F RT FOR REACTOR VESSELS EXPOSED TO 550 NDT F............ B 3/4 4-11 3/4.4-10 STRUCTURAL INTEGRITY.....................................

_ B 3/4 4-15 1

3/4.5 EMERGENCY CORE COOLING SYSTEMS '

.3/4.5.1 ACCUMULATORS..............................................

. B 3/4 5-1 l 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS............................... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.............................. B 3/4 5-2 l 3/4.6 CONTAINMENT SYSTEMS _. 1_ ..._...._._._. _ _._.___ __._. _ _ __ _ _ .

3/4.6.1 PRIMARY CONTAINMENT....................................... B 3/4 6-1 3/4.-6.2 DEPRESSURIZATION AND COOLING SYSTEMS..........'............ B 3/4 6-4 ,

3/4.6.3 CONTAINMENT ISOLATION VALVES.............................. B 3/4 6-4 i 3/4.6.4 -COMBUSTIBLE GAS CONTR0L................................... B 3/4 6-4 3/4.6.5 ICE CONDENSER............................................. B 3/4 6-5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE............................................. B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION........... B 3/4 7-2 3/4.7.3 COMPONENT COOLING WATER SYSTEM............................ B 3/4 7-3 [

1 3/4.7.4 NUCLEAR SERVICE WATER SYSTEM.............................. B 3/4 7-3

3/4.7.5 STANDBY NUCLEAR SERVICE WATER P0ND........................ B 3/4 7-3 l 3/4.7.6 CONTROL ROOM AREA VENTILATI M..... . . .............. B 3/4 7-3 1 '3/4.7.7 AUXILIARY BUILDING FILTE D 4TNfftAft9N EX UST SYSTEM.... B 3/4 7-4 3/4.7.8 SNUBBERS................... - . . . . . -

. . . . . .............. B 3/4 7-4

l 3/4.7.9 SEALED SOURCE CONTAMINATION............................... B 3/4 7-6 .

t 3/4.7.10 FIRE SUPPRESSION SYSTEMS.................................. B 3/4 7-6 t

i 3/4.7.11 FIRE BARRIER PENETRATIONS................................. B 3/4 7-7 3/4.7.12 AREA TEMPERATURE MONITORING............................... B 3/4 7-7 3/4.7.13 GROUN0 WATER LEVEL......................................... B 3/4 7-7 l

CATAWBA -. UNIT 1 XIV JUN 81984

- - , . , w w-9+w -ff.- ,. _ 3.g.p.---. -

qm. , _ . - , y. , .--, , __..-..,..,_,g..-mp,_.-.m3 -,.- ., -, , , . _ . . . . . ,

k .

n TA' E,LE 2.2-1 (Continued) ~-

(-4 TABLE NOTATIONS E NOTE 1: 0VERTE;4PERATURE AT

~-

e S 1 AT gy (1 ,Tzg)

+ ri ) (y ,1 Tsg) 5 AT, {Kg-K 2 (1 + gs) r,S)-[T (1 + g S) - T'] + K3 (P - P') - fg(AI)}

ix.l 1 - '

l

[ Where: ' AT =

Measured AT by RTO Manifold Instrumentation; , , , , 'j ; il f

~

= Lead-lag compensnor on measured AT; A *

's ,

=

, , It, r2 Time constants utilized in. lead-lag compensator for AT, rt =8s, (2 =s3 s; 1 #

= Lag compensator 'on measured AT; '

y, 3 O

=

T3 Time constantx u tilized in f.he lag compensator for AT, r3 ,-

'? = '1

[ y AT Indicated AT at RATEI THERMAL POWER; s o '>

2,

= / i i ir' i

K2 l 411; '

K2 = 0.02401/*F; ,

1 =

The function dynamic generated by the lea 6 tag compensator for T,yg.

compensation; ,

14, Ts =

Time 1 = 4 constants s; utilized in the lead-lag compensator for T,yg, 14 = 28 s, 3 .

T = Average' temperature, *F; 1

= i y, 3 Lag compensator on measured T,yg; o

.c

c. - Is =

Time constant utilized in the measured T avg lag compensator, s=[s-

x ca . . -

Yo T

. \. .

x LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux -

\ ~

$o tection The Intermediate and Source Range, Neutron Flux trips provide core pro-during reactor STARTUP to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.

pyy[ Power Range, Neutron Flux channels.These trips provide redundant prote The Source Range channels will initiate t

- a_ Reactor trip at about 10s counts per second unless manually blocked when

~D-6 becomes active] The Intermediate Range channels will initiate a Reactor ;

s trip at a currentJlevel equivalent to approximately 25% of RATED THERMAL POWER 3 unless manually blocked when P-10 becomes active.

t bk, Overtemperature AT g

J The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure ',

trips. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water ;

and includes dynamic compensation for piping delays from the core to the loep temperature detectors, (2) pressurizer pressure, and (3) axial power distribu-tion. With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.2-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced according to the notations in Table 2.2-1.

Overpower AT The Overpower AT trip provides assurance of fuel integrity (e.g. , no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature AT trip, -

and provides a backup to the High Neutron Flux trip. The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water, and (2) rate of change l of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors,.to ensure that the allowable heat generation l rate (kW/ft) is not exceeded. The Overpower AT trip provides protection to i

mitigate the consequences of various size steam breaks as reported in WCAP-9226,

" Reactor Core Response to Excessive Secondary Steam Releases."

1 I

CATAWBA - UNIT 1 B 2-5 1UN 8 IN t

D O

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION .

~

r 3.1.2.6 As a minimum, one of the iollowing borated water source (s) shall be .

OPERABLE as required by Specification 3 1.2.2i

a. A Boric Acid Storage System with: '

1) A minimum contained borated water volume of 19500 gallons,

2) A, minimum boron concentration of 7000 ppm, and

3) A minimum solution temperature of 65*F. ~

b. The refueling water storage tan'k with: /

3M,6/3

~

1) A contained. borated water' volume of at'I ast % 200 ga lons,

2) A minimum boron concentration of 20b0 ppm,

3) A minimum solu' tion temperature of J0 F, and

4) A maximum solution temperature of * ;0 F.

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

ACTION:

a.

With the Boric Acid Storage System-inoperable and being used as one of the above required borated water sources,< restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWH MARGIN equivalent to at least 1% Ak/k at 200*F; restore the Bor tc Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l

b. With the refueling water storage tank inoperable, restore tneftank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY i

within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWtt'within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . '

s. ,

t

'N e

CATAWBA - UNIT 1 3/4 1-12 dVN 8 N

O POWER DISTRIBUTION LIMITS \

3/4.2.3 REACTOR CHANNEL FACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT !

l LIMITING CONDITION FOR OPERATION !

3.2.3 The combination of' indicated Reactor Coolant System total flow rate and

~ ~

R shallfor 3.2-3 befour maintained within the region of allowable operation shown on Figure loop operation. !

~

W'he' re:

N a* "

R = 1.49 [1.0 + 0.3 (1.0 - P)] '

t b* P THERMAL POWER , and

= RATED THERMAL POWER ,

i

c. 'Fh=MeasuredvaluesofFhobtainedbyusingthemovableincore !

detectors to obtain a power distribution map. The measured

{

l valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 .

- -includes. penalties for' undetected feedwate -

ri fouling of ;

0.1% and for measurement uncertainties of 2.y f r flow and 4%

for incore measurement of F .

APPLICABILITY: MODE 1. .

t ACTION: .

f With the combination of Reactor Coolant System total flow rate and R outside the region of acceptable operation shown on Figure 3.2-3:

i

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore the combination of Reactor Coolant System total flow

~

rate and R to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER f and reduce the Power Range Neutron Flux - High Trip Setpoint '

to less than or equal to 55% of RATED. THERMAL POWER within ;

-the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . !

b. - Within-24 hours of initially being outside the above limits, verify through incore flux mapping and Reactor Coolant System total flow ;

rate comparison that the combination of R and Reactor Coolant System total flow rate are restored to within the above limits, or reduce '

THERMAL POWER to less than 5% of RATED THERMAL POWER within the j 7 next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

i CATAWBA - UNIT 1 3/4 2-9 '

JUN 81984 -

6 PENALTIES OF 0.1% FOR UNDETECTED FEEDWATER VENTURI FOUUNG AND MEASUREMENT UNCERTAINTIES OF2.0% FOR FLOW AND 4.0% FOR INCORE MEASUREMENT OF F$H ARE INCLUDED IN THIS FIGURE. ,

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y. .

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e.e ae .ea

-e

..eeg>m >w

. . ee .

~gue.

.. = .e ..'

.." .~.46~e.~..... .

~

.h

.~e~ <

36 0.90 0.95 1.00 1.05 1.1 O R = F$Hl1.49 (1 + 0.3(1-P)

FIGURE 3.2-3 REACTOR COOLANT SYSTEM TOTAL FLOW RATE VERSUS R - FOUR LOOPS IN OPERATION CATAWBA - UNIT 1 3/4 2-10

TABLE 3.3-3 (Continued)

S g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 4

5 2

8 MINIMUM c TOTAL NO. CHANNELS CHANNELS APPLICABLE h FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

8. Auxiliary Feedwater (Continued)

g. Auxiliary Feedwater Suction Pressure-Low ~

1) 1 CAPS 5220, 5221, 6-3/ pump 2/ pump 2/ pump 1,2,3 15 5222

2) 1 CAPS 5230, 5231, 5231 6-3/ pump 2/ pump 2/ pump 1,2,3 15

9. Containment Sump y Recirculation '

a. Automatic Actuation 2 1 2 1,2,3,4 14 T Logic and Actuation U Relays

b. Refueling WaterpS rage Tank Level-Low % 4 2 3 1,2,3,4 16 i

Coincident With Safety Injection See Item 1. above for all Safety Injection initiating functions and requirements.

10. Loss of Power l a. 4 kV Bus Undervoltage- 3/ Bus 2/ Bus 2/ Bus 1,2,3,4 15*

! Loss of Voltage i

b. 4 kV Bus Undervoltage-h

a:

Grid Degraded Voltage 3/ Bus 2/ Bus 2/ Bus 1,2,3,4 15*

$ 11. Control Room Area , .

. Ventilation Operation i

I a. Automatic Actuation Logic ,

j and Actuation Relays 2 1 2 All 24 i

i f ,.. _ . , - - - , - - . - , . , _ . , , , . . . _ _ . . . r__ .. ,_ _ . _ . _ _ _ . . . _ . . -. __._. ___._. . _ _ _ _ . - - _ _ _

TABLE 3.3-4 (Continued) k 5 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR c- FUNCTIONAL UNIT i'i ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE e 6. Turbine Trip

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation N.A. N.A. N.A. H.A.

Logic and Actuation Relays N.A.

c. Steam Generator Water 5.4 2.18 1.5 < 82.4% of Level-High-High (P-14) < 84.2% of narrow iiarrow range range instrument instrument span R span ,

d. Trip of All Main N.A. -

Y N.A. N.A. N.A. N.A.

Feedwater Pumps o" .

e. Doghouse Water Level-High 1. 0 0 0.5 11 inches 12 inches above~

above 577' 577' floor level floor level

f. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

7. Containment Pressure Control System

a. Start Permissive N.A. N.A. N.A A7siil < O'.iS'$1d c_ b. Termination N.A. N.A. N > 0.25 psid > 0.3 psid E .

~

g 8. Auxiliary Feedwater -

j g a. Manual Initiation N.A N.A. N.A. N.A N.A. t

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays ,

.~

f b .

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS .

h '

SENSOR TOTAL ERROR e FUNCTIONAL UNIT ALLOWANCE (TA) Z

$ (S) TRIP SETPOINT ALLOWABLE VALUE

] 8. Auxiliary Feedwater (Continued)

c. Steam Generator Water 15 12.18 1.5 Level - Low-Low 117% of span 1 10.25% of from 0% to span from 0% to 30% RTP 30T RTP increasing increasing linearly to -

i arly to 1 53.2% of span l 5 54. of span from 30% to 100%

rom 30% to RTP 100% RTP

d. Safety Injection See Item 1. above for.all Safety Injection Setpoints and Allowable Values.

e. Loss-of-Offsite Power N.A. N.A. N.A.

w 1 3500 V 1 3200 V J, f. Trip of All Main Feedwater Pu.aps N.A. N.A. N.A. N.A. N.A.

g. Auxiliary Feedwater Suction Pressure-Low

1) 1 CAPS 5220, 5221, 5222 N.A. N.A. N. A.- 3 9.6 psig 1 9.5 psig

2) 1 CAPS 5230, 5231, 5232 N.A. N.A. N.A. 1 10 psig 1 9.9 psig

9. Containment Sump Recirculation i

c_,

a. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. .A. Nb

g y i?A/5 ( m2.1

b. Refueling Water St N.A. N.A.

g irage N.A.

((1300 inches a Tank Level-Low 2 3as- in es Coincident Wi th S fety l Injection See Item 1. above for all Safety Inject nSetpo(intand Allowable Values.

i I ^ _ _. . _ _ _ _ _ - . ___ _

l ..-

TA8'E L 3.3-4 (Continued) 9 g -

ENGINEERED SAFEfY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS E

SENSOR TOTAL ERROR g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE Q

g 13. Annulus Ventilation Operation ,

a. Manual Initiation N.A. N.A. N.A. N.A. N.A.

b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays'

c. Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

14. Nuclear Service Water Operation p ,

$ a. Manual Initiation N.A. N.A. N.A. N.A. N.A. .i

,Y b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. y w and Actuation Relays m

c. Loss-of-Offsite Power N.A. N.A. N.A. > 3500 V > 3200 V

d. Containment Spray See Item 2. above for all Containment Spray Setpoints and Allowable Values.

e. Phase "B" Isolation See Item 3.b. above for all Phase "B" Isolation Setpoints and Allowable Values. '

f. Safety Injection Se tem' . above all S fety j b n Se e hs Al.lowable ses,

g. Suction Transfer-Low Pit Level , _

M ___M --

54

' -- "#!##4

~

15. Emergency Diesel Generator Operation (Diesel Building PAGEMPETiii5 REM Ventilation Operation, Nuclear ,

INFORMATUN ..a)s THE AP*ICANT Service Water Operation)

^-

-=

a. Manual Initiation N.A N.A. N.A. N.A. N.A.

p '

f' 7

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS

, 1. Manual Initiation '

a. Safety Injection (ECCS) N.A.

b. Containment Spray i N.A.

c. I

+

Phase ~"A" Isolation ~ N.A.

d. Phase "B" Isolation '

N.A.

e. Purge and Exhaust Isolation N.A.

f. . Steam Line Isolation N.A.

g. Diesel Building Ventilation Operation N.A.

h. Nuclear Service Water Operation N.A.

1. Turbine Trip N.A.

j. Component Cooling Water N.A.

k'FArmulus Ventilation Operation N.A.

( [ W.

Mt

? t r ^ = "!:.7ti'-+4aa 0 =t': ;

Auxil; ary Building Filtered Ventilation Exhaurt Operation

.1. A c N.A.

gp Reactor Trip N. A.

4 A: Emergency Diesel Generator Operation N.A.

o# Conta neent Air Return and Hydrogen Ski r Operation N.A.

fW Aux iary Feedwater N.A.

2.

l 'Contai nt Pressure-High a./fafety Injection (ECLS) 1 27(1)/12(3)

1) Reactor Trip i2 t

2) Feedwater Isolation <7 i

3) Phase "A" Isolationf) 8( 3/2 (4)

4) Purge and Exhaust Isolation < b

5) Auxiliary Feedwater(5) N. .

6) Nuclear Service Water Operation 1 65(3)/76(4)

7) Turbine Trip N.A.

8) Component Cooling Water 1 65(3)/76(#)

9) Emergency Diesel Generator Operation i 11

10) Control Room Area Ventilation Operation N.A.

CATAWBA - UNIT 1 3/4 3-37 %

s .

~

TA8LE*3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

_ RESPONSE TIME IN SECONDS

2. Containment Pressure-High (Continued)

11) Annulus' Ventilation Operation

_< 23

12) Auxiliary Building Filtered N.A.

Ventilation Exhaust Operation l

13) C6ntainment Sump Recirculation N.A.

, 3. Pressurizer Pressure-Low

a. Safety Injection (ECCS) l 5 27(1)/12(3)

1) Reactor Trip

~<2

2) Feedwater Isolation <7

3) Phase "A" Is,olation(2) O r 18N)

,p 8(4) f

4) Purge and Exhaust Isolation </

5) 'kbiiliary Feedwater(5) N. . '

6)- Nuclear Service Water Operation < 65(3) 76(4)

7) Turbine Trip

8) Component Cooling Water 1 65(3)/76(4)

9) Emergency blesel Generator Operation 1 11 .

10) Contr(. Room Area Ventilation .

Operation

11) Annulus Ventilation Operation _< 23

, 12) Auxiliary Building Filtered N.A.

Ventilation Exhaust Operation -

13) Containment Sump Recirculation N.A. ~I

4. Steam Line Fressure-Low '

a. Safety Injection (ECCS) 1 12(3)/22(4) .

1) Reactor Trip _2

2) Feedwater Isolation <7

3) Phase "A" Isolation (2) < 3)/28(4)

Purge and Exhaust Isolation

4) '

5) Auxiliary Feedwater(5) _

<y .

6) Nuclear Service Water Operation 1 65(3)/76(4) ;

7) Turbine Trip N.A.

8) Component Coolin) Water 1 65(3)/76(4)

9) Emergency Diesel Generator Operation i 11 -

CATAWBA - UNIT 1 3/4 3-38 JUN 221984

.- ,.. - - - ,- -.. - -.,.--,__,_.-__..__,_.w_...__mm, , , , - - - _ ,mo..--._- . . -

TABLE ?.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

4. Steam Line Pressure-Low (Continued)

10) Control Room Area Ventilation N.A.

Operation

11) Annulus Ventilation Operation <_. 23

12) Auxiliary Building Filtered Ventilation Exhaust Isolation N.A.

13) Containment Sump Recirculation N.A.

b. Steam Line Isolation <7

5. Containment Pressure-High-High g

a. Containment Spray < 45 g /

b. Phase "B" Isolation < 'dd /s A[

Nuclear Service Water Operation N. y h ej

^

c. Steam Line Isolation 57

d. Contcinment Air Return and Hydrogen < 600 8k

( Skimmer Operation cc c_

e< ,

6. Steam Line Pressure - Negative Rate-High 9,

e-- '. '.d-Steam Line Isolation 57 7

- i to 5

7. Steam Generator Water Level-High-High I,

a. Turbine Trip 13 , t,; ' .f .

I

b. Feedwater Isolation <7

- O2 to w c.D <

8. T -Low <2 -

avg Q- ce Feedwater Isolation N.A.

Q@

-Z

9. Doghouse Water Level-High *-

a. Feedwater Isolation N.A.

b. Turbine Trip N.A.

10. Start Permissive Containment Pressure Control System N.A.

11. Termination -

Containment Pressure Control System N.A.

L CATAWBA - UNIT 1 3/4 3-39 Juli 8194'

........-.......-n.-.-7.-....--,- ....----.,..,..,......... .

r - . - - . .

\.

) '

i.

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES

' INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN' SECONDS -

12.

Steam Generator Water Level-Low-Low

a. Motor-Driven Auxiliary Feedwater Pumps

< 60

b. Turbine-Driven Auxiliary Feedwater Pump i 60

13. Loss-of-Offsite Power ' ' '

N.A.

a. Motor-Driven Auxillary Feedwater Pumps < 60

b. Turbine-Driven Auxiliary Feedwater Pumps 1 60

c. Control Room Area Ventilation Operation N.A.

d. Emergency Diesel Generator Operation i 11 w

1) Diesel Building Venti 1'ation Operation N.A. O H

2) Nuclear Service Water Operation Z e.

5 65(3)/76(4) _<

Auxiliary Building Filtered Ventilation ua O Exhaust Operation o ~._a N.A. $g

14. Trip of All Main Feedwater Pumps en <

a.

= tu Motor-Driven Auxiliary Feedwater Pumps c

b. Turbine Trip 1 60 C t-

15. Auxiliary Feedwater Suction Pressure-Low N.A.

i kh T.

{ uT3 ' '--

Auxiliary Feedwater (Suction Supply 5 15(6) f g {g Automatic Realige. ment) -

i.o

16. f c.D <

RefuelingWaterStorageTankLevel-Lo$

3 Coincident with Safety Injection Signal

h e.,

(Automatic Switchover to Containment 52 @

Sump) Cz

"~

$ 60 -

17. Loss of Power i a. 4 kV Bus Undervoltage -

Loss of Voltage 5[ -

~

b. 4 kV Bus Undervoltage-Grid Degraded Voltage $g [o00

18. Suction Transfer-Low Pit Level '

Nuclear Service Water Operation N.A.

CATAWBA - UNIT 1 3/4 3-40 J uel 8 198 k

. , w-.,. . . v c. ... i

o TA8LE 4.3-2 (Continued) -

O ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

, SURVEILLANCE REQUIREMENTS TRIP g ANALOG ACTUATING q CHANNEL DEVICE MODES CHANNEL MASTER SLAVE FOR WHICN g CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY FUNCTIONAL UNIT CHECK RELAY SURVEILLANCE CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

8. Auxiliary Feedwater

a. Manual Initiation N.A. N.A. , M.A. R N.A. N.A. N.A. 1,2,3

b. Automatic Actuation N.A. M.A N.A. N.A.

Logic and Actuation M(1) M(1) Q 1,2,3 Relays

, c. Steam Generator S R M N.A. N.A. M. A ' N. A 1,* 2, 3 g Water Level-tow-Low .

y d. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

$ e. ' Loss-of-Offsi te N.A. R N.A. M N.A. N.A. N.A 1,2,3 Power

f. Trip of All Main N.A. N.A. N.A. R N.A. M.A. N.A 1, 2 Feedwater Pumps

g. Auxiliary Feedwater Suction Pressure-Low S R H N:A. N.A. N.A. N.A. 1,2,3

9. Containment Sump Recirculation

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

Logic and Actuation M(1) M(1) Q 1,2,3,4 h Relays a b. Refue ing Water S .R M N.A. N.A. N.A.

88 St N.A. 1,2,3,4 ge nk Level -

g L M C incident With 5 fety jection '

See Item 1. above for all Safety Injection Surveillance Requirements.

s A

~

.~

TABLE 3.3-10 9

g ACCIDENT MONITORING INSTRUMENTATION 5

-TOTAL MINIMUM NO. OF CHANNELS

'E INSTRUMENT CHANNELS OPERABLE

--e g 1. Containment' Pressure 2 . 1

~

2.

3.

Reactor Coolant Oatlet Temperature - THOT (Wide Range) 2 1 Reactor Coolant Inlet' Temperature - TCOLD (Wide Range) 2 -

1

4. Reactor Coolant Pressure - Wide Range 2 1

-r

--- t

5. Pressurizer Water Level 2 1 w 6. Steam Line Pressure 2/ steam generator 1/ steam generator 1

i w 7. Steam Generator Water Level - Narrow Range 2/ steam generator 1/ steam generator

! E j 8. Refueling Water Storage Tank Water Level 2 1

9. Auxiliary Feedwater Flow Rate 2/ steam generator 1/ steam generator

10. Reac r Coolan System Subcooling Margin Monitor 1 1 -

%:N.4 I

11. P RV-F4ew I cator* 2/ Valve 1/ Valve i '

12. PORV Block Valve Position Indicator ** 2/ Valve 1/ Valve ,

. 13. Pressurizer Safety Valve Position Indicator lve 1/ Valve

14. Containment Sump Water Level (Wide Range) 2 1

INSTRUMENTATION e u FIRE DETECTION INSTRUMENTATION h LIMITING CONDITION FOR OPERATION '

3.3.3.8 As a minimum, the fire detection instrumentation for each fire Qr y detection zone shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY: Whenever equipment protected by the fire detection instrument 1 , M is required to.be OPERABLE. _ _ .. . _ ._ .. _

N 3 Q l ACTION:

$ E

a. With any, but not more than one-half the total in any fire zone, 6

Function A fire detection instruments ~shown in Table 3.3-11 inoperable, yR g h E. restore the inoperable instrument (s) to OPERABLE status within 14 days o g or within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the zone (s)

G x with the inoperable instrument (s) at least once per hour, unless the w tr 3 instrument (s) is located inside the containment, then inspect that b % a containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor the containment

-< * ? Ii air temperature at least once per hour at the locations listed in M%EJ Specification 4.6.1.5.

I b. With more than one-half of the Function A fire detection instruments ty J ps in any fire zone shown in Table 3.3-11 inoperable, or with any Function B fire detection instruments shown in Table 3.3-11 inoperable,

'd Isa)'

Qwi or with any two or more adjacent fire detection instruments shown in Table 3.3-11 inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol

.$ g c* to inspect the zone (s) with the inoperable instrument (s) at least T h) da once per hour, unless the instrument (s) is located inside the contain-ment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months M }i or monitor the containment air temperature at least once per hour

%a e at the locations listed in Specification 4.6.1.5.

(jf c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, wh C N s cr SURVEILLANCE REQL'IREMENTS

).yg k$ I .3.3.8.N.1Each of the above required fixed temperature / rate of rise detection 4

4

,3 instruments shall be demonstrated OPERABLE as follows:

a. For nonrestorable spot-type detectors, at least two detectors out of E.

w ) kk Ib every hundred, or fraction thereof, shall be removed every 5 years Jy j> and functionally tested. For each failure that occurs on the detectors 3} removed, two additional detectors shall be removed and tested; and J I b. For restorable spot-typo heat detectors, at least one detector on

+ g each si 0nal initiating circuit shall be demonstrated OPERABLE at, j)j

't o

@ }a ig jA least once per 6 ' months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST.

test.

Different detectors shall be selected for each Fire detectors which are not accessible during plant operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING

{ DEVICE OPERATIONAL TEST during each COLD SHUTOOWN exceeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless performed in the previous 6 months.

" ,4g 4.3.3.8.X' The NFPA Standard 72D supervised circuits supervision associated F. m M

with the detector alarms of each of the above required fire detection instruments

, J I , shall be demonstrated OPERABLE at least once per 6 months.

tr-ww5 CATAWBA - UNIT 1 3/4 3-71

.t .

TABLE 3.3-11 k FIRE DETECTION INSTRUMENTS I

E l g MINIMUM INSTRUMENTS OPERABLE

g FIRE I

q ZONE DESCRIPTION LOCATION ,

SM0KE FLAME HEAT FUNCTION **

1 R.H.R. Pump IB GG-53 E1.522 + 0 2 1 0 1 A R.H.R. Pump 1A FF-53 E1.522 + 0 1 0 1 A 3 Cont. Spray Pump IB GG-54 E1.522 + 0 3 4 Cont. Spray Pump 1A 0 3 A .-

9 GG-55 E1.522 + 0 2 0 2 A Aux. F. W. Pumps BB-51 E1.543 + 0 13 i

10 0 11(6) A(B)

Mech. Pene. Room JJ-52 E1.543 + 0 3 11 0 3 A Corridor / Cables NN-51 E1.543 + 0 6 12 0 6 A Recip. Chg. Pump JJ-53 E1.543 + 0 1 0 1 A 13 Safety Inj Pump IB HH-53 E1.543 + 0 1 0

, 14 Safety Inj Pump 1A 1 A ,

g 15 GG-53 E1.543 + 0 1 0 1 A Cent. Chg. Pump 18 JJ-54 E1.543 + 0 2 I

16 0 2 A

, Cent. Chg. Pump 1A JJ-55 E1.543 + 0 2 0 2 A 4 17 Aisles / Cables KK-56 E1.543 + 0 18 0 18 A h> 18 Aisles / Cables EE-55 j

E1.543 + 0 6 0 6 A 21 Aisles / Cables I NN-61 E1.543 + 0 6 27 Aisles / Cables KK-59 E1.543 + 0 15 0

0 6 A O

28 Aisles / Cables EE-58 E1.543 + 0 15 A y 29 30 SW Gear Equip. Room Elect. Pene. Room AA-50 E1.560 + 0 1

7 0 0

0 1 A A

- r":

31 CC-50 E1.560 + 0 8 0 0 A Corridor / Cables EE-53 E1.560 + 0 5 0 32 5 A Corridor / Cables KK-52 E1.560 + 0 8 0 33 8 A Corridor / Cables NN-54 E1.560 + 0 10 0 34 10 A Aisles / Cables JJ-56 E1.560 + 0 14 0

',~

35 14 A Motor Control Centers GG-56 ' E1.560 + 0 2 0 2 36 Cable Tray Access A FF-56 E1.568 + 0 2 0 2 A 37 Equip. Batteries , DD-55 E1.554 + 0 5 0 4 A 38 Equip. Batteries CC-55 E1.554 + 0 5 0 t

39 Battery Room 4 A ..

CC-56 E1.554 + 0 17 0 0 A !

45 Aisles / Cables NN-60 E1.560 + 0 1 0 1 A 46 Aisles / Cables HH-59 E1.560 + 0 0 E A V

. ... . . . . ..... . . .- . a. . . . .. . a ... ,. . .a u . , ..

.1 l.

I TABLE 3.3-11 (Continued) l FIRE DETECTION INSTRUMENTS MINIMUM INSTRUMENTS OPERABLE

l t

E FIRE ZONE DESCRIPTION LOCATION SMOKE FLAME HEAT FUNCTION ** i 53 SW Gear Equip. Room AA-49 E1.577 + 0 7 0 0 A 54 Aisles / Cables CC-50 E1.577 + 0 10 0 0

{

A i 55 Aisles / Cables NN-52 E1.577 + 0 9 0 9 A I 56 Afsles/ Cables -PP-55 E1.577 + 0 13 0 13 A l 57 Aisles / Cables LL-55 E1.577 + 0 11 0 11 A l 58 Aisles / Cables HH-55 E1.577 + 0 21 0 21 A 59 Motor Control Center EE-54 E1.577 + 0 2 0 2 A ;

60 Cable Room CC-56 E1.574 + 0 18 0 15 A ,

65 Aisles / Cables PP-59 E1.577 + 0 15 0 15 '

A w 66 Aisles / Cables LL-59 E1.577 + 0 4 0 4 A D 71 Elect Pene. Room CC-51 E1.594 + 0 10 0 0 A i w 72 Control Room CC-56 E1.594 + 0 23 0 6 A 4

73 Vent. Equip. Room FF-56 E1.594 + 0 9 0 0 A . i 74 Aisles / Cables LL-56 E1.594 + 0 25 0 25 A 75 Aisles / Cables PP-54 E1.594 + 0 15 0 15 A 80 Control Room 81 Ven. Equip. Room BB-59 FF-58 E1.594 + 0 E1.594 + 0 22 0 0

6 A O 82 Aisles / Cables KK-58 E1.594 + 0 0 A Z 84 Aisles / Cables NN-58 E1.594 + 0 0 A b 89 Fuel Pool Area #1 A y PP-50 E1.605 + 10 7 A q l

-128 UHI Bldg. HH-44 E1.550 + 0 2 3 2 A 129 Fuel Pool Purge Room NN-50 E1.631 + 6 6 0 6 A

' 131 Reactor B1dg. Be1.0"-45* E1.565 + 3 4 0 0 A 132 Reactor Bldg. Be1.45*-90 E1.565 + 3 3 0 0 A 133 Reactor Bldg. Be1.90 -135 E1.565 + 3 4 0 0 A 134 Reactor Bldg. Be1.135*-180 E1.565 + 3 5 0 0 A i 135 Reactor Bldg. Be1.180*-225 E1.565 + 3 4 0 0 A 136 Reactor Bldg. Be1.270*-315* E1.565 + 3 3 0 0 A 137 Reactor Bldg. Be1.315*-0* E1.565 + 3 8 0 0 A 138 Reactor Bldg. Be1.0 -45* E1.586 + 3 6 0 0 A 139 Reactor B1dg. Be1.45 -90 E1.586 + 3 4 0 0 A ,

140 Reactor Bldg. Bel.90*-135* E1.565 + 3 3 0 0 A 5 141 Reactor Bldg. Be1.135*-180 E1.586 + 3 8 0 0 A

TABLE 3.3-11 (Continued) 9 g FIRE DETECTION INSTRUMENTS E

MINIMUM INSTRUMENTS OPERABLE

e C

5

-4 FIRE ZONE DESCRIPTION LOCATION SM0KE FLAME HEAT FUNCTION **

w 150 e. i uy+.. u.'+-- Exch: ;;r 50 00, 000 0- 1(Duct)' O O A

--155 -o s ;.. On":y729 01 ::':, 5 0 1(Duct) 0 0 A

": u;g ;;;,_ a;;;g$n;7

_____c_

u..s.

e_

. .u 3 . _ . ,

150 L. lu Au;u T. ...J: . 50 pp, 500:0-- 1(Duct) 0 0 A

" .c;; !?., 19, ?? *. 29 RF1A Diesel Generator IA EE-41, 556 + 0 0 0 0(10) A(B)

RFIB Diesel Generator 18 AA-41, 556 + 0 0 0 0(10 A(B)

N.

Y' u

j ui

The fire detection instruments located within the containme'nt are not required to be OPERABLE during the performance of Type A Containment Leakage Rate tests.

- Function A: Early warning fire detection and notification only.

Function B: Actuation of fire suppression system and early warning and notification.

j tgf %VM.'. 'bwd foe Lc.s FF - 53, 545 +o 31I Q 312.

\$5 Q)g_. %{ hr $cce1S M M '" G'O > 5 3 f O ZoS , zoS ,7. oSA , 20GA , 7d*S, 1 E 207 W 2074 i z

, oc l8(p lb/AC hch br Ecokts NN'4 '

7 E

J 30s, 3o2 > 3o5 ) ~ 30 7

REACTOR COOLANT SYSTEM HOT STANDBY I LIMITING CONDITION FOR OPERATION !

ree 3.4.1.2 At lea the reactor coolant loops listed below shall be OPERABLE and at leas ene' f these reactor coolant loops shall be in operation:*

iwo .

a. Reactor Coo ant Loop A and its associated steam generator and reactor coolant pump,

b. Reacto'r Coolant Loop 8 and its associated steam generator and reactor coolant pump,

c. Reactor Coolant Loop C and its associated steam generator and reactor i a

coolant pump, and

d. Reactor Coolant Loop D and its associated steam generator and reactor coolant pump.

APPLICABILITY: MODE 3.

ACTION:

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

L

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

required reactor coolant loop to operation.

SURVEILLANCE REQUIREMENTS '

h r

4.4.1.2.1 At' least the above required reactor coolant pumps, if not in operation, shall be determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

, 4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 12% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

! 4.4.1.2.3 At least one reactor coolant loop shall be verified 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 .

All reactor coolant pumps may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided:

(1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.

CATAWBA - UNIT 1 3/4 4-2

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

. - . - - , , - . _ _ , , . , - - - _ .. - , - . - -, . - - - . - , _ . , - - - , _ _ _ _ - , . , . - , _ - , . - - -, -_-_-7--

_ ,, _ ...-e- --~ - . . - ~ - - . ~ ~~ -- * -* ---~ ~~---- no.-e-o..' - - - -

h' .

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK DRAT i

~

LIMITING CONDITION FOR OPERATION 3.5.4 The refueling water storage tank shall be OPJRA

a. %3513 Aminimumcontainedboratedwatervolume((460,499 7 l 3 ons,

b. A boron concentration of between 2000 and 2100 ppm of boron, i

c. A minimum solution temperature of 70*F, and

d. A maximum solution temperature of 100*F.

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

ACTION:

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

SURVEILLANCE REQUIREMENTS 9

4.5.4 The refueling water storage tank shall be demonstrated OPERABLE:

a. At least once per 7 days by:

1) Verifying the contained borated water level in the tank, and

2) Verifying the boron concentration of the water.

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the refueling water storage tank temperature when the outside air temperature is less than 70*F or greater than 100*F.

CATAWBA - UNIT 1 3/4 5-11 O

. . . . . . . . . - - . .. - --....-- - . . . . . . . - - - - - - . ---~: -~-~~~~.------=----~~~-

- - ~ ~ - - - -

TABLE 3.6-1 (Continued)

S g SECONDARY CONTAIMcENT BYPASS LEAKAGE' PATHS c

E PENETRATION s

TEST NUPEER SERVICE RELEASE LOCATION c TYPE

(

M386 Containment Air Release Auxiliary Building Type C i

M204 Containment Air Addition Auxiliary Building Type C

! M316 Int. Fire Protection Header - Auxiliary Building Type C Hose Racks M337 Demineralized Water Auxiliary Building Type C M220 Instrument Air Auxiliary Buf1 ding Type C M219 Station Air Auxiliary Building s Type C

[ M215 Breathing Air Auxiliary Building Type C !

M329 Reactor Coolant Pump Motor Oil Fill Auxiliary Building Type C i t

M361 Int. Fire Protection Header - Auxiliary Building Type C i Sprinklers

. M119 Containment Purge Exhaust Auxiliary Building Type C g '

r1311

~

Z phre3ea SgIy -h 6Q L g,;f;,y Bu;/ Jog -Typa C-- p

Acca,nul./or.s T !

f1321 $[eh Ljec/no Tic / hae. I"

'47 N

. Msi uut %{ h,a .Auch~y O '") W t

i

+

A

CONTAINMENT SYSTEMS

/ P4RGE CONTAINMENT VGiT:LATI^% SYST,EM f -

w -

LIMITING CONDITION FOR OPERATION .

3.6.1.9 Each containment purge supply and exhaust isolation valve shall be OPERABLE and:

a. Each containment purge supply and/or exhaust isolation valve for the lower compartment (24-inch), instrument room (12-inch), and the Hydrogen Purge System (4-inch) shall be sealed closed, ;

b. The containment purge supply and/or exhaust isolation valve (s) for

'the upper compartment (24-inch) may be open for up to 250 hours0.00289 days 0.0694 hours 4.133598e-4 weeks 9.5125e-5 months during a calendar year provided no more than two penetrations are open at one time, and ,

c. The Containment Air Release and Addition System (4-inch) isolation valve (s) may be open for up to 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months during a calendar year.

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

ACTION:

a.

With any containment purge supply and/or exhaust isolation valve for the lower compartment, or instrument room, or Hydrogen Purge System

. open or not sealed closed, close and/or seal closed that valve or '

. isolate the penetrations (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

b. With the containment purge supply and/or exhaust isolation valve (s) for the upper compartment open for more than 250 hours0.00289 days 0.0694 hours 4.133598e-4 weeks 9.5125e-5 months during a calendar year, close the open valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

c. With the Containment Air Release and Addition System isolation valve (s) open for more than 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months during a calendar year, close the open valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

d. With'a containment purge supply and/or exhaust isolation valve (s) having a measured leakage rate in excess of the limits of Specifi-cations 4.6.1.9.3 and/or 4.6.1.9.4, restore the inoperable valve (s) to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , otherwise be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

O CATAWBA - UNIT 1 3/4 6-16

( . i i

CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS -

l CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2 Two independent Containment Spray Systems shall be OPERABLE with each

~ Spray System capable of taking suction from the refueling water storage tank and transferring suction to the containment sump.

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

ACTION:

With one Containment Spray System inoperable, restore the inoperable Spray System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the inoperable Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMEh.S 4.6.2 Each Containment Spray System shall be demonstrated OPERABLE:

a. At least once per 31 days by 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. Bf'ved fyl g that on recirculation flow, each pump d vele a ,

rea4, - - '#d hcharge pres ure of greater than or equal to 185 sigwhen

}; tested pursuan to Specification 4.0.5; (g ss __

c. At least once per 18 months during shutdown, by:

i

1) Verifying that each automatic valve in the flow path actuates to its correct position on a Phase "B" Isolation test signal, and

2) Verifying that each spray pump starts automatically on a Phase "B" Isolation test signal.

3) starting by Verifying that each the Containment spray Pressure Control pump System is prevented whe he' from ,& ntainment i atmosphere pressure is less than or equal o 0.26 sid, t 4 4

i CATAWBA - UNIT 1 3/4 6-18 4

y , , . , _ _ _ - --

.,y-_.. - . _ _ _ _ _ __ _ _ -,..--,_--.-_-----._m, -

-. - = - - - - - ;

L

?,

CONTAINMENT SYSTEMS i.

(' SURVEILLANCE REQUIREMENTS (Continued) l

~.9

4) Verifying that each spray pump discharge valve is prevented

.: fr ,.

opening by the Containment Pressure Control System when

. a c ntainment atmosphere pressure is less than or equal to j . $ sid, and

i r

! 5) rifying that each spray pump is automatically deenergized

~; . by the Containment Pressure Control System when the containment at phere pressure is reduced to less than or equal to l . psid.

d. At ea once per 5 years by performing an air or smoke flow test ;

through each spray header and verifying each spray nozzle is '

unobstructed.

?

i l

l

'~

j t l

l 1

a l

l\i i

l

[ CATAWBA - UNIT 1 3/4 6-19 F

l' so CONTAINMENT SYSTEMS, j

q

SURVEILLANCE REQUIREMENTS (Continued) -

't 4.6.3.2 Each isolation valve specified in Table 3.6-2 shall be demonstrated 4' OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:

i 3

a. Verifying that on a Phase "A" Isolation test signal, each Phase "A" isolation valve actuates to its isolation position;

b. Verifying that on a Phase "B" Isolation test signal, each Phase "B" isolation valve actuates to its isolation position; r

c. Verifying that on a Containment Radioactivity-High test signal, each purge and exhaust valve actuates to its isolation position; and

d. Verifying that o i h Relative Humidity (>,70%) isolation test eig-e! N h pu

. post fg f ly eandehau]stvalveactuatestoitsisolation M The isolation time of each power-operated or automatic valve of Table 3.6-2 shall be determined to be within its limit when tested pursuant to peefff,= tion 4 0.5.

(Age t and \ower con $0/nnoen

_l P

P CATAWBA - UNIT 1 3/4 6-21

.1 L.

o * . [r i

i n TABLE 3.6-2 (Continuedl !,

g CONTAINMENT ISOLATION VALVES '

t

$ l.

8 MAXIMUM g VALVE NUMBER FUNCTION ISOLATION TIME (s)

G '

r 1. Phase "A" Isolation (Continued) '

NI-266A UHI Check Valve Test Line Inside Containment Isolation <10 i NI-267A ;JHI Check Valve Test Line Inside Containment Isolation 210 NI-153A# Hot Leg Injection Check NI156, NI159 Test Isolation 210 t

NM-3A Pressurizer Liquid Sample Line Inside Containment Isolation $10 >

NM-6A Pressuri7er-S Sample Line Inside Containment Isolation $10 NM-78 surizer Header Outside Containment Isolation <10 '

NM-22A NC Hot Leg A Samp Line Inside Containment Isolation 210 w NM-25A g " NC Hot L M S Line Inside Containment Isolation 310 .

g NM-26B NC Hot e Hdr Outside Containment Isolation $10 *

, NM-72B TG in.cumulator IA Sample Line Inside Containment Isolation $10 4 NH-75B NI Accumulator IB Sample Line Inside Containment Isolation $10

NM-788 NI Accumulator IC Sample Line Inside Containment Isolation $10 l NM-81B NI Accumulator ID Sample Line Inside Containment Isolation <10 '

'. NM-82A NI Accumulator Sample Hdr Outside Containment Isolation $10 C

NM-187A# SG 1A Upper Shell Sample Containment Isolation Inside $10 y 3 l

NM-190A# SG 1A Blowdown Line Sample Containment Isolation Inside 510 p HM-1918# SG 1A Sample Hdr Containment Isolation Outside $10 m i NM-1978# SG IB Upper Shell Sample Containment Isolation Inside $10 m NM-200B# SG IB Blowdown Line Sample Containment Isolation Inside <10 NM-201A# SG IB Sample Hdr Containment Isolation Outside 710 l NM-207A# SG IC Upper Shell Sample Containment Isolation Inside 10

' 5 NM-210A# SG IC Blowdown Line Sample Containment Isolation Inside {10

, NM-2118# SG IC Sample Hdr Containment Isolation Outside 310 i l' NM-2178# SG ID Upper Shell Sample Containment Isolation Inside $10 ;

NM-220B# SG ID Blowdown Line Sample Containment Isolation Inside <10 i NM-221A# SG ID Sample Hdr Containment Isolation Outside <

_10 l NV-15B Letdown Containment Isolation Outside <10 i NV-89A NC Pumps Seal Return Containment Isolation Inside $10 NV-918 NC Pumps Seal Return Containment Isolation Outside <10 l NV-3148# Charging Line Containment Isolation Outside 510 .

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

6 CONTAINMENT SYSTEMS 3/4.6.5 ICE CONDENSER ICE BED LIMITING CONDITION FOR OPERATION 3.6.5.1 The ice bed shall be OPERABLE with:

a. The s.tored ice having~a boron' concentration of at least 1800 ppm boron as sodium tetraborate and a pH'of 9.0 to 9.5,

b. Flow channels through the ice condenser,

c. A maximum ice bed temperature of less than or equal to 27 F,

d. A total ice weight of at least 2,368,652 pounds at a 95% level of confidence, and

e. 1944 ice baskets. , ,

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

ACTION:

With the ice bed inoperable, res' tore the ice bed 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 STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUT-DOWN within.the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.5.1 The ice condenser shall be determined OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by using the Ice Bed Temperature Monitor-ing System to verify that the maximum ice bed temperature is less than or equal to 27 F,

b. At least once per 9 months by:

1) Chemical analyses which verify that at least nine represen_t ive samples of stored ice have a boron concentration of at 1

.1800 ppm as sodium tetraborate and a pH of 9.0 to 9.5 C;

2) Weighing a representative sample of at least 144 ice ba and verifying that each basket contains at least 1218 lbs of ice. The representative sample shall include six baskets from each of the 24 ice condenser bays and shall be constituted of 6

CATAWBA - UNIT 1 3/4 6-33

i

?

{ PLANT SYSTEMS

l 3/4.7.3 COMPONENT COOLING WATER SYSTEM

,1 nD A T Ul\n . I 1 '

{ LIMITING CONDITION FOR OPERATION ~

i 3.7.3 At least two independent component cooling water loops shall be OPERABLE.

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

ACTION: .

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .-

SURVEILLANCE REQUIREMENTS 4.7.3 At least two cemponent cooling water loops shall be demonstrated- OPERABLE

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or auto.matic) servicing safety-related equipment that is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At least once per 18 months during shutdown,,-Wrifying that:

1) Each automatic valve servicing safe y-related equip ant wtuatas to its correct position on a Safot Injection test signal,Nand

2) Each Component Cooling Water.Systen pump starts automatically on a Safety Injection test signal.

' ', Pl.m *A' ol* b'* *r fh,u "6 " isoleS'*4 9

\ %

+

CATAWBA - UNIT.1 3/4 7-10

t PLANT SYSTEMS r ;

3/4.7.7 AUXILIARY BUILDING FILT RED *Y MT!LATI0s EXHAUST SYSTEM

( /

LIMITING CONDITION FOR OPERATION -

3.7.7 The Auxiliary Building Filter V;c/_f!;t: m Exhaust System shall be OPERA 8LE.

APPLICA8ILITY: MODES 1, 2, 3, and 4.

ACTION: .

With the Auxiliary Building Filtered Y;nti!;tt;; E haust System inoperable, restore the inoperable system to OPER LE <te__ within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

f SURVEILLANCE REQUIREMENTS P

4.7.7 The Auxiliary Building Filte d Ven d&eMon E aust System shall be demonstrated OPERA 8LE:

a. At least once per 31 days by initiating, from the control room, flow through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating; f

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release'in any ventilation zone communicating with the system by:

i l: 1) Verifying that the cleanup system satisfies the in place L? penetration and bypass leakage testing acceptance criteria of less than 1% and uses the test procedure guidance in Regula-tory Positions C.S.a. C.S.c and C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 30,000 cfm t -

i 10%;

h:-

2). Verifying, within 31 days after removal, that a laboratory -

E analysis of a representative carbon sample obtained in accor-dance with Regulatory Position C.6.b of Regulatory Guide 1.52, i Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revi-l sfon 2, March 1978, for a methyl iodide penetration of less i than 1%; and CATAWBA - UNIT 1 3/4 7-16 4

6 *

94 9& -+

^

PLANT SYSTEMS ' '

SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying a system flow, rate of 30,000 cfm 110% during system

, operation when tested in accordance with ANSI N510-1980.

c.

After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of. Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%;

d. At least once per 13 months by:

1) Verifying that the pressure drop across the combined HEPA

~

filters, charcoal adsorber banks ,and c moisture separators of less than 8 inches Water Gauge while c,perating the system at a flow rate of 30,000 cfm i 10%,

/r M m

2) erffying that the syste3 st' arts on a Safety Injectio 1^ee cf-Offaii.e Funrr test signal, and directs its ex, t ow through__the-HEPA filters and charcoal adsorbers,

3) Verifying that the system maintains the ECCS pupp-room 5t a negative pressure relative to the eutride at: .,.hre,adjacenf

^

amtS

4) Verifying that the filter cooling by asm ves can be manuai sy opened, and

,' 40 k 4 -

5) Verifying that the. heaters diss ate-30 p when tested in accordance with ANSI N510-1980.

e. After each complete or partial replacement of a HEPA filter ba'nk, i

- by verifying that the cleanup system satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than

'4

1% in accordance with ANSI N510-1980 for a DOP test aerosol while

> operating the system at a flow rate of 30,000 cfm i 10%; and l f.

After each complete or partial replacement of a charcoal adsorber Le bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less j than 1% in accordance with ANSI N510-1980 for a halogenated hydro-l carbon refrigerant, test gas while operating the system at e flow

, rate of 30,000 cfm i 10%.

CATAWBA - UNIT 1 3/4 7-17

l i .

PLANT SYSTEMS 3 T l

"[

SPRAY AND/OR SPRINKLER SYSTEMS i

LIMITING CONDITION FOR OPERATION [

3.7.10.2 The following Spray and/or Sprinkler Systems shall be OPERABLE:

a. Elevation 522 + 0 ft - Auxiliary Building Room No. Equipment 100 RHR & Containment S ump P 101 Corridor 104 RHR Pump 18 105 RHR Pump 1A Corridor

-ttt- 112. Corridor

b. Elevation 543 + 0 ft - Auxiliary Building 230 Cent. Chg. Pump 1A 231 Cent. Chg. Pump 18 250 Unit 1 Aux. Feedwater Pump Room Elevation 554 + 0'ft - Auxiliary Building

~~

c.

350 Battery Room Corridor (DD-EE)

d. Elevation 560 + 0 ft - Auxiliary Building 300 Component Cooling Pumps 1A1, 1A2, 181 &lB2

e. Elevation 574 + 0 ft - Auxiliary Building 490 Cable Room Corridor (DD-EE)

f. Reactor Building Annulus Pipe Corridor APPLICABILITY: Whenever equipment protected by the Spray / Sprinkler System is required to be OPERABLE.

ACTION:

a. With one or more of the above required Spray and/or Sprinkler Systems inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged; for other areas, establish an hourly fire watch patrol.

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

CATAWBA - UNIT 1 3/4 7-28 -

A

fraf sed 6 Iedhc f /* M MfEi! SEPA.% RED if:0.'.1 E?,LCD.: iEs M:XE Tl!!3 tocu*.st;;7 ps c5co,ycty_gj SURVEILLANCE REQUIREMENTS (Continued)

2. Verifyin e-battery-to-batt y and-termitial j

co ne ions are clean, tight ree of corr,osiop, e d :::ted "ith r,ti : rr::i r,, n tri A jl , ;

4.7.14.2 The St'a ndby Makeup Pump water supply shall be demonstrated OPERABLE by:

a. Verifying at least once per 7 days:

1. .That the requirements of Specification 3.9.10 are met, or 1

2. That a contained water volume of at least 112,320 gallons is available and capable of being aligned to the Standby Makeup Pump.

i

b. Verifying on a quarterly basis that the Stand y Makeup' Pump develops a flow of greater than or equal o .~ gp at a pressure greater than or equal to 2488 psi 26 h

t CATAWBA - UNIT 1 3/4 7-42

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

_ _ . .- ,__.. _ . . _ _ _ _ - _ . _ - , - - - - - - - - - - - - . -- - - - - - - - - + - -

1 DRATi j ELECTRICAL POWER SYSTEMS

j. t

+

q SURVEILLANCE REQUIREMENTS (Continued) ;

b i

2) Verifying the fuel level in the fuel storage tank, i n

i 3) Verifying the fuel transfer valve can be operated to allow i

L I fuel to be transfer ed from the storage system to the day tank, i W/ f

.. 4) Verifying t Ties starts from ambient condition and accelerates ;

i to at leas M4 m in less than or equal to 11 seconds. >

l The gener oltage and frequency shall be 4160 + 420 volts !

.l . and 60 + 1.2 Hz within 11 seconds after the start signal.

] The diesel generator shall be started for this test by using t,

one of the following signals: ,

i i

'i

j a

a) Manual, or l

b) Simulated loss of offsite power by itself, or I l

c) Simulated loss of offsite power in conjunction with an ESF

Actuation test signal, or ,

, i

d) An ESF Actuation test signal by itself. ,

, 5) Verifying the generator is synchronized, loaded to greater than or equal to 7000 kW in less than or equal to 60 seconds, and operates for at least 60 minutes, and l

6) Verifying the diesel generator is aligned to provide standby power to the associated emergency busses.

N

b. At least once per 31 days and after each operation of the diesel !

where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the day tank; i l c. At least once per 31 days by checking for and removing accumulated

! water from the fuel oil storage tanks; f', ,

Ir d. By verifying that the Cathodic Protection System is OPERABLE by ,

verifying: '

,. a

i- 1) At least once per 60 days that cathodic protection rectifiers l l .. are OPERABLE and have been inspected in accordance with the y manufacturer's inspection procedures, and ir t M 2) At least once per 12 months that adequate protection from j p corrosion is provided in accordance with manufacturer's ;

/ inspection procedures. '

, e. By sampling new fuel oil in accordance with ASTM-04057 prior to ;

L addition to storage tanks and: i

?*

CATAWBA - UNIT'l 3/4 8-3

,=g_..-.__-.

,- 7; , _~y------- - - - - - - - ; ; - - ;- y - 3- -~~; ~~ ; ~ 7

, -v ELECTRICAL POWER SYSTEMS IL i SURVEILLANCE REQUIREMENTS (Continued) h t 8 1) By verifying in accordance with the tests specified in

~*

-t- 4 ASTM-0975-81 prior to addition to the storage tanks that the sample has:

d m_ a) An API Gravity of within 0.3 degrees at 60 F, or a specific

S gravity of within 0.0016 at 60/60 F, when compared to the W

supplier's certificate, or an absolute specific gravity at y')-3

' 60/60*F of greater than or equal to 0.83 but less than or equal to 0.89, or an API gravity of greater than or equal y to 27 degrees but less than or equal to 39 degrees;

$A -

b) A kinematic viscosity at 40 C of greater than or equal to

'85 1.9 centistokes, but less than or equal to 4.1 centistokes, if gravity was not determined by comparison with the JP supplier's certification; E' L M

c) A flash point equal to or greater than 125*F; and fQ

[ fi gz t.

e T

d) A clear and bright appearance with proper color when 4 te.sted in accordance with ASTM-D4176-82.

E c A

-: E-

2) By verifying within 30 days of obtaining the sample that t. 1

@ ]p other properties specified in Table 1 of ASTM-0975-81 are met when tested in accordance with ASTM-0975-81 except that the analysis for sulfur may be performed in accordance with ASTM-01552-79 or ASTM-D2622-82.

f. At least once every 31 days by obtaining a sample of fuel oil in accordance with ASTM-D2276-78, and verifying that total particulate contamination is less than 10 mg/ liter when checked in accordance with ASTM-D2276-78, Method A;

g. At least once per 18 months, during shutdown, by:

1) Subjecting the diesel to an inspection in accordance with procedures prepared in conjunction with its. manufacturer's recommendations for this class of standby service; j

2) Verifying the generator capability to reject a load of greater than or equal to 825 kW while maintaining voltage at 4160 1 420 volts and frequency at 60 1 1.2 Hz;

3) Verifying the generator capability to reject a load of 7000 kW l, without tripping. The generator speed shall not exceed 500 rpm l

,i during and following the load rejection; l 4) Simulating a loss-of-offsite power by itself, and:

a) Verifying deenergization of the emergency busses and load shedding from the emergency busses, and l

CATAWBA - UNIT 1 3/4 8-4 P84W Oe

ELECTRICAL POWER SYSTEMS -

SURVEILLANCE REQUIREMENTS (Continued) limits during this test. Within 5 minutes after completing this 24-hour test, perform Specification 4.8.1.1.2g.6)b);*

8) Verifying that the auto-connected loads to each diecel generator do not exceed the 2-hour rating of 7700 kW;

9) Verifying the diesel generator's capability to:

'a) Synchronize with the offsite power source while the generator is loaded witn its emergency loads upon a simulated restoration of offsite power, b) Transfer its loads to the offsite power source, and c) Be restored to its standby status.

10) Verifying that with the diesel generator operating in a test mode, connected to its. bus, a simulated Safety Injection signal overrides the test mode by: (1) returning the diesel generator to standby operation, and (2) automatically energizing the emergency loads with offsite power;

11) Verifying that the fuel transfer valve transfers fuel from each fuel storage tank to the day tank of each diesel via the in-stalled cross-connection lines;

12) Verifying that the automatic load sequence timer is OPERABLE with the interval between each -Mock within the tolerances given in Table 4.8-2; diesel speed ~

13) Verifyi g hat the volta anda fr: g:ncy t eranc for t e acceleraled sequen e)( ermissives'are .- + W e. 8 + 1%, ective ,

with a ' imum time delay of M , and -

' o.2

14) Verifying that the following di generator lockout features prevent diesel generator starting only when required:

a) Turning gear engaged, or b) Maintenance mode.

If Specification 4.8.1.1.2g.6)b) is not satisfactorily completed, it is not necessary to repeat the preceding 24-hour test. Instead, the diesel generator may be operated at 7000 kW for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temperature has stabilized.

~

CATAWBA - UNIT 1 ,

3/4 8-6 ~

9

ELECTRICAL POWER SYSTEMS j SURVEILLANCE REOUIREMENTS (Continued)

h.

At least once per 10 years or after a'ny modifications which could affect diesel generator interdependence p ting both diesel

generators simultaneously, during shutd(wn, verifying that both diesel generators accelerate to at leas t, p in less than or l

I i equal to 11 seconds; and '

i. At least once per 10 years by:

4

1) . Draining each fuel oil storage tank, removing the accumulated f sediment and cleaning the tank using'a sodium hypochlorite ;

solution or its equivalent, and

2) Performing a pressure test of those portions of the diesel fuel i oil system designed to Section III, subsection ND of the ASME ,

Code at a test pressure equal to 110% of the system design' pressure.

4.8.1.1.3 Reports - All diesel generator failures, valid or non-valid, shall be reported in a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days. Reports of diesel generator failures shall include the informa- ,

tion recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revi-sion 1, August 1977. If the number of failures in the last 100 valid tests (on a per nuclear unit basis) is groater than or equal to 7, the report shall be supplemented to include the additional information recommended in Regulatory Position C.3.b of Regulatory Guide 1.108, Revision 1, August 1977.

4.8.1.1.4 Diesel Generator Batteries - Each diesel generator 125-volt battery bank and charger shall be demonstrated OPERABLE:

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

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

2) The overall batter < voltage is greater than or equal to 125 volts on float charge. -

t i

b. At least once per 92 days and within 7 days after a battery discharge with battery terminal voltage below 110 volts, or battery

overcharge with battery terminal voltage above 150 volts, by verifying that:

l l 1) There is no visible corrosion at either terminals or '

l connectorsf-m um ;;rm:: tier .;siste..;; cf t % it;;.; i: 'aee-

! tt.ei, (150) x 10 " .;...., a..d

2) The average electrolyte temperature of six connected cells is i above 60'F.

THIS PAGE 0?EN PEN 91NG RECEIPT OF INFORMib: M.RX1 THE APPLICANT CATAWBA - UNIT 1 3/4 8-7

, - - , . - . . _ , . - ---w. ,. - ,. , _ - , - _ , , _ , w---. , , -

ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

4) The battery charger will supply at least 200 amperes at a minimum of 125 volts for at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

d. At least once per 18 months during shutdown, by verifying that the battery capacity is adequate to either:

1) Supply and maintain in OPERABLE status all of the actual emergency loadsfor1hourpheqthebatteryissubjectedtoabattery gf./ W service test; r

2) Supplyadmmy[ load to or equal o 4&M BatterisIEBAandif of great r than pe es for t e first min te of the fi st hour, greater t an e ual o

~

amperes for he' next 59 inutes of the first hou an} g t than or equa to mperes for the second hour hile mai ing the batt .inal voltage greater than or equal /06 volts

^

e. Atleastonc$perkmoths, uring shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. Once per 60 month interval this performance discharge test may be performed in lieu of the battery service test required by Specification 4.8.2.1.1d.; and

f. At least once per 18 months, during shutdown, by giving performance ,

discharge tests of battery capacity to any battery that shows signs of degradation or has reached 85% of the service life expected for the application. Degradation is indicated when the battery capacity drops more than 10% of rated capacity from its average on. previous -

performance tests, or is below 90% of the manufacturer's rating.

( 4.8.2.1.2 Each D.C. channel shall be determined OPERABLE and energized with tie breakers open between redundant busses at least once per 7 days by veri-

) fying correct breaker alignment, indicated power availability from the charger l

and battery, and voltage on the bus of greater than or equal to 125 volts.

l l

l THIS PAGE 0?E.1 PEN 9ING RECEIPT OF INFORMATidil;El TriE AP?UCANT l

l t

CATAWBA - UNIT 1 3/4 8-14 i Jutt 8 E84 l

e TABLE 3.8-1 CONTAINMENT PENETRATION CdNDUCTOR OVERCURRENT PROTECTIVE DEVICES

~

-0EVICE NUMBER & LOCATION SYSTEM POWERED

1. 6900 VAC Swgr Primary Bkr RCP1A Reactor Coolant Pump 1A Backup Bkr 1TA-3 Primary Bkr RCPIB Reactor Coolant Pump 1B Backup Bkr ITB-3 '

Primary BKR RCP1C Reactor Coolant Pump 1C Backup Bkr 1TC-3 1

Primary BKR RCP1D Reactor Coolant Pump 10 Backup Bkr 1TD-3

2. 600 VAC MCC t

IEMXC-F01B. b Primary Bkr 'Accumulato 1/ 0 charge Backup Fuse Isol,ylv. NI7 . . _ .

. 1EMXC-F01C - I Primary Bkr Check Valve Test Header Backup Fuse Co',ntfIsol V7v'1NI95A ,-

1EMXC-F02A Primary Bkr Train A Alternate Power Backup Fuse To ND LTDN V1v 1ND1B '

1EMXC-F028 Primary Bkr Hot Leg Inj. c Viv Backup Fuse Test Isol V v iM11.53 A I

l 1EMXC-F02C '

l Primary Bkr Cont Isol at 134 Deg Backup Fuse Annulus Area Viv IVI312A l ,

1EMXC-F03A l Primary Bkr NC Pump 1C Thermal Barrier Outlet Backup Fuse Isol Viv IKC345A l

1EMXC-F038 -

Primary Bkr N2 to Prt Cont Isol Inside Backup Fuse Viv 1NC54A 1EMXC-F03C Primary Bkr Pressurizer Power-0perated Backup Fuse Relief Isol Viv 1NC33A CATAWBA - UNIT 1 3/4 8-21 i

JUN 221934

~ , . .. - .. . ,..

TABLE 3.8-1 (Continued) ,

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES '

OEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued)

IEMXC-FOSA Primary Bkr NCDT Vent Inside Cont Isol Backup Fuse Viv 1WL450A IEMXC-F05B Primary Bkr Cont Sump Pumps Discharge Inside Backup Fuse Cont Isol Vlv 1WL825A 1EMXC-F05C Primary Bkr Vent Unit Cond Drn Tank Backup Fuse Outside Cont Isol Viv 1WL867A 1EMXC-!- 1 Prit.a: s 'kr NCDT Pumps Disch Inside Cont Isol Backup Fuse Viv IWL805A 1EMXC-F07B, Primary Bkr Cont H2 Purge Outlet Cont Isol Backup Fuse Viv IVY 17A 1EMXD-F01A Primary Bkr ND Pump 1A fon From NC Backup Fuse Loop B Viv B 1EMXD-F018 Primary Bkr Accumulator IB Discharge Backup Fuse Isol Viv 1NI658 IEMXD-F01C Primary Bkr NI Pump A to Hot Leg Check Backup Fuse Viv Test Isol Viv 1NI122B 1EMXD-F02A Primary Bkr ND Pump 18 Suction from NC Backup Fuse Loop C Vlv 1ND36B 1EMXD-F028 Primary Bkr ND to Hot Legs Chk 1NI125, 1NI129 Backup Fuse Test Isol Viv 1NI1548 CATAWBA - UNIT 1 3/4 8-22 ;

=

TABLE 3.8-1 (Continued)

DRA:T ,

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES i

., DEVICE NUMBER & LOCATION SYSTEM POWERED

[

2. 600 VAC MCC (Continued) '

t 1EMXL-F108 '

Primary Bkr Reactor Vessel Head Vent Backup Fuse Viv INC2518 l

1EMXL-F10C !

Primary Bkr Reactor Vessel Head Vent Viv ;

Backup Fuse INC252B

IEMXL-F11A '

Primary Bkr Containment Air Return I Backup Fuse Fan Motor 1B I IEMXL-F11B Primary Bkr Hydrogen Skimmer Fan Motor 1B Backup Fuse 1EMXS-F018 Primary Bkr NC Pumps Seal Rtn

. Backup Fuse Inside Cont Isol Viv 1NV89A 1EMXS-F02A Primary Bkr ND Pump 18 Suction from NC i Backup Fuse Loop C Viv 1ND37A 1EMXS-F02B '

Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 1NC250A

-1EMXS-F03C Primary Bkr NO Pump 1A Suction from NC Backup Fuse Loop B Viv 1ND2A 1EMXS-F030 Primary Bkr Reactor Vessel Head Vent Viv Backup / F INC253A 1EMXS- 04 Pri S/G ID Blowdown Inside Cont Backu p sa Isol Viv 1BB8A 1EMXS Pri ary S/G IB Blowdown Inside Cont Backup Fuse Isol Viv 1BB19A .

i CATAWBA - UNIT 1 3/4 8-27

,-n ,-n.-,- ..

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

TABLE 3.8-1 (Continued) ,

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1EMXS-F05A Primary Bkr S/G 1A Blowdown Inside Cont l Backup Fuse Isol Viv 1BB56A 1EMXS-F058 Primary Bkr S/G 1C Blowdown Inside Cont i Backup Fuse Isol Viv 1BB60A 4

1EMXS-F05C Primary Bkr Pzr Liquid Sample Line Inside Backup Fuse Cont Isol Viv 1NM3A 1EMXS-F06A '

Primary Bkr Pzr Steam Sample Line Inside Backup Fuse Cont Isol Viv 1NM6A 1EMXS-F068 Primary Bkr NC Hot Leg A Sample Line Backup Fuse Inside Cont Isol Viv 1NM22A 1EMXS-F06C Primary Skr NC Hot Leg C Sample Line Backup Fuse Inside Cont Isol Viv 1NM25A 1MXM-F01A Primary Bkr Reactor Coolant Pump Motor Backup Fuse Drain Tank Pump Motor 1MXM-F02A Primary Bkr NC Pump 18 Oil Lift Backup Fuse Pump Motor 1 '

1MXM-F028 Primary Bkr NC Pump 1C 011 Lift Backup Fuse Pump Motor 1 1MXM-F03A '

Primary Bkr Ice Condens wer Backup Fuse Transforme ($CT14 1MXM-F03B 3-Primary Bkr Ice Condenser A r Handling Unit Backup Fuse 186 Fan Mot &B r CATAWBA - UNIT 1 3/4 8-28

,-r-,,--w,-- -ng,. ---m- -m--sr - --- ~ n ,-

-

e, ---+-w- --n

. a TABLE 3.8-1 (Continued) ;

CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES !

4 DEVICE NUMBER & LOCATION SYSTEM POWERED '

l 2. 600 VAC MCC (Continued) 1MXN-F03A 5

Primary Bkr Backup Fuse (fe3C ondenser Power Transformer

, $T B 1MXN-F038 Primary Bkr Ice Condenser Bridge Crane 1 Backup Fuse Crane No. R011 1MXN-F03E Primary Bkr Stud Tensioner Hoist 1A

, Backup Fuse 4

1MXN-F04D Primary Bkr Lighting Transformer ILR5 Backup Fuse 1MXN-F04E Primary Bkr Lighting Transformer ILR6 ,

Backup Fuse l

1MXN-F05A Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Defrost Heater 1B

, IMXN-F05B >

Primary Bkr Ice Condenser Refrigeration Floor Backup Fuse Cool Pump Motor 1B

- ~

1MXN-F05C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 1B IMXN-F06A Primary Bkr Ice Condenser Air Handling .

Backup Fuse Unit 181 Fan Motor A & B IMXN-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A2 Fan Motor A & B 1MXN-F06C t Primary Bkr Ice Condenser Air Handling Backup Fuse -

Unit 183 Fan Motor A & B

+

CATAWBA - UNIT 1 3/4 8-31 e, , , , , , - - , . , , . , , ,

- . ---i TABLE 3.8-1 (Continued) i CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

L

2. 600 VAC MCC (Continued) 1MXZ-F070 Primary Bkr Reactor Cavity Manipulator s Backup Fuse

, Crane No. R007 & R027 1MXZ-F08A Primary Bkr Stess' Generator Drain Pump Backup Fuse 1MXZ-F08C '

Primary Bkr Ton Equipment Access Hatch Backup Fuse Hoist Crane No. R009 1MXZ-F080 Primary Bkr Control Rod Drive 2 Ton Jib Backup Fuse Hoist Crane No. R017 1

1MXZ-F08E Primary Bkr Reactor Side Fuel Handling

, Backup Fuse Control Console SMXG-F01C Primary Bkr Standby Makeup Pump Drain Isol i

Backup Fuse Viv 1NV876 SMXG-F05C Primary Bkr Pressurizer Heaters 28, 55 & 56 Backup Fuse SMXG-F06A _

Primary Bkr Standby Makeup Pump to Seal Backup Fuse Water Line Isol V1v 1NV877

'3. 600 VAC Pressurizer Heater Power Panels

, PHP1A-F01A Primary Bkr Pressurizer Heaters

?

Backup Fuse 1, 2, & 22 PHP1A-F018 Primary Bkr Pressurizer Heaters Backup Fuse 5, 6, & 27 CATAWBA - UNIT 1 3/4 8-38

+

8 REFUELING OPERATIONS ff

] .

3/4.9.4 CONTAINMENT BUILDING PENETRATIONS 3

_i LIMITING CONDITION FOR OPERATION I

J 3.9.4 The containment building penetrations shall be in the following status:

}-

a. The equipment hatch closed and held in place by a minimum of four

. bolts,

. b. A minimum of one door in each airlock is closed, and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:

1) Closed by an isolation valve, blind flange, or manual valve, or

2) Exhausting through an OPERABLE Reactor Building Containment Purge System HEPA filters and charcoal adsorbers.

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

ACTION:

With the requirements of the above specification not satisfied, immedfately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in'the containment building.

SURVEILLANCE REQUIREMENTS 4.9.4.1 Each of the above required containment building potetrations shall be determined to be either in its closed / isolated condition or exhausting through an OPERABLE Reactor Building Containment Purge System with the capability of being automatically isolated upon heater failure within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS or movement of 4

irradiated fuel in the containment building by:

a. Verifying the,penM s are in th M /is lated condition,

ha,,) law yyty ad ukott

b. Verifyi g the[containm tp ge/ksolationvale close upon a High Relative idity_t es- sig al 5

r i

CATAWBA - UNIT 1 3/4 9-4

~ . _ _ _ _ . . - -

i

! REFUELING OPERATIONS l SURVEILLANCE REOUIREMENTS (Continued) t E

4.9.4.2 The Reactor Building Containment Purge System shall be demonstrated OPERABLE - ; ;_ __

7--~_ , _ __ _ ._ _

i

a. At least once per 31 days by initiating, from the control room, flow

.' through the HEPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heaters operating;

b. At least once per 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire, or chemical release in any ventilation zone communi-cating with the system by:

1) Verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of I

less than 1% and uses the test procedures guidance in Regula-tory Positions C.5.a, C.5.c, and C.S.d of Regulator 1.52, Revision 2, March 1978, and the system flow rate i . .;Guld .,000 fm i 10% (both exhaust fans operating); 26,ooci

2) Verifying within 31 days after removal, that a labora ory analysis of a presentative carbon sample obtained in accordance with Regulatory Position C.6.b of Regulatory Guide 1.52, Revi-sion 2, March 1978, meets the laboratory testin iteria of '

Regulatory Position C.6.a of Regulatory Guide 1 Revision 2, March 1978, for a methyl iodide penetration of than 6%;

and 5,ooo

3) Verifying a system flow rate f 20,000 fm i 10% (both exhaust m

fans operating) during syste operatio when tested in accordance with ANSI N510-1980. .

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, within 31 days after removal, that a laboratory analysis of a repre-sentative carbon sample obtained in accordance with Regulatory ^

Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,

meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 6%;

d. At least once per 18 months by: cephea

1) Verifying that the pressure drop act ssthe(combinedHEPA filters, charcoal adsorber banks, d -^ # " -- - ---";r- s less than 8 i hv Water Gauge whi e operating the s stem at a flow rate of 48;006- fm-*-10%-(bot ust fan operating);

,/ z-S,oco f) '/o AIy!.s *h,[_1)^ ItCT C00liSs byyaaa valv6a ;;n E2 ???"?d

-by c? r;ter ;;t.;uus J

CATAWBA - UNIT 3/4 9-5 /

~ --

9'U# e - e aps ema e tte * - em. 3 - w am. g a e. . Ge e- e.we o e

(... . .. .

' REFUELING OPERATIONS t

SURVEILLANCE REQUIREMENT (Continued) 2 X) Verifying hat), atehpdissipae3 0 12 kW when tested in accordance with ANSI N51 M 980.

s

e.

After each complete or partial replacement of a HEPA filter bank, by i verifying that the cleanup system satisfies the in place penetration 4

and bypass leakage testing acceptance. criteria of_less than 1% in ;

accordance with ANSI N510-19J DOP test aerosol while operating the system at a flow ratef 00,000 e i 10% (both exhaust fans .

cperating); and 15,ocx) i

f. After each con.plete. or partial replacement of a charcoal adsorber !

bank, by veaifying that the cleanup system bank satisifies the !

in place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydrocarbon f gera test gas while operating the system at a [

flow rate. f-20,000 cf i 10% (both exhaust fans operating). i 25,ood i

E i

.t e

i CATAWBA - UNIT 1 3/4 9-6 t

I REFUELING OPERATIONS 3/4.9.11 FUEL HANDLING VENTILATION EXHAUST SYSTEM LIMITING CONDITION FOR OPERATION

.i i

3.9.11 At least one train of the fuel Handling Ventilation Exhaust System

, shall be OPERABLE.. .. _... .. ~ , .

APPLICABILITY: Whenever irradiated fuel is in.the storage pool.

ACTION:

a. With both trains of the Fuel Handling Ventilation Exhaust System t

inoperable, suspend all operations involving movement of fuel within ;

! the storage pool or crane operation with loads over the storage pool until the Fuel Handling Ventilation Exhaust System is restored to OPERABLE status.

t

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

i SURVEILLANCE REQUIREMENTS 4.9.11.1 One train of the Fuel Handling Ventilation Exhaust System shall be determined to be operating and discharging through the HEPA filter and charcoal adsorbers at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever irradiated fuel is being moved in the storage pool and during crane operation with loads over the storage pool.

4.9.11.2 Both trains of the Fuel Handling Ventilation Exhaust System shall be demonstrated OPERABLE:

a. At least once per 31 days by initiating, from the control room, l flow through the HEPA filters and charcoal adsorbers and verifying

that the system operates for at least 10 continuous hours with the heaters operating; l

'a . At least once per 18 months or (1) after any st.uctural maintenance on the HEPA filter or charcoal adsorber housings, or (2) following l painting, fire, or chemical release in any ventilation zone communicating with the system by:

l

1) Verifying that the cleanup system satisfies the in place

penetration and bypass leakage testing acceptance criteria

of less than 1% and uses the test procedure guidance in t sguistoty Positions C.5.a, C.5.c, and C.S.d of Regulatory ideT1.5 Revision 2, Maren 1978, and the system flow rate i} s m i 10%;

u ss 1

CATAWBA - UNIT 1 3/4 9-14 l .-

\.

REFUELING OPERATIONS ['

SURVEILLANCE REQUIREMENTS (Continued)

Verifying, within 31' days aftef renioval,' that a labora~ tory -

~

~2) analysis of a representative carbon sample obtained in accor-dance with Regulatory Positions C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of RegulaJorg Guide 1.52, Revision 2,

' ~ March 1978, for a methyl iodide pe tfition of less than 1%;

and 4

, lkbY '

3) Verifying a system flow rate o 14r000 m i 10% during system operation when te'sted in accor ith ANSI N510-1980.

c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation in any train by verifying, within 31 days after removal, that a laboratory analysis '

of a representative carbon sample obtained in accordance with Regula-tory Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets the laboratory testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl iodide penetration of less than 1%.

d. At least once per 18 months for each train by:

1) VerifyingihatthepressuredropacNssthecombinedHEPA filters, charcoa adsorber banks, and moisture separators is less than 8 jfn es ter Gauge while operating the systc.n at a flow rate 91 19,000 c m i 10%.

\ t6'

2) Verifying at ,S4.6 system maintains the spent fuel storage pool area at a negative pressure of greater than or equal to % inch Water Gauge relative to the outside amosphere during system operation, f 3) Verifying that the filter cooling bypass valves can be manually opened, and

4) Verifying that the heaters dissipate 80 1 8 kW when tested in accordance with ANSI N510-1980.

e. After each complete or partial replacement of a HEPA filter bank in any train, by verifying that the cleanup system satisfies the in-place penetration and bypass leakage testing accep'tance criteria of less than 1% in accordance with ANSI N510-198 r a 00P test aerosol while operating the system at a flow rate o IS,00? cfm i 10%; and After each complete or partial replacement 4[5

~

f. charcoal adsorber bank in any train, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1980 for a halogenated hydrocar o er erant test gas while operating the system at a flow rate o , cf 10%.

e 14,5(,

CATAWBA - UNIT 3/4 9-15

.e e-= ,e--

, ,y, = . er Oe **-d'we **h * ** * * * ' *

,----.e e w --m-.. * --- . - - ,- , , - - - - - -

TABLE 3.12-1 (Continued) DR) FT

.},g. TABLE NOTATIONS ,

J (1) g5 ecifi MpeT par rs9of distance and direction sector from the centerline pfl Wand additional description where pertinent, shall be ;

r ded fo M ach and every sample location in Table 3.12-1 in a table '

and figure (s) in the ODCM. Refer to NUREG-0133, " Preparation of Radio-

' logical Effluent Technical Specifications for Nuclear Power Plants,"

October 1978, and to Radiological Assessment Branch Technical Position, Revision.1, November 1979. Deviations are permitted from the required

sampling schedule if specimens are unobtainable due to circumstances such 1

as hazardous conditions, seasonal unavailability, and malfunction of i automatic sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to comple .orrec-tive action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report pursuant to Specification 6.9.1.6. It is recognized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable specific alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program given in the ODCM. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report . documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples for that pathway and justifying the selection of the new location (s) for obtaining samples.

(2) One or more instruments, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.

l Film badges shall not be used as dosimeters for measuring direct radia-tion. The 40 stations is not an absolute number. The number of direct radiation monitoring stations may be reduced according to geographical l limitations; e.g., at an ocean site, some sectors will be over water so i

that the number of dosimeters may be reduced accordingly. The frequency l of analysis or readout for TLD systems will depend upon the characteristics

' of the specific system used and should be selected to obtain optimum dose information with minimal fading.

(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

CATAWBA - UNIT 1 3/4 12-7 A %-

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POWER DISTRIBUTION LIMITS j

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BASES t i

HEAT FLUX HOT CHANNEL FACTOR, and REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR '

ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

When Reactor Coolent System flow rate and F are measured, no additional allowances are neces ar (

rior to comparison with the limits of Figure 3.2-3. !

Measurement errors f 2. r Reactor Coolant System total flow rate and 4% ,

for F H have been al owed for in determination of the design DNBR value.

O The measurement or or Reactor Coolant System total flow rate is based .

upon performing a precision heat balance and using the result to calibrate the '

Reactor Coolant System flow rate indicators. Potential fouling of the feedwater j j venturi which might not be detected could bias the result from the precision ;

heat. balance in a nonconservative manner. Therefore, a penalty of 0.1% for <

undetected fouling of the feedwater venturi is included in Figure 3.2-3. Any fouling which might bias the Reactor Coolant System flow rate measurement I

greater than 0.1% can be detected by monitoring and trending various plant performance parameters. If detected, action shall be taken before performing i subsequent precision heat balance measurements, i.e., either the effect of the l fouling shall be quantified and compensated for in the Reactor Coolant System :

flow rate measurement or the venturi shall be cleaned to eliminate the fouling. !

The 12-hour periodic surveillance of indicated Reactor Coolan't System l flow is sufficient to detect only flow degradation which could lead to opera- [

tion outside the acceptable region of operation shown on Figure 3.2-3. :

l- 3/4.2.4 QUADRANT POWER TILT RATIO

The QUADRANT POWER TILT RATIO limit assures that the radial power distribu-tion satisfies the design values used in the power capability analysis. !

Radial power distribution measurements are made during STARTUP testing and i periodically during power operation. l

. E

.The limit of 1.02, at which corrective action is required, provides DNB !

and linear heat generation rate protection with x y plane power tilts. A f limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt. )

The 2-hour time allowance for operation with a tilt condition greater !

than 1.02 but less than 1.09 is provided to allow identification and correction i of a dropped or misaligned control rod. In the event such action does not :

correct the tilt, the margin for uncerta'nty on Fqis reinstated by reducing L the maximum allowed power by 3% for each percent of tilt in excess of 1. ;

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore ;

detector is inoperable, the moveable incore detectors are used to confirm that -

i the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore l I

i i

! CATAWBA - UNIT 1 8 3/4 2-5 !

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l REACTOR COGLANT SYSTEM BASES PRESSURE /TEMPERATURELIMITS(Continued)

Although the pressurizer operates in temperature ranges above those for which there is reas n fer wi. sir, r,f-nonductile failure, operating limits are provide ssure compatibility of opeh on with the fatigue analysis

_performe in accordance with the ASME Code requ ements.

% Tv4PfAA*Ttatr attunews__nr k'rGeYtosi

The OPERABILITY of two PORVs or a Rea dr Coolant System vent opening of l at least 4.5 square inches ensurac he Reactor Coolant System will be

_rotec+ad N . inEssiire transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or scre of the cold legs are less than or equal to 285'F. Either PORV has adequate relieving capability to protect the Reactor Coolant System from overpressurization when the transient is limited to either: l (1) the start of an idle reactor coolant pump with the secondary water temperature of the steam generator less than or equal to 50*F above the cold leg temperatures, or (2) the start of a Safety Injection pump and its injection into a water

. solid Reactor Coolant System.

!Y -1P pM pG4 3 4.4.10 STRUCTURAL INTEGRITY

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The inservice ir.spection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in acccrdance with Section XI of the ASME. Boiler end Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) 3xcept where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g.)(6)(1).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, and applicable Addenda as required by i 10 CFR 50.55a(g) except where specific written relief has been granted by the -

Commission pursuant to 10 CFR 50.55a(g)(6)(1).

CATAWBA - UNIT 1 8 3/4 4-15 ggg

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(.m.dT# W The Max ' mum Allowed PORV Setpoint for theJinM Overpressure "'E N i L..

is derived by analysis which models the performance of L7 Y h ** Operation teS stemwith a PORV assuming various mass input and heat input transients.

setpoint less than or equal to the maximum 1-setpoint ensures that Appendix G criteria will not be violated with ^~

consideration for,g4 a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal processing &

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single failure. To ensure that mass and heat input transients more ,_._.

severe than those assumed cannot occur, technical specifications }

require lockout of both safety injection pumps and all but one centri- -

fugal charging pmip while in MODES 4, 5 and 6 with the reactor vessel head installgd and disallow start of a RCP if secondary tenperature is i

G more than 50 F above primary temperature. M Lute +,Jm. Popeeh &

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! System will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required l

j by 10 CFR 50, Appendix H and in accordance with the schedule in Table -

I 4.4-5.

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t-EMERGENCY CORE COOLING SYSTEMS .

BASES '

ECCS SUBSYSTEMS (Continued)

The limitation for a maximum of one centrifugal charging pump and one i Safety Injection pump to be OPERA 8LE and the Surveillance Requirement to verify all charging pumps and Safety Injection pumps except the required OPERABLE centrifugal charging pump to be inoperable below 285'F provides assurance that a mass addition pressure transient can be relieved by the ;

operation of a single PORV. ;

The Surveillance Requirements provided to ensure OPERA 8ILITY of each !

component ensures that at a minimum, the assumptions used in the safety l analyses are met and that subsystem OPERABILITY is maintained. Surveillance !

Requirements for throttle valve position stops and flow balance testing provide j assurance that proper ECCS flows will be maintained in the event of a LOCA. !

Maintenance of proper flow resistance and pressure drop 'in the piping system l to each injection point is necessary to: (1) prevent total pump flow from !

exceeding runout conditions when the system is in its minimum resistance ,

l configuration, (2) provide the proper flow split between injection points '

in accordance with the assumptions used in the ECCS-LOCA analyses, and

(

(3) provicle an acceptable level of total ECCS flow to all injection points -

equal to or above that assumed in the ECCS-LOCA analyses. ,

(

l 3/4.5.4 REFUELING WATER' STORAGE TANK -

'The OPERABILITY of the refueling water storage tank as part of the ECCS ensures that a sufficient supply of borated water is available for injection by the ECCS in the event of a LOCA. The limits on minimum volume and boron concentration ensure that: (1) sufficient water is available within containment i to permit recirculation cooling flow to the core, and (2) the reactor will.

remain subcritical in the cold condition following mixing of the refueling water storage tank ar.d the Reactor Coolant System water volumes with all control ~

l rods inserted except for the most reactive control assembly. These assumptions i

! are consistent with the LOCA analyses. !

The contained water volume limit includes an allowance for water not usable ;

because of tank discharge line location or other physical characteristics. ;

r.

j The limits on contained water volume and boron concentration of the i

refueling water storage tank also ensure a pH value of between 8.5 and 10.5 for the solution recirculated within containment after a LOCA. This pH band mini- ;

mizes the evolution of iodine and minimizes the effect of chloride and caustic

, , stress corrosion on mechanical systems and components. I CATAWBA - UNIT 1 8 /4%52 ggg

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CONTAINMENT SYSTEMS .

BASES i f

f 3/4.6.1.8 ANNULUS VENTILATION SYSTEM ' I i

The OPERABILITY of the Annulus Ventilation System ensures that during LOCA i conditions, containment vessel leakage into the annulus will be filtered through !

the HEPA filters and' charcoal adsoEber trains' prior to discharge to the atmosphere. !

Operation of the system with the heaters operating to maintain low humidity using automatic control for at least 10. continuous hours in a 31-day period is !

sufficient to re' duce the. buildup of moisture on the adsorbers and HEPA filters. '

This requirement is necessary to meet the assumptions used in the safety ar.alyses and limit the SITE BOUNDARY radiation doses to within the dose guide- l line values of 10 CFR Part 100 during LOCA conditions. ANSI N510-1980 will be used as a procedural guide for surveillance testing.

l

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3/4.6.1.9 CONTAINM T* S

. The containment nd exhaust isolation valves for the lower .

}- compartment (24-inch), and instrument-room (12-inch), and the Hydrogen Purge [

System (4-inch) are required to be sealed closed during plant operation since [

these valves have not been demonstrated capable of closing during a LOCA. [

Maintaining these valves sealed closed during plant operations ensures that !

c excessive quantities of radioactive materials will not be released via the ;

! Containment Purge System. To provide assurance that these containment valves [

cannot-be inadvertently opened, the valves are sealed closed in accordance with !

Stanqard Review Plan 6.2.4 which includes mechanical devices to seal or lock l the valve closed, or prevents power from being supplied to the valve operator, k i

L The use of the containment purge lines is restricted to the 24-inch purge ;

supply and exhaust isolation valves for the upper compartment and the 4-inch l Containment Air Release and Addition System valves since, unlike the lower i compartment, instrument room, and the Hydrogen Purge System valves, these i 24-inch valves and 4-inch valves are capable of closing during a LOCA. There- .j

( fore, the SITE BOUNDARY dose guideline values of 10 CFR Part 100 would not be l l exceeded in the event of an accident during containment purging operation. (

Operation with the line open will be limited to 250 hours0.00289 days 0.0694 hours 4.133598e-4 weeks 9.5125e-5 months during a calendar year i

-for the 24-inch valves and 2000 hours0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months during a calendar year for the 4-inch !

l valves. The total time the containment purge (vent) system isolation valves l may be open during MODES 1, 2, 3, and 4 in a calendar year is a function of l anticipated need and operating experience. Only safety-related reasons; e.g., !

containment pressure control or the reduction of airborne radioactivity to- l l facilitate personnel access for surveillance and maintenance activities, may be j l

used to support the additional time requests. l Leakage integrity' tests with a maximum allowable leakage rate for contain- f ment purge supply and exhaust valves will provide early indication of resilient t material seal degradation and will allow opportunity for repair before gross t leakage failures could develop. The 0.60 L, leakage limit of Specification [

3.6.1.2b. shall not be exceeded when the leakage rates determined by the leakage integrity tests of these valves are added to the previously determined total for all v11ves and penetrations subject to Type B and C tests. ,

CATAWBA - UNIT 1 B 3/4 6-3 JUN 22 B84

. - .._...-. 2 . . - - - - - - - . - ....-.. . ... -.. : .- .- - . . . . .

4 U DRA-..

PLANT SYSTEMS l l~

l BASES 3/4.7.7 AUXILIARY BUILDING FILTER XHAUST SYSTEM i

!i The OPERABILITY of the Auxiliar Filtered ti t Mxhaust Systemensuresthatradioactivematerialsleakingfromthh'ECCSeq$ipmentwithin l the auxiliary building following a LOCA are filtered prior to reaching the

environment. Operation of the system with the heaters operating to maintain

! low humidity using automatic control for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers l

j and HEPA filters. The operation of this system and the resultant effect on i

offsite dosage calculations was assumed in the safety analyses. ANSI N510-1980 will be used as a procedural guide for surveillance testing.

L

3/4.7.8 SNUBBERS '

All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is main-tained during and following a seismic or other event initiating dynamic loads.

Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related system.

Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2-kip,10-kip, and 100-kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this Technical Specification would be of a different type, as would hydraulic snubbers from either manufacturer.

I A list of individual snubbers with detailed information of snubber location and size and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The accessibility of each snubber shall _

be determined and approved by the Catawba Safety Review Group. The determination shall be based upon the existing radiation levels and the expected time to per-form a visual inspection in each snubber location as well as other factors asso-ciated with accessibility during plant operations (e.g., temperature, atmosphere, location etc.), and the recommendations of Regulatory Guides 8.8 and 8.10. The addition or deletions of any hydraulic or mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

CATAWBA - UNIT 1 8 3/4 7-4

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T CHAARMAN OF THE SOARD t ANO CHIEF EXECUTIVE OFF4CER PRESIDENT AND CHIEF OPER ATING Of F4CER EXECUTIVE VICE PRESIDENT w

1 I I I I VICE PRESIDENT MANAGER aAANAGER VICE PRESIDENT VICE PRESIDENT VtCE PRESIDENT ION ESIGN CONSTRUCTION COR QUALITY PROJECT CONTROL PR NWLMR NM NUCLE AR SAFETY REVIEW BOARD I I NUCLEAR PRODUCTION GENERAL MANAGER .

STAFF NUCLEAR STATIOhs l

MANAGER McGUIRE NUCLEAR STATION

! DUKE POWER CORPORATE

! ORGANIZATION t rem ICATAHBANUCLEARSTATION l Figure 6.2-1 i

I

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ADMINISTRATIVE CONTROLS

/

RECORD RETENTION (Continued) m.

Records of secondary water sampling and water quality; and

n. -

Records of analyses required by the Radiological Environmental Monitoring the analysisProgram at a laterthat would permit evaluation of the accuracy of date. This should include procedures effective at were procedures specified times and QA records showing that these followed.

6.11 RADIATION PRO'TECTION PROGRAM 6.1k.Proceduresforpersonnelradiationprotectionshallbe repared co istent with the requirements of 10 CFR Part 20 and shall p adhered to for all operations involving personn ed, maintained, and diation exposure.

6.12 HIGH RADIATION AREA b he kn oc epd 6 100 mg/hr se Me44Uvs<m 7

6.12.1 In lieu of the " control device" o " alarm signal" required b paragraph 20.203(c)(2) of 10 CFR Part 20, each high radiation area, yas defined (t8;Q a in 10 CFR Part 20, in which the intensity }

1000 mR/h at 45 cm (18 in.) from the radiat\i (source or from any surface w; the radiation penetrates shall be barricaded and conspicuourly pvai.eu as a--

high radiation area and entrance thereto shall be controlled by requiring

( issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g. , Health Physics Technician) or personnel continu-ously escorted by such individuals may be exempt from the RWP issuance require-ment during the performance of their assigned duties in high radiation areas with exposure rates equal to or less than '000 mR/h, provided they are other-wise radiationfollowing areas. plant radiation protection procedures for entry into such high Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following: '

a.

A radiation monitoring device which continuously indicates the radiation dose rate in the area; or .

b.

A radiation monitoring device which continuously integrates the '

radiation dose dose rate in the area and alarms when a preset integrated is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been estab-lished and personnel have been made knowledgeable of them; or c.

An individual qualified in radiation protection procedures with a t radiation dose rate monitoring device, who is responsible for pro-viding positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency '

specified by the Station Health Physicist in the RWP.

t i

e CATAWBA - UNIT 1 6-21 JUN 8 W

. ~.. . .

. . . . . - . . - - -

ML20092L140

Text

Dear Mr. Denton:

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