Text

Forwards marked-up Proof & Review Tech Specs Re Reactor Trip Sys Instrumentation Trip Setpoints,Reactor Core,Limiting Safety Sys Settings,Flow Paths,Charging Pump Shutdown & Power Distribution Limits
	ML20083J988
Person / Time
Site:	Catawba
Issue date:	04/04/1984
From:	Tucker H; DUKE POWER CO.
To:	Adensam E, Harold Denton; Office of Nuclear Reactor Regulation
References
	NUDOCS 8404160017
	Download: ML20083J988 (67)
	v • d • e

_

e54 DUKE POWER GOMPANY P.O. Box 03180 CIIAltLOTTE, N.C. 28242 a Em = ( os 3 a

= = = - - April 4, 1984 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Ms. E. G. Adensam, Chief Licensing Branch No. 4 Re: Catawba Nuclear Station, Unit 1 Docket No. 50-413 Proof and Review Technical Specifications

Dear Mr. Denton:

Please find attached proposed changes to the Proof and Review Technical Specifications for Catawba Unit 1. These changes reflect corrections to errors presently contained in the Specifications.

Very truly yours, sd b v.

Hal B. Tucker' RWO/php Attachment cc: Mr. James P. 0'Reilly Mr. Jesse L. Riley Regional Administrator Carolina Environmental. Study Group U. S. Nuclear Regulatory Commission 854 Henley Place Region II .

Charlotte, North Carolina 28207 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 Palmetto Alliance 21351 Devine Street NRC Resident Inspector. Columbia, South Carolina 28207 Catawba Nuclear Station Mr. Robert Guild, Esq. -

Attorney-at-Law .

P..O. Box 12097 Charleston, South Carolina 29205 t

^.

8404160017 840404 g PDR ADOCK 05000413 L, .A., .

PDR.

PRDOF & RBTl COPY l BASES SECTION PAGE 3/4.0 APPLICABILITY........ ...................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1' B0 RATION CONTR0L.......................................... B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.......................................... B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES................................ B 3/4 1-3 3/4.2 POWER CISTRIBUTION LIMITS................................... B 3/4 2-1 3/4.2.1 AXIAL FLUX DIFFERENCE..................................... B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR....... B 3/4 2-2 FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL P0WER............................................. B 3/4 2-3 3/4.2.4 QUADRANT POWER TILT RATI0................................. B 3/4 2-5 3/4.2.5 DNB PARAMETERS............................................ B 3/4 2-6 3/4.3 INSTRUMENTATION 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION............... B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION................................ B 3/4 3-3 3/4.3.4 TURBINE OVERSPEED PROTECTION.............................. B 3/4 3-7 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1- REACTOR COOLANT LOOPS AND COOLANT CIRCU! ATION. . . . . . . . . . . . . B 3/4 4-1 3/4.4.2 SAFETY VALVES............................................. B 3/4 4-1 3/4.4.3 PRESSURIZER............................................... B 3/4 4-2 3/4.i 4 RELIEF VALVES............................................. B 3/4 4-2 3/4.4.5 STEAM GENERATORS.......................................... B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE............................ B 3/4 4-3 3/4.4.7 CHEMISTRY................................................. B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY......................................... B 3/4 4-S 3/4.4.9 ..................... B 3/4 4-7 PRESSURE TABLE B 3/4.4-1 REACTOR/TEMPERATURELIMITS.....I)[)I!?s..................

VESSEL TOUGH 5$ B 3/4 4-9 FIGURE B 3/4.4-1 FAST NEUTRON FLUE E( 1MeV)AkAFUNCTIONOF FULL POWER SERVICE LIFE. ...... ..?"....)................ B 3/4 4-10 FIGURE B 3/4.4-2 EFFECT OF FLUENC AND COPPER CO,NTENT ON SHIFT XPOSED'To 550*F............ B 3/4 4-11 OF RTNDT FOR REACTOR VESSE 3/4.4.10 STRU CTURAL INT EGRITY. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B 3/4 4-15 CATAWBA - UNIT 1 X .

h b O '

n TABLE 2.2-1 h

t REACTORTRIPSYSTEMINSTRUMENTATIONTRIDETFOIN15 E TOTAL -

JSENSOR e

ALLOWANCE P l.lRROR ;

-g FUNCTIONAL UNIT (TA) Z (S) TRIP 4 ETPOINT ALLOWABLE VALUE U 1. Mar.ual Reactor Trip N.A. N. . Mr N.A. N.A.

w

2. Power Range, Neutron Flux

a. High Setpoint 7. 5 4.56 0 $109% of RTP* $111.2% of RTP*

b. Low Setpoint 8.3 4.56 0 $25% of RTP* 127.2% of RTP*

3. Power Range,' Neutron Flux, 1.6 0.5 0 $5% of RTP* with High Positive Rate 16.3% of RTP* with a time constant a time constant 1 2 seconds 2 2 seconds

4. Power Range, Neutron Flux, High Negative Rate '

1.6 0.5 0 <5% of RTP* with i time constant

<6.3% of RTP* with i tis::e constant

, 12 seconds 32 seconds Q

m

. 5. Intermediate Range, 17.0 8.4 0 <25% of RTP* ~<31% of RTP*

Neutron Flux-

~

I ~$

6. Source Range,' Neutron Flux 17.0 10 0 $105 cps $1.4 x 105 cps

7. Overtemperature AT 6.4 3.92 2. 2 See Note 1 See Note 2 cJ c3

8. Overpower AT 4.6 1.4 u

1. 2 See Note 3 See Note 4 .- <

~

9. Pressurizer Pressure-Low 3.0 0.71 1.5 21945 psig 11934 psig

10. Pressurizer Pressure-High 3.1 0.71 1. 5 12385 psig $2396 psig

11. Pressurizer Water Level-High 5.0 2.18 1. 5 $92% of instrument $93.8% of instrument span span

12. Reactor Coolant flow-Low 2.5 1.77 0.6 190% of loop 188.8% of loop design flow ** design flow"*

RTP = RATED THERMAL POWER

- Loop design flow = 96,900 gpm

I g TAGLE 2.2-1 (Continued) h _ REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E '

, TOTAL SENSOR

-e ALLOWANCE ERROR

--e FUNCTIONAL UNIT (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE g 13. Steam Generator Water 12 12.18 1.5 , @ % of span 14evefr% of span from

' Level Low-Low / from 0% to 30% 0% to 30% RTP*

~

p RTP* increasing increasing linearly

' linearly to , to j ?.' " :' ;,_:

M

-^

59 97 af h, _ .

spn at -full leal M

14. Undervoltage - Reactor 6 -!! ^^: 1,o th77'1,,f L':7::; . 't: 7/o 7. (6016 offs)

Coolant Pumps 8.57 o.o 63 v.ti.3e a al .7 s.c

.. **5f*a5*l. _i S 55"8f 1 # d_... _ ___ _ _ --

15. Underfrequency - Reactor =4,4= Me> ;^ ':

Efeilua H w.A . ["?. :; ilz y et Coolant Pumps 4, o o.o su m m 1,o .2 m. 55 9

16. Turbine Trip rese n h e Q r,

a. Low Control Valve EH. N.A. N.A. N.A.

Ro 1550 psig 1500 psig Pressure tv

b. Turbine Stop Valve N.A. N.A. N.A. 11% open ->1% oper. Cf3 Closure #

c :2 Safety Injection Input 17.

from ESF N.A. N.A. N.A. N.A. N.A. $

sW W 5W5& Y

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

'h b N '

g TABLE 2.2-1 (Continued)

$ REACTOR TRIP SYSTEN INSTRUNENTATION TRIP SETPOINTS Ei

~7 TOTAL SENSOR ALLOWANCE ERROR E FUPCTIONAL UNIT (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE

" 18. Reactor Trip System '

T Interlocks -10

a. . Intermediate Range N.A. N.A. N.A. 11 x 0 s 16 x 10 81 amps

Neutron Flux, P-6

. b. Low Power Reactor Trips Block. P-7 ,

1) P-10 input N.A. N.A. N.A. 510% of RTP* 112.2% of RTP*

.2). P-13 input N.A. N.A. N.A. $10% RTP* Turbine $12.2% RTP* Turbine - c3 m -

Impulse Pressure Equivalent Impulse Pressure Equivalent.

j

' c2 c.- Power Range Neutron N.A. N.A. N.A. <48% of RTP*

Flux, P-8 $50.2% of RTP*

d. Power Range Neutron M.A. N.A. .N.A.

2 Flux, P-9 169% of RTP* $70% of RTP"

e. Power Range Neutron ~ N.A. N.A. N.A 210% of RTP" 37.8% of RTP*

-Flux, P-10 ca c_o

- f. Turbine Impulse Chamber N.A. N.A. N.A. <10% RTP* Turbine <12.2% P.TP* Turbine vJ Pressure, P-13 Impulse Pressure Tmr.ulse Pressure W Equivalent Equivalent

19. Reactor Trip Breakers N.A. N.A. N.A N.A. N.A.

20. Automat 1'c Trip.and N.A. N.A. H.A. N.A. N.A.

Interlock Logic

RTP = RATED THERNAL POWER i

1

a 2.1 SAFETY LIMITS PR0nF d U D N COPY C BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

' Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate bofling (DNB) and the resultant sharp reduction in heat transfer coefficient. DN8 is not a directly measurabie parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been re!ated to DN8 through the WRB-1 correlation. The WRB-1 DNB correlation ha:;

bean developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions. The local DNB heat flux ratio, (DNBR), is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

( The DNB design basis is as follows: there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is grcater than or equal to the DNDP. limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNDR limit is established based en the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit.

~

I:. meeting this design basis, uncertainties in plant operating para-meters, nuclear and thermal paramaters, and fuel fabrication parameters are considered statistically such that.there is at least a 95 confidence that the minimum DNBR for the limiting rod is~ greater than or equal to the DH8R limit.

The uncertainties in..the above plant parameters are used to determine the

plant DN8R uncertainty. This DNBR uncertainty, combined with the correlation

! DNBR limit, establishes a design DNBR-value which must be met in plant ty

! analysac using values of input parameters without uncertainties.

N I The curves of Figures 2.1-1 and 2.1-2 show the loci of points q fTfRMA POWER, Reactor Coolant System pressure and averago temperature below whic he calculated DNBR is no less than the design DNBR value, or tne average alpy at the vessel exit is less than the enthalpy of saturatud liquid.

CATAWBA - UNIT 1 B 2-1

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

PR00F & REB'l P"Y

(

4 # LIMITING SAFETY SYSTEM SETTINGS i

BASES i

i Steam Generator Water Level The Steam Generator Water Level Low-Low trip protects the reactor frem loss of heat sink in the event of a sustained steam /feedwater flow mismatch resulting from loss of normal feedwater. The specified Setpoint provides

, allowances for starting delays of the Auxiliary Feedwater System.

j: Undervoltage and Underfrecuency - Reactor Coolant Pumo Busses i

The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant

!- flow. The specified Setpoints assure a Reactor trip signal is generated before the Lew Flow Trip Setpoint is reached. Time celays are incorporated in l: the Underfrequency and Undervoltage trips to prevent spurious Reactor trips i from mecentary electrical power transients. For undervoltage, the delay is

set so that the time required for a signal to reach the Reactor trip breakers '

i following the simultaneous trip of two or more reactor coolant pump bus circuit

! breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set

! so that the time required for a signal to reach the Reactor trip breakers.

l after. the 'Jnderfrequency Trip Setpoint is reached shall not exceed 0.3 second.

On decreasing power the.Undervoltage and Underfrequency Reactor Coolant Punp

'( Bus trips are automatically blocked by P-7 (a power level of approximately 10%

of RATED THERMAL POWER with a turbine impulse chamber pressure et approximately .

10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trio

A Turbine . i itiates a Reactor trip. On decreasing power the Reactor trip from t urbine ip is automatically blocked by P-9 (a power level of

approxima ly of RA ED THERMAL POWER); and on increasing power, ruinstated l automatic ly y P'9.

WI Safety-InbioAJnput from ESF .

If a Reactor trip has not already ber.n generated by the Reactor Trip System instrumentation ~, the ESF automatic actuation logic channels will initiate a Reactor trip upca any signal which initiates a Safety Injection. The ESF instrun.entation channels which initiate a Safety Injection signal are shown in

-Table 3.3-3.

I

' CATAWBA - UNIT 1 B 2-7

!" c 2

~

1 _ . . . . .

. - . . . . ;-. . - - . - . . . . ~ - - - . - - -- . -. - . . - . . -

REACTIVITY CONTROL SYSTEMS PR00F & REYEW COPY FLOW FATHS - OPERATI'NG j LIMITING CONDITION FOR OPERATION

.3.1.2.2 At least two* of the following threc boron injection flow paths shall i

be OPERABLE:

a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump-to the Reactor Coolant System, and

b. Two flow paths from the refueling water storage tank via charging ;

, pumps to the Reactor Coolant 3ystem. '

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

! ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant Systeem OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant 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 STAND 8Y and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to GPERA8LE status within the next 7 days or be in COLD SHUTDOWN within the

'. next'30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

' SURVEILLANCE REQUIREMENTS

' i 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERA 8LE: N c,

a. At least once per 7 days by verifying that ti.e temperature of the heated portion of the'f:1ow path from the boric acid tanks is greater than or equal to'65'F when'.it is a required water source; ,

b. 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;

c. At least once per 18 months during shutdown by verifying-that each

- automatic valve in the flow path actuates tc its correct position on E a. Safety Injection-test s;gnal;.and d.- At least once per 18 months by verifying that the flow path required

~

by Specification 3.1.2.2a.-delivers at least 30 gpa to the Reactor Coolant System.

0nly one boron injection flow path is required to be OPERA 8LE whenever the temperature of one or more of the cold legs is less than or equal to

>= r .

w c.w spw CATAWBA - UNIT 1 3/4 1 - '

REACTIVITY CONTROL SYSTEMS '

\ CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION s

3.1.2.4 At least two* charging pumps shall be OPERABLE.

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

ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to 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 and borated to a SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying that a differential pressure across each pump of greater than or equal to 2380 psid is developed when tested pursuant to Specification 4.0.5.

4.1.2.4.2 All charging pumps, except the above required OPERABLE pump, shall be demonstrated inoperable at1.least once per 31 hys whenever the temperature of one or mor's of the cold legs is less than or equal to 300*F by verifying that the motor circui breakers are secured in the open position or that the discharge of each arging pump ha' s been isolated from the Reactor Coolant System by at least two isolation' valves with power removed from the valve motor operators.

M c Ced ant k 5 b WA maximum of one centrifugal har pump shall be OPERABLE whenever the temperature of one or more of cold legs is less than or equal to 3CO*F.

CATAWABA - UNIT 1 3/4 1-10

( REACTIVITY CONTROL SYSTEMS BORATED VATER SOURCE - SHUTDOWN PRODF & RMW COPgi i i

LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE: ,

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume of 5100 gallons,

2) A minimum boron concentration of 7000 ppm, and

3) A minimum solutf or temperature of 65'F.

b. The refueling water storage tank with:

1) A minimum contained borated water volume of 26,000 gallons,

2) A minimum baron concentration of 2000 ppm, and

3) A minimum solution temperature of 70*F.

APPLICABILITY: MODES 5 and 6. .

ACTIC ri.

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

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

a. At least once per 7 days by:

1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and

3) Verifying the boric acid storage tank solution temperature when it is the source of borated 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 it is the source of borated water and the outside air temperature is less than 70*F.

CATAWABA - UNIT.1 3/4 1-11 b

,,4. .- ---

REACTIVITY CONTROL SYSTEMS {g g ((Af ))

C BORATED WATER SOURCE - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum, the following borated water source (s) shall be OPERABLE as required by Specification 3 1.2.2:

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 tank with:

1) A contained borated water volume of at least 350,000 gallons

_ . _ _ . . . _ _ _ - '-o' _ ' ^ ~~ 1,

2) A minimum boron concentration of 2000 ppm,

( 3) 4)

A minimum solution temperature of 70*F, and A maximum solution temperature of 100*F.

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

ACTION:

a. WiththeBoricAciN-StorageSysteminoperableandbeingusedasone 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 SHUTDOWN MARGIN equivalent to at least 1% ak/k at 200*F; restore the Boric Acid Stot ige System to OPERABLE status within the next 7 days or be in COLD SHUTOOWN within i the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , i

. b. With the refueling water storage tark 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 leas +. 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 CATAWABA - UNIT 1 3/4 1- 12

PR00F & RMW COPY

{ 140 -

'm

. W 3

~

120 8+o n

<4 m2

. u. m OW 100 U'NA'CCEPTABLE #z

UNACCEPTABLE OPERATION OPERATION N11'90) _ __ '(11.90) 80 / \

60

( N 2

'"81,5 0)i

/ ACCEPTABLE OPERATION i'

' T-

.(X31)50) 40 _ _

20 7--+ :

II~ Milli IIii ..T. _.

i=.i.i..i.iM. . .i.i.i..iE.E.

iiill!!!Iliiifri Ii555 --

50 40 30 -20 10 0 10 20 30 40 50 FLUX DIFFERENCE (41) %

FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS A5 A FUNCTION OF RATED THERMAL POWER CATAWBA - UNIT 1 3/4 2-4

POWER DISTRIBUTION LIMITS 3/4.2.2 HEATFLUXHOTCHANNELFACTOR-Fg LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) q shall be limited by the following relationships:

! F0 (Z) 1 [2.32] [K(Z)] for P > 0.5 P

Fq (Z) i [4.64] [K(Z)] for P 1 0.5

, and Where: P _ THERMAL POWER RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: MODE 1.

ACTION:

With Fq (Z) exceeding its limit: 2

( a.

Reduce THERMAL POWER at least 1% forq each 1% F (Z) exceeds within 15 minutes and similarly reduce the Power Range Neutron <

te Flux-High Trip 5etpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints have been reduced at least 1% for each 1% F q (Z) exceeds the limit, and

b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced x ACTION a.1;, above; THERMAL POWER may then,be increased ,1Jait-required-by provided Fq (Z) is demonstrated through incore ma inQobewithinitslimit 3

/

i L

l CATAWBA - UNIT 1 3/4 2-5 1 .. _ _ . - _ ._

i I

PR00F & REVIEW COPY POWER DISTRIBUTION LIMITS 3/4.2.3 REACTOR COOLANT SYSTEM FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System total flow rate and R shall be maintained within the region of allowable operation shown on Figure 3.2-3 for four loop operation.

Where: N Fg a.

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

THERMAL POWER , and b*

P = RATED THERMAL POWER

c. Fh=MeasuredvaluesofFhobtainedbyusingthemovableincore detectors to obtain a power distribution map. The meas.ured valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 includes penalties for undetected feedwater venturi fouling of 0.1% and for measurement uncertainties of 1.9% for flow and 4%

forincoremeasurementofFh. .

APPLICABILITY: MODE 1.

ACTION:

1. With the combination of Reactor Coolant System total flow rate and R outsideQhearegion of acceptable operation shown on Figure 3.2-3:3

~

N_)

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 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 .

L CATAWBA - UNIT 1 3/4 2-9

g TABLE 3.3-2

--s REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES d ,

c FUNCTIONAL UNIT RESPONSE TIME 5

1. Manual Reactor Trip N.A.

2. Power Range, Neutron Flux $ 0.5 second"

3. Power Range, Neutron Flux. -

High Positive Rate N.A. m

o

4. Power Range, Neutron Flux, O c.

High Negative Rate 1 0.5 second* n Go

5. Intemediate Range, Neutron Flux N.A.

m, r.

$ 6. Source Range, Neutron Flux N.A. 5 rn

7. Overtemperature AT $ 4 seconds

l 8. Overpower AT -<4sebndsk -o v

9. Pressurizer Pressure-Low 1 2 seconds .-

l

10. Pressurizer Pressure-High < 2 seconds

! 11. Pressurizer Water Level-High N.A.

e

Neut.ron detectors are exempt from response time testing. Response time of the neutron flux signal portion of ti.e channel shall be measured from detector output or input of first electronic component in channel.

- w- w.

( .(, ; -

V) ,

f')

V' n

> TABLE 3.3-3 (Continued)

U ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION Ei MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE 4 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w

8. Auxiliary Feedwater

a. Ma'nual Initiation 1 1 1 1,2,3 22

b. Automatic Actuation Logic and Actuation Relays 2' 1 2 1,2,3 21

c. Sta. Gen. Water Level- .

w Low-low ,

A :::a c:)

w 1) Start Motor-g Driven Pumps 4/sta gen. 2/stm. gen. 3/stm. gen. 1, 2, 3 19*

M in any opera- in each g ting stm. gen operating stm. gen. M f $

rs

2) Start Turbine- F G Driven Pump 4/stm. gen. 2/stm. 3/stm. gen. 1,2,3 19*

in a a

R in each C ope stm.

tincf e :,

operating stm. gen. Q

d. Safety Injection-Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions and requirements.

e. Loss-of-Offsite Power-Start Motor-Driven Pumps and Turbine-Driven Pump 6-3/ bus 2/ bus 2/ bus 1, 2, 3 19*

either bus

f O O '

, TABLE 3.3-3 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5

MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE Z FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION y

8. Auxiliary Feedwater (Continued)

f. Trip of All Main Feedwater Pumps- ~

Start Motor-Driven Pumps 2/ pump 1/ pump 1/ pump 1, 2# 14

~c

g. Auxiliary Feedwater w Suction Pressure-Low 6-3/ pump 2/ pump 2/ pump 1, 2, 3 22 c3 1 9 4 w 9. Containment Sump Ro 4 Recirculation m

a. Automatic Actuation 2 1 2 1,2,3,4 21 2 Logic and Actuation 1

Relays h

ca

b. Refueling Water Storage Tank Level-Low 4 . 2 3 1,2,3,4 16 O

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

! and requirements.

10. Loss of Power 4 kV Bus Undervoltage-Grid Degraded Voltage 3/ Bus 2/ Bus 2/ Bus 1, 2, 3. 4 15*

11. Control Room Area 4

Ventilation 4ae&aMes> Operm[ ton

a. Manual Initiation 2 1 2 All 18

f' O '

y y {, dove & a(( So Infechen int]iaktag &ne] Ions o nck ~

n 3

d. Sd*h L eclion i T i ,.e.,,a d c. TABL'E 3.3-3 (Continued)

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 5

e MINIMUM E TOTAL NO. CHANNELS CHANNELS APPLICABLE Q FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE H0 DES ACTION w

11. Control Room Area Ventilation 4eenottuseQersfion (Continued)

b. Automatic Actuation Logic and Actuation Relays 2 1 2 All 14

c. Loss-of-Offsite Power R

3 2 2 1,2,3 19* y

12. Containment Air Return and

] Hydrogen Skimmer Operation @ n M a. Manual Initiation RO 2 1 2 1,2,3,4 18 .

r <1

b. Automatic Actuation Logic and Actuation Relays 2 1 2 1,2,3,4 21 3

c. Containment Pressure- c3 High-High 4 2 1,2,3 _]

3 16

13. Annulus Ventilation Operation -<

a. Manual Initiation 2 1 2 1,2,3,4 18

b. Automatic Actuation Logic and Actuation Relays 2 1 2 1,2,3,4 21

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

14. Nuclear Service Water Operation

a. Manual Initiation 2 1 2 1,2,3,4 18

TABLE 3.3-4 9

y ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION TRIP SETPOINTS E.

2 SENSOR E TOTAL ERROR Q FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE s.*

1. Safety Injection (Reactor Trip, Phase "A" Isolation, Feedwater

Isolation, Auxiliary Feedwater-

Motor-Driven Pump, Purge &

+

Exhaust Isolation, Annulus Ventilation Operation Cwdrol Room Ace Y'4N*b*^ Ope n Auxiliary Buildin9 O **84 Ventilatiohlsolation,Emer- 4 Fhd y

gency Diesel Generator Opera- d tion, Component Cooling Water, c3

. Turbine Trip, and Nuclear 9

} Service Water Operation) Ro w :C j a. Manual Initiation N.A. M.A. N.A. N.A. N.A.

b. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A. h c:3

c. Containment Pressure-High 3.0 5 1.2 psig 0.71 1. 5 5 1.37 psig y

d. Pressurizer Pressure-Low 12.5 10.71 1.5 1 1845 psig 1 1835 psig

e. -Steam Line Pressure-Low 24.6 10.71 1.5 1 710 psig 1 671 psig*

2. Containment Spray (Nuclear Service Water Operation)

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

b. AutomaticActuationLkic N.A. N.A. N.A. N.A. N.A.

and Actuation Relays

c. Containment Pressure-High-Nigh 3.0 0.71 1. 5 5 3 psig 5 3.17 psig

TABLE 3.3-4 (Continued)

U '

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

> SENSOR TOTAL ERROR l g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE "4

g 10. Loss of Power 4 kV Bus Undervoltage- N.A. N.A. N.A. 3500 1 175 > 3200 volts Grid Degraded Voltage volts with a 8.5 1 0.5 second time delay

11. Control Room Area Venttiation ~

h Operah.n w :=.3

} a. Manu:' Initiation N.A. N.A. N.A. N.A. N.A. c.-.J c3 Y '

b. Automatic Actuation Logic U and Actuation Relays N.A. N.A. N.A. N.A. N.A. @

t3

c. Loss-of-Offsite Power N.A. N.A. N.A. N.A. N.A. 2

~

. onta n A etu and *g"*#*

'* *N

'hN'E9 F" b **# M*"*N

. Hydrogen Skimmer Operation a co

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 ll c. Containment Pressure- 3.0 0.71 1.5 <3 psig <3.17 psig Nigh-High O

( TABLE 3.3-5 i

PROSF & REVIEW COPY ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation

a. Safety Injection (ECCS) N.A.

b. Containment Spray' N.A.

c. Phase "A" Isolation N.A.

d. Phase "B" Isolation N.A.

Purge and Exhaust Isolation t

e. N.A.

f. Steam Line Isolation N.A.

g. Diesel Building Ventilation Isolation N. A.

h. Auxiliary Feedwater N.A.

1. Nuclear Service Water Operation N.A.

j. Turbine Trip N.A.

k. Component Cooling Water N.A.

1. .

Annulus Ventilation Operatio g ,,f; , N.A.

(

m.

Control Room Area Ventilatic 2::' t(23 N. A.

n. Auxiliary Building, Ventilation3 Isolation N.A.

o. Reactor Trip CH*nA 64=d N.A.

p. Emergency Diesel Generator Operation N.A.

q. Containment Air Return and Hydrogen Skimmer Operation N.A.

2. Containment Pressure-High

a. Safety Injection (ECCS). 1 27(1)

1) Reactor Trip i2

2) Feedwater Isolation <7

3) Phase "A" Isolation (2) 1 18(3)/28(4)

4) Purge and Exhaust Isolation N.A.

5) Auxiliary Feedwater(5) N.A.

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

7) Turbine Trip N.A.

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

9) Emergency Diesel Generator Operation i 11 CATAWBA - UNIT 1 3/4 3-36

,O-

TABLE 3.3-5 (Continued) i - PR00F & REV!EW C ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

?. Containment Pressure-High (Continued) i

10) Annulus Ventilation Operation

11) Auxiliary BuildinggVentilation f M N ShF MS,A.

Isolation g ,J

12) Containment Sump Recirculation N.A.

3. Pressurizer Pressure-Low

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

1) Reactor Trip <2

2) Feedwater Isolation <7

3) Phase "A" Isolation (2) 18(3)/28(4)

4) Purge and Exhaust Isolation N.A.

5) Auxiliary Feedwater(5) N.A.

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

7) Turdine Trip N.A.

( 10) 8)

9)

Component Cooling Water Emergency Diesel Generator Operation Annulus Ventilation Operation i 76(1)/65(3) 1 23 g5

11) Auxiliary Building, Ventilation -

geh G W Isolation gpgg $

AI A. c

12) Containment Sump Recirculation fs

{[2 pb t

e b1

4. Steam Line Pressure-Low Q9\ % b

(\h'p4

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

1) Reactor Trip , 12 ,

2) Feedwater Isolation <7 .

3) Phase "A" Isolation (2) } #18)(3)/428%I4) l 4) Purge and Exhaust Isolation N.A.

5) Auxiliary Feedwater(3) < 60

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

7) Turbine Trip N.A.

8) Component Cooling Water 1 65(3)/76(4) l 9) Emergency Diesel Generator Operation < 11

10) Annulus Ventilation Operation L

CATAWBA - UNIT 1 3/4 3-37 6ambe M O A'

TABLE 4.3-2 h ENGINEERED SAFETY FEATtIRES ACTUATION SYSTEM INSTRUMENTATION h SURVEILLANCE REQUIREMENIS TRIP E ANALOG ACTUATING MODES U CHANNEL DEVICE MASTER SLAVE FOR WHICH CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

1. Safety Jnjection (Reactor Trip, Phase "A" Isolation, Feeduater Isolation, Auxiliary Feedwater-Motor-Driven Pump, Purge and y Exhaust Isolation, Annulus m C3 Ventilation Operation, Auxiliary Building Ventilation Operation.,Confrel R a Arc. Vedl.lt.a Ope d .n @

w Emergency Diesel Generators Opera- g 1 tion, Component Cooling Water, D w Turbine Trip, and Nuclear n, 1 Service Water Operation) $

t -t

.g'

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

b. Automatic Actua- N.A. N.A. M.A.

6 N.A. M(1) M(1) 1, 2, tion Logic anJ Q 3Qg Actuation Relays

c. Containment S R M N.A. N. A. N.A. N.A. 1, 2, 3

~

Pressure-High

d. Pressurizer S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low

e. Steam Line 5 R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low I

2. Containment Spray (Nuclear Service Water Operation)

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

b. Automatic Actua- N.A. M.A. N.A. N.A. M(1) M(I) Q 1,2,3,4 tion Logic and Actuation Relays -

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

~

t

} b ~ '

TABLE 4.3-2 (Continued) 9

-* ENGINEERED SAFETY FEA10RES ACTUATION SYSTEN INSTRUNENTATION SURVEILLANCE REQUIRENENIS I

TRIP ANALOG ACTUATING NODES E CNANNEL DEVICE NASTER SLAVE FOR WHICH

-! Q CHAISEL CNAleIEL CNAleIEL j g FINICTIonAt inIIT CNECK CAL 18 RATION TEST OPERATIONAL OPERATIONAL ACTUATION TEST LOGIC TEST TEST RELAY

~

RELAY SURVEILLANCE TEST 15 REQUIRED

5. Feedseter Isolatten (Continued)

b. Stede Generater -S R N N.A. N.A. N. A. N.A. 1, 2 h tet level-Nigh-Nigh (P-14)

c. T ,,-Lew (P-4 Interlock) S R N N.A. M.A. N.A. N.A. 1, 2

d. Dogheese W ter AAA. j~

t' Level-Nigh # #y' M I N.A. N.A.

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

N.A. 1, 2

]

I A 6. Turbine Trip pa

a. Manual Initiatfen N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2 ::x3 i, b. Automatic M.A. N.A. M.A. M.A. N(1) N(1) Q 1, 2 d

Actuation Logic and Actuation Relays C'3 q <

c. Steam Generater Su j S R N N.A. N.A. N. A. N.A. 1, 2 i Water Level-Nigh-Nigh (P-14) W !

d. Trip of All Main N.A. N.A. N.A. R N.A. N.A. N.A. 1,2 ;

Feessater Pumps

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

7. Centainment Pressure

Centrol System j a. Start Permissive S R N N.A. N.A. N.A. N.A. 1, 2, 3, 4 ;

I

b. Termination 5 R N N.A. N.A N.A. N.A. 1, 2, 3, 4 1

, - - - m_, ,,ec.

TABLE 4.3-2 (Continued)

D ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION g , SURVEILLANCE REQUIREMENIS TRIP ANALOG ACTUATING MODES

~

E CHANNEL DEVICE MASTER SLAVE FOR WHICH O CHANNEL CHAleIEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

10. Loss of Power 4 kV' Bus N.A. /

N.A. effk N.A. N.A. N.A. 1, 2, 3, 4m R

Undervoltage-Grid Degraded Voltage 9 ,Q @

0

11. Control Room Area Ventilation, h

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. All m

b. Automatic Actuation M.A. N.A. N.A. N.A. M(1) M(1) Q All Q l 1 Logic and Actuation M '

l y Relays G

! A c. Loss-of-Of fsite M.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3 C7 h

p d.'*"'_L-scG h 1+ i.. .bw<. &c,U kiely %1aN, Amd/ece Ago.cowe/r.

. 12. Containnent Airje% turn ia b l and .",_". +.7 Skimmer

! Operation

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. M.A. 1, 2, 3, 4 l b. Automatic N.A. N.A. M.A. N.A. M(1) N(1) Q 1, 2, 3, 4 8

Actuation Logic and Actuatim Relays -

c. Containment S R M N.A. M.A. N.A. N.A. 1,2,3 i . Pressure-High-Nigh

13. Annulus Ventilation Operation

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

=

1 PR007& RElmy copy TABLE 3.3-6 (Continued) a TABLE NOTATION $

a With fuel in the fuel storage pool areas.

With irradiated fuel in the fuel storage pool areas.

Must satisfy the requirements of Specification 3.11.2.1.

ACTION $7ATEMENT$

ACTION 26 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge and exhaust valves are maintained closed.

ACTION 27 - With the number of operable channels one less than the Minimpm Channels 0PCRABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate theg adie' fem Control Room Ventilation System " ' 't' n _a '- - r y.. _ . .....-__ _ _ _ _ _ Mia ke from oulsde air unfh recircddwg floul %roujh Os HCPA fikers and ebrs..

ACTION 28 - With less than the Minimum C'hannels CPLRAOLE requirement, opera- reded 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 CPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel building.

ACTION 29 - Must satisfy the ACTION requirement for $pecification 3.4.6.1.

ACTION 30 - With the number of OPERABLE channels less than the Minimum Channels CPERABLE requirement, operation may continue provided -

the fuel Handling ventilation Exhaust System is nperating and discharging through the HEPA ft1ters and charcoal adsorbers.

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

ACTION 31 - With the number of OPERA 8LE channels less than the Minimum Channels OPERABLE requirement, o ration may continue provided the Auxiliary 76MA Building} Ventilation yttem is operating and discharging through the HCPA filter and charcoal dsorbers.

Eshad' CATAWBA - UNIT 1 3/4 3 $2

i

( '"5'au"5"'^" o" SE!$MIC INSTRUMENTATION PROD & REV!EW CD?Y l f

LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be i CPERABLE.

i APPLTCABILITY: At all times, ACTION:

a. With one or more of the above required seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERA 8LE status.

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

l SURVEILLANCC REQUIREM(NTS 4.3.3.3.1 tech of the above required seismic monitoring instruments shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL CAL 1-BRATION and ANALOG CHANNEL OPERATIONAL TEST operations at the frequencies i

shown in Table 4.3 4. ,

ages.85tMt 4.3.3.3.2 Each of the above[see M eetsmic monitoring instruments actuated during a seismic event greatdr than or equal to 0.01 g shall be restored to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. Data retrieved from the triaxial time-history .

acceleregraph shall include a post event CHANNEL CALIBRATION obtained by actuation of the internal test and calibrate function immediately prior to removing data. CHANNEL CALIBRATION shall be performed immediately after insertion of the new recording media in the triamiel time-history accelero-graph recorder. A Special Report shall be prepared and submitted to the Commission pursuant to Specification 4.9.2 within 10 days describing the magnitude, frequency spectrus, and resultant effect upon facility features important to safety.

L CATAWSA = UNIT 1 3/4 3 55

. 4. . ~. . - .. .. . . . ,

PRODF & REWEty CDPy EBLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE

1. Triaxial Time-History Accelerographs

a. 1MIMT 5070 (Remote Sensor A) -1 g to + 1 g 1 Containment Base Slab

b. 1MIMT 5080 (Remote Sensor B) -1 g to + 1 g 1 Containment Vessel Elev 619'5"

c. 1MIMT 5090 (Starter Unit) 0.005 g to 0.05 g 1 Containment Base Slab

2. Triaxial Peak Accelerographs o

a. 1MIMT 5010 - Containment Bldg. kgto+2g 1 Elev 613'8 9/16" o

b. 1MIMT 5020 - Containment Bldg. \ g to + 2 g 1 Elev 567'2h"

c. IMIMT 5030 - Auxiliary B1dg. '

g to + 2 g 1 Elev 543'

3. Triaxial Seismic Switch ,

IMINT 5000 - Containment 0.025 g to 0.25 g 1*

Base Slab

4. Triaxial Response-5pectrum Recorders

a. 1MIMT 5040 - Containment 'O to 34 g at la Base Slab 2 to 25 Hz

b. 1MIMT 5050 - Containment Bldg. O to 34 g at 1 .

Elev 579'34" 2 to 25 Hz

c. 1MIMT 5060 - Auxiliary Bldg. O to 34 g at 1 Elev 577' 2 to 25 Hz i

s

With reactor control room indication.

CATAWBA - UNIT 1 3/4 3-56

f O O

n TABLE 3.3-10 E ACCIDENT MONITORING INSTRUMENTATION TOTAL MINIMUM I

NO. OF CHANNELS E INSTRtNENT CHANNELS OPERABLE n

w 1. Containment Pressure 2 1

2. Reactor Coolant Outlet Temperature - THOT (Wirle Range) 2 1

3. Reactor Coolant Inlet Temperature - TCOLD (Wik Range) 2 1

4. Reactor Coolant Pressure - Wide Range 2 1

5. Pressurizer Water Level 2 1

6. Steam Line Pressure 2/ steam generator 1/ steam generator y 7. Steam Generator Water Level - Narrow Range 2/ steam generator 1/ steam generator e

8. Refueling Water Storage Tank Water Level 2 1

9. Auxiliary Feedwater Flow Rate 2/steambeoerator 1/ steam generator _- t3

10. Reactor Coolant System Subcooling Margin Monitor d

co

/ -i / I

11. PORY Flow Indicator

2/Valife 1/ Valve kM

x2

12. PORY Block Valve Position Indicator ** 2/ Valve 1/ Val've g

13. Pressurizer Safety Valve Position Indicator 2/ Valve 1/ Valve h f 14. Containment Sump Water Level (Wide Range) 2 1 Q t3 e

1 f b '

n TA8LE 3.3-10 (Continued)

D

> ACCIDENT MONITORING INSTRUMENTATION 6

, TOTAL MINIMUM NO. OF CilANNELS

!! E INSTRLSENT CHANNELS OPERABLE

-4 w 15. In Core Thermocouples /0 g/ core quadrant Y $ core quadrant

16. Unit Vent - High Range Noble Gas Monitor (EMF .%) 1 1

17. Steam Relief Valse Exhaust Monitor 1/ steam line 1/ steam line

18. Containment Atmosphere - High Range Monitor 2 1

19. Reactor Vessel Water Level 2 1 '

20. Reactor Coolant Radiation Level (EMF-48) / ,{ 1 Y

3 i

'o' C"3 i C3 TABLE NOTATIONS m i

Qo

Not applicable if the associated block valve is in the closed position. p

- Not appilcable if the associated block valve is in the closed position and power is removed. @_.

M J -

4 c3 l C:3

-o l.

1 I -

TABLE 4.3-7 (Continued) ,

U ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

" CHANNEL CHANNEL INSTRtMENT (Continued) CHECK CALIBRATION E

'I M 15. In Core Thermocouples M R

16. Unit Vent - High Range Noble Gas Monitor (ENF -% M R

17. Steam Relief Valve Exhaust Monitor M R

18. Containment Atmosphere - High Range Monitor M R*

19. Reactor Vessel Water Level M R

20. M R Reactor Coolant Radiation Level (EMF- 48)

R

~. '

Y E $

a CD

CHANNEL CALIBRATION may consist of an electronic calibration of the channel, not including the detector, for range decades above 10R/h and a one point calibration check of the detector below 10R/h with an Q0 installed or portable gamma source. y 5

LO

-=

cm CD tJ3 M

i l

f b '

n TABLE 4.3-8 D

g RADIOACTIVE LIQUID EFFLUENT MDNITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS s

$ ANALOG ,

CHANNEL k

-4 INSTRt#ENT CHANNEL CHECK SOURCE CHECK CHANNEL CALIBRATION OPERATIONAL TEST w

1. Radioactivity Monitors Providing Alarm and Automatic Termination of- Release

a. Waste Liquid Discharge Monitor (Low Range - D P R(3) Q(1)

EMF-49)

b. Turbine Building Sump Mor.itor (Low Range - D M R(3) Q(1) '

EMF-31)

2. Radioactivity Monitors Providing Alarm But '

y Not Providing Automatic Termination u

of Release rz

c2

a. Nuclear Service Water System Effluent Line (EMF-45 A&B, H&L)

D M R(3) Q(2) $

-r1 Ro

b. Component Cooling Water System Effluent m Line (EMF-46 A&8) D M R(3) Q(2) e-i 3. Continuous Composite Samplers and Sampler Flow Monitor i/,A .

]

c3 Conventional Waste Water Treatment Line D N.A. R

/ $

I t 4. Flow Rate Measurement Devices

a. Waste Liquid Effluent Line D(4) N.A. R

b. Conventional Waste Water Treatment D(4) N.A. R Mg

c. Low Pressure Service Water Minimum Flow D(4) N.A. R Q Interlock

r o 6 -

TABLE 4.3-9

> RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS

. 5>

, ANALOG CHANNEL MODES FOR WHICH k-4 CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE INSTRt#ENT CHECK CHECK CALIBRATION TEST IS REQUIRED i H

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor -

Providing Alarm and Automatic Termination of Release P P R(3) Q(1) ***

(Low Range - EMF-50 Low Range -EMF-36) g

b. Effluent System Flow Rate Measuring Device P N.A. R 4[.A ***

a w 2. WASTE GAS MOLOUP SYSTEM Explosive h Gas Monitoring System $

C3

a. Mydrogen Monitor (Recombiner N.A. **

Outlet) ,

D Q(4) M

b. Oxygen Monitors (Recombiner Outlet)

D N.A. Q(5) M **

y

.c:::

FR

3. Condenser Evacuation System G c3 Noble Gas Activity Monitor D M R(3) Q(2) 1, 2, 3, 4 C3 (Low Range - EMF-33)

Q

4. Vent System

a. Noble Gas Activity Monitor D M R(3) Q(2) *

(Low Range - EMF-36)

b. Iodine Sampler (EMF-37) W N.A. N.A. N.A.

i '

n TABLE 4.3-9 (Continued)

D

> RADIOACTIVE GASEOUS EFFLUENT NONITORING INSTRUMENTATION "fRVEILLANCE REQUIREMENTS E

( ANALOG CHANNEL MODES FOR WHICH e

z CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE

-* INSTRtMENT CHECK CHECK CALIBRATION TEST IS REQUIRED ea -

! 4. Vent System (Continued)

c. Particulate Sampler (EMF-35) W N.A. N.A. N.A. *

d. Flow Rate Monitor D N.A. R [g), *

e. Sampler Flow Rate Monitor 0 N.A. R Q

5. Containment Purge System j w I, 1 Noble Gas Activity Monitor -

1' e Providing Alarm and Automatic cIn

iermination of Release D P R(3) Q(1) ***

m l

(Low Range - EMF-3?, Low Range - EMF-36) g ra

' mt

6. Containment Air Release and D P R(3) Q(1) ***

Addition System-Providing Alarm and Automatic Termina- m,r r ,

tion of Release (Low Range - $

p i

EMF-36) Q O

C T3 W

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.1.1 Each Cold Leg Injection Accu:nulator System shall be OPERABLE with:

a. The discharge isolation valve open,

b. A contained borated water volume of between 7743 and 7965 gallons

c. A boron concentration of between 1900 and 2100 ppm,

d. A nitrogen cover pressure of between 400 and 454 psig, and

e. A water level and pressure channel OPERABLE.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

a. With one Cold Leg Injection Accumulator System inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With one Cold Leg Injection Accumulator System inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS - 4.5.1.1.1 Each Cold Leg Injection Accumulator System shall be demonstrated OPERABLE:

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

1) Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and

2) Verffying that each cold leg injection accumulator isolation valve is open.

" Pressurizer pressure above 1000 psig.

CATAWBA - UNIT 1 3/4 5-1 h-

i !

4

(

I

! EMERGENCY CORE COOLING SYSTEMS j d' k [ [, N{}g !

l f SURVEILLANCE REQUIREMENTS (Continued) i :

j g 75 p u.am l

. b. At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution f

} _. ... volume increase of greater than or equal to :^.~ J i . _ J by ^

verifying the boron concentration of the accumulator solution; ;

l l c. At least once per 31 days when the Reactor Coolant System pressure l

! is above 2000 psig by verifying that power to the isolation valve j operator is disconnected by removal of the breaker from the circuit; ;

j and .

i d. At least once per 18 months by verifying that each cold leg I injection accumulator isolation valve opens automatically under i each of the following conditions:

l l

- 1) When an actual or a simulated Reactor Coolant System pressure

} signal exceeds the P-11 (Pressurizer Pressure 31ock of Safety (

j Injection) Setpoint, and ;

I j 2) Upon receipt of a Safety Injection test signal. !

) 4.5.1.1.2 Each Cold Leg Injection Accumulator System water level and pressure :

channel shall be demonstrated OPERABLE: !

s. At least once per 31 days by the performance of an ANALOG CHANNEL l I, OPERATIONAL TEST, and :

t I b. At least once per 18 months by the performance of a CHANNEL CALIBRATION, 1 .

l !

8 Y

0 0 i

4 CATAWBA - UNIT 1 3/4 6-2

~

s . c. ._: -

_ ~_ _ _ :~ ~ ~ ~ <

( EMERGENCY CORE COOLING SYSTEMS UPPER HEAD INJECTION I Pt hhh I

LIMITING CONDITION FOR OPERATION 3.5.1.2 Each Upper Head Injection Accumulator System shall be OPERABLE with:

i a. The discharge isolat'fon valves open, -

b. A minimum contained borated water volume of 1807 cubic feet,

c. A boron concentration of between 1900 and 2100 ppa, and

~

d. The nitrogen-bearing accumulator pressurized to between 1206 and 1264 psig.

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

ACTION:

a. WiththeUpperHeadInjectionAccumulatorSysteminoperable,except as a result of closed isolation valve (s), restore the Upper Head Injection Accumulator System to OPERA 8LE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least NOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. WiththeUpperHeadInjectionAccumulatorSysteminoperabledueto the isolation valve (s) being closed, either immediately open the isolation valve (s) or be in HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS

4. 5.1. 2 EachUpperHeadInjectionAccumulatorSystemshallbedemonstrated OPERA 8LE:

s. At least once per 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

1) Verifying the cent borated water level in the surge tank and nitrogen press the accumulators, and

2) Verifying that each accumulator discharge isolation valve is open.

Pressurizer pressure above 1900 psig.

CATAWBA - UNIT 1 3/4 5 3

- - ~ ~ -~~

l . .

l l

( EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 REFUELING WATER STORAGE TANK li.*h,2(( h,h7[

LIMITING CCNDITION FOR OPERATION 3.5.4 The refueling water storage. tank shall be OPERA 8LE with:

a. A minimum contained borated water volume of 350,000 gallons 4964

b. A boron concentration of between 2000 and 2100 ppa of boron, I c. A minimum solution temperature of 70'F, and l

l d. A maximum solution temperature of 100'F.

APPLICABIt,!TY: MODES 1, 2, 3, and 4.

ACTION:

With the refueling water storage tank inoperable, restore the tank to OPERABLE ststus within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l

_SURVE!LLANCE REQUIREMENTS 4.5.4 ,The refueling water storage tank shall be demonstrated CPERABLE:

a. At least once per 7 days by:

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

2) Verifying the beren concentration of the water, 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 stora e b.

tank temperature when the outside air temperature is less than 70 F er greater than 100*F.

t l

l CATAWBA = UNIT 1 3/4 5 11

r j__ "

CONTAINMENT SYSTEMS bu' c

s SURVEILLANCE REOUIREMENTS (Continued) l c. After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, l within 31 days after removal, that a laboratory analysis of a repre-l 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 l penetration of less than 1%;

d. At least once per 18 months by: brs and enois b re 588 i

1) Verifying that the pressure drop across the combined iEPA filters,en6 charcoal adsorber banks is less than inches Water Gauge while operating the system at a flow rate of 9000 cfm i 10%;

2) Verifying that the system starts automatically on any Phase "A" Isolation test signal,

3) Verifying that the filter cooling electric motor-operated bypass valves can be manually opened,

4) Verifying that each system produces a negative pressure of greater than or equal to 0.5 inch Water Gauge in the annulus within 1 minute after a start signal, and

5) Verifying that the pre-heaters dissipate 45 i 6.7 kW when tested in accordance with ANSI N510-1975. . ,

e. After each complete or partial replacement of a HEPA filter bank, by - ^

verifying that the cleanup system satisfies the in place penetra-tion and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1975 for a DOP test aerosol while operating

-the system at a flow rate of 9000 cfm i 10%; and

f. After each complete or partial replacement of a charcoal adsorber -

bank, by verifying that the cleanup system satisfies the in place e penetration and bypass leakage testing acceptance criteria of less -

than 1% in accordance with ANSI N510-1975 for a haloge~nated hydro-: ' -

carbon refrigerant test gas while operating the system"at=a flow , '

N rate of 9000 cfm i 10%.

t e

1 I,s i.

%. i h'

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- [ .

4-CATAWBA - UNIT 1 3/4 6-15 ,

s,%_

~~

- ,. . l 1-- -. . L ------= -'

P n TABLE 3.6-2 (Continued)

D g CONTAINMENT ISOLATION VALVES 2

[ MAXIMUM VALVE NUMBER FUNCTION ISOLATION TIME (Sec.)

1. Phase "A" Isolation (Continued)

KC-3058# Excess Letdown Hx Supply Containment Isolation (Outside) $20 KC-315B# Excess Letdown Hx Return lleader Containment Isolation (Outside) 520 KC-320A# NCOT Hx Supply Hdr Containment Isolation (Outside) $20 KC-3328# NCOT Hx Return Hdr Containment Isolation (Inside) $20 KC-333A# NCDT Hx Return Hdr Containment Isolation (Outside) $20 KC-4298 RB Drain Header Inside Containment Isolation <10 KC-430A RB Drain Header Outside Containment Isolation 510 m NB-260B Reactor hakeup Water Tank to Flush lleader $10 u '

s ::o I NC-53B Nitrogen to Pressurizer Relief Tank #1 Containment Isolation outside $10 @

4 NC-54A Nitrogen to Pressurizer Relief Tank #1 Containment Isolation Inside $10 m w NC-568 RMW Pump Olsch Cont Isolation $10 no NC-195B NC Pump Motor 011 Containment Isolation Outside $10 m NC-196A NC Pump Motor 011 Containment Isolation Inside $10 g NF-228A Unit 1 Air Handling Units Glycol Supply Containment Isolation Outside <10 C NF-233B Unit 1 Air llandling Units Glycol Return Containment Isolation Inside 710

'G NF-234A Unit 1 Air llandling Units Glycol Return Containment Isolation Outside 310 Q u 6 " :- :..;_ _:_.. !_.f 1 : .._ 1_ ": ' f ' 7 64h *<

N ":.;._ ::.j::- !_./ _:.. 1: "-!d i_,_ @-. _;

NI-47A Accumulator N 2 Supply outside Containment Isolation slo NI-95A Test Hdr Inside Containment Isolation <10 HI-968 Test Hdr Outside Containment Isolation 710 NI-120B* Safety Injection Pump to Accumulator Fill Line Isolation 510 NI-1228# llot Leg Injection Check 1NI124, INI128 Test Isolation $10 NI-154B# Hot Leg Recirculation Check 1NI125, 1NI129 Test Isolation $10 NI-2558# UHI Check Valve Test Line Isolation <10 NI-258A# UHI Check Valve Test Line Isolation 710 NI-2648 Ull! Check Valve Test Line Outside Containment Isolation 510

r o 6 -

TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES 5

2

. MAXIMUM j

E c-.

VALVE NUMBER FUNCTION ISOLATION TIME (Sec.)

e 1. Phase "A" Isolation (Continued) i NV-11A 45 gpa Letdown Orifice Outlet - Containment Isolation $10 NV-13A- 75 gpa Letdown Orifice Outlet - Containment Isolation $10

. NV-10A High Pressurizer Letdown Orifice Outlet - Containment Isolation 110 NV-872A Standby Makeup Pump to RCS seals $10 N -. . . . : _ . . . . . .m i. r a i. n,i i . . . . ; ..,, _ 2 : " " -- "; - ' - " ' ' - - -

4se4st4evr9ptTNo- 4 m _ _ _ .

.. . c _ , _ _ .

__7 _ _ _

R N

i i

Di RF-3898 Interior Fire Protection Containment Hose Rack Isolation Valve (Outside Containment) 15 E

=3 RF-4478 Reactor Building Sprinklers Containment Isolation Valve 15 @

(Outside Containment) - ri

!?o VB-838 Breathing Air Unit 1 Containment Isolation $10 y rn VY-188** Containment H 2 Purge to Annulus Inside Containment Isolation <10 "'E VY-17A** Containment H 2 Purge to Annulus Outside Containment Isolation 710 C']

VY-ISB** Containment H 2 Purge Blower Outlet, Containment Isolation (Outside) 710

^

c .>

VI-312A RB Isolation Valve for VI Supply to annulum Vent. 110 0 -;

VP-1B** Upper Containment Purge Supply #1 Outside Isolation 15 .

VP-2A** . Upper Containment Purge Supply #1 Inside Isolation $5 VP-3B** Upper Containment Purge Supply #2 Outside Isolation 55 VP-4A** Upper Containment Purge Supply #2 Inside Isolation 55 VP-6B** Lower Containment Purge Supply #1 Outside Isolation 15 VP-7A** Lower Containment Purge Supply #1 Inside Isolation 15 VP-8B** Lower Containment Purge Supply #2 Outside Isolation $5 VP-9A** Lower Containment Purge Supply #2 Inside Isolation 55 VP-10A** Upper Containment Purge Exhaust #1 Inside Isolation 15

P A A -

TABLE 3.6-2 (Continued) 9 C CONTAINMENT ISOLATION VALVES

5 MAXIMUM E VALVE NUMBER FUNCTION ISOLATION TIME (Sec.)

Z s 2. Phase "B" Isolation (Continued) 446 4384 -

^ " ^'

_ . . _; " . _ . . ". ,. . _, ' ' . ; ' ^ _ . _ f - !

N @

m .

- - - . 7_ .;. . . .. g , . ; 9 m _

s c..... , s -. in.. .syg y RN-437B Supply to NC Pumps and LCVU Supply Outside Containment Isolation 160 RN-484A Return from NC Pumps and LCVU Return Inside Containment Isolation 160 RN-487B Return from NC Pumps and LCVU Return Outside Containment Isolation 160 '

RN-404B Supply to Upper Containment Supply Ventilation Units Containment ~<10 a R

RN-429A Isolation (Outside)

Return from Upper Containment Ventilation Units Containment Isolation -<10 ca T (Inside) 8 O RN-432B Return from Upper Containment Ventilation Units Containment Isolation 110 Qo (Outside) :: .s tr:

VI-77B Instrument Air Containment Outside Isolation $10 h als SM-1 # Main Steam ID Isolation <5 i

SM-3 #

SM-5 #

Main Steam IC Isolation Main Steam IB Isolation 55

<5 O

7J SM-7 # Main Steam 1A Isolation . 75 "

SM-9 # Main Steam ID Isolation Bypass Ctr1. 75 L -- J.

SM-10 # Main Steam IC Isolation Bypass Ctri. 35 SM-11 # Main Steam IB Isolatinn Bypass Ctri. <5 SM-12 # Main Steam 1A Isolation Bypass Ctri. 55

,' SV-19 # Main Steam 1A PORV <5 SV-13 # Main Steam IB PORV 75 SV-7 # Main Steam IC PORY 75 SV-1 # Hain Steam 10 PORV 75 WL-867A** Containment Vent Unit Drains Inside Containment Isolation ~10 WL-869B** Containment Vent Unit Drains Outside Containment Isolation 510

f O O TABLE 3.6-2 (Continued)

CONTAINMENT ISOLATION VALVES e

c HAXIMUM

$ VALVE NUMBER FUNCTION ISOLATION TIME (Sec.)

-e H 3. Manual

, NC-141 NC Pump H2 Drain Tank Pump Discharge N.A.

NC-142 .

2 N . A. ..

(-W-06ef/ [NCPump-HDrainTankPump-Discharge P:7. f. 3.q Tr =kntLE:_e 4rA,.

?; NI-3' s Boron. Injection Tank-Line to Cold legs N. Ac' FW-11 Refueling Water Pump Suction N.A.

FW-13 Refueling Water Pump Suction N.A.

CF-91# Feedwater IA N.A.

CF-93# Feedwater 18 N.A.

m CF-95# Feedwater 1C N.A.

N CF-97# Feedwater ID N.A. o CA-121# Aux. Feedwater 1A l

I 4 BW-1# Aux. Feedwater IA N.A.

N.A.

to m

CA-120# Aux. Feedwater IB N. A. ~n BW-26# Aux. Feedwater IB N.A. Ro CA-119# Aux. Feedwater IC N.A. .a BW-17# Aux. Feedwater IC N.A. 02 CA-118# Aux. Feedwater ID N.A. _,-]

BW-10# Aux. Feedwater ID N.A. q SM-16# Main Steam 1A N.A.

SM-73# Nain Steam 1A N.A. F;-5 SM-105# Hain Steam 1A . N.A. ~i3 SM-121#

Main Steam 1A N.A.

SM-143#

Main Steam 1A N.A. - - . .

SM-72# Main Steam IB N.A.

SM-104# Main Steam IB N.A.

SM-120# Main Steam IB N.A.

SM-142# Main Steam IB N.A.

SH-1# Main Steam IB N.A.

SM-17# Main Steam IB N.A.

I SH-18# Main Steam IC N.A.

SM-71# Main Steam IC N.A.

PLANT SYSTEMS . a SURVEILLANCE REOUIREMENTS (Continued)

3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in its correct position; and

4) Verifying that each automatic valve in the flow path is in the fully open position wh9never the Auxiliary Feedwater System is placed in automatic control or when above 10% RATED THERMAL POWER.

5) Verifying that the isolation valves in the auxiliary feedwater pump suction lines are open and that power is removed from the valve operators on valves CA-2, CA-7A, CA-98, and CA-11A.

b. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of an Auxiliary Feedwater Actuation test signal, and

2) Verifying that each auxiliary feedwater pump starts as designed

( 3) automatically upon receipt of an Auxiliary Feedwater Actuation test signal.

Verifying that the valve in the suction line.of each auxiliary feedwater mp from the Nuclear Service Water System automatically actuates o i s full open position within less than or equal to 15 +,)lf sec ds*o a Loss-of-Suction test signal.

4.7.1.2.2 An auxilia eedwater flow path to each steam generator shall be l demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days

! prior to entering MODE 2 by verifying normal flow to each steam generator.

l i

l hn, i,-uwau,.

CATAWBA - UNIT 1 3/4 7-5

PLANT SYSTEMS

{]w'lt ,'{ r). ,Lpt *.- y y ;y

.w u ,: 1 3/4.7.6 CONTROL ROOM AREA VENTILATION staitn LIMITING CONDITION FOR OPERATION 3.7.6 Two independent Control Room Area Ventilation Systems shall be OPERABLE.

APPLICABILITY: ALL MODES ACTION:

MODES 1, 2, 3 and 4:

With one Control Room Area Ventilation System inoperable, restore the inoperaolo system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following i

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

M0GES 5 and 6:

a. With one Control Room Area Ventilation System inoperable, restore the inoperable system to OPERABLE status within 7 days or initiate O and maintain operation of the remaining OPERABLE Control Room Area-Ventilation System in the recirculation mode.

I .

b. With both Control Room Area Ventilation Systems inoperable, or with the OPERABLE Control Room Area Ventilation System, required to be in the recirculation mode by ACTION a., not capable of being powered by an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or, positive reactivity changes.

? .

l .-

SURVEILLANCE REQUIREMENTS - c.

4.7.6 Each Control Room Area Ventilation System shall be demonstrated OPERABLE:

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air temperature is less than or equal toQ g

b. At least once per 31 days on a STAGGERED TEST BASIS 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; l CATAWBA - UNIT 1 3/4 7-13

t

{ h 6

PLANT SYSTEMS l-SURVEILLANCE REQUIREMENTS (Continued)

c. At least once per 18 months or (1) after any structural maintenar i t 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:

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 Regulatory Position C.S.a. C.S.c, and C.S.d of Regulatory Guide 1.52, Revi-sions 2, March 1978, and the sytem flow rate is 6000 cfm i 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance 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, Revi-sion 2, March 1978, for a methyl iodide penetration of w less than 1%; and

3) Verifying a system flow rate of 6000 cfm 1 10% during system operation when tested in accordance with ANSI N510-1975.

d. 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 1%;

e. At least once per 18 months by: g (88i 5hre 3df"tdY'Sg

1) Verifying that the pressure drop across the combined EPA filters W charcoal adsorber banks is less than nches Water Gauge while operating the system at a flow rate of 6000 cfm + 10%;

2) Verifying that on a Loss-of-Offsite Power, or High Radition-Air '

Intake, or Smoke Density-High test signal, the system automatically switches into a recirculation mode of operation with flow through the HEPA filters and charcoal adsorber banks;

3) Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere during system operation;

4) Verifying that the heaters dissipate 2512.5 kW when tested in accordance with ANSI N510-1975; and

5) Verifying that en a High Chlorine / Toxic Gas test signal, the system automatically isolates the affected intake from outside air with recirculacing flow through the HEPA filters and char-( coal adsorbers banks within 10 seconds.

CATAW8A - UNIT 1 3/4 7-14 a a e 4 as -

( PLANT SYSTEMS pl'# / #PR00F & iBF *'"'!

3/4.7.7 AUXILIARY BUILDING, VENTILATION, SYSTEM n n LIMITING CONDITION FOR OPERATION Vil4< red Ehusl^

3.7.7 The Auxiliary Building, Ventilation System j shall be OPERABLE.

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

gdQMA gxlwak With the Auxiliary Building 3Ventilation gystem inoperable, restore the inoperable system to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least HOT 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 .

SURVEILLANCE REQUIREMENTS g;\}er*A Svh*b

( 4.7.7 The Auxiliary Building Ventilation OPERABLE:

a.

g System 3 shall be demonstrated At least once per 31 days by initiating, from the, control room, flow l through the HEPA filters and charcoal adsorbers and verifying that the j system operates for at least 10 continuous hours with the heaters operating; i 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:

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 Regula-

! tory Positions C.S.a. C.5.c, and C.S.d of Regulatory Guide 1.52, l Revision 2, March 1978, and the system flow rate is 30,000 cfm i 10%;

2) Verifying, within 31 days after removal, that a laboratory analysis of a representative carbon sample obtained in accor-dance 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, Revi-sion 2, March.1978, for a methyl iodide penetration of less than 1%; and CATAWBA - UNIT 1 3/4 7-16

I PLANT SYSTEMS I : L . 1 SURVEILLANCE RE0VIREMENTS (Continued)

3) Verifying a system flow rate of 30,000 cfm + 10% during system operation when tested in accordance with ANSI N510-1975.

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%;

p,

d. At least once per 18 months by: g pis/d g

1) Verifying that the pressure. drop cross the combined HEPA filters W charcoal adsorber banks of less than inches Water Gauge while operating the system at a flow rate of 30,000 cfm + 10%, and

2) Verifying that the system starts on a Safety Injection or Loss of-Offsite Power test signal, and directs its exhaust flow through the HEPA filters and charcoal adsorbers,

3) Verifying that the system maintains the ECCS pump room at a negative pressure of greater than or equal to 1/8 inch water gauge relative to ,the outside atmosphere,

4) Verifying that the filter cooling bypass valves can be manually opened, and

5) Verifying that the heaters dissipate 30 1 3 kW when tested in accordance with ANSI N510-1975.

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place pene- .

tration and bypass leakage testing acceptance criteria of less than 1% in accordance with ANSI N510-1975 for a DOP test aerosol while l 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 bank, 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 N510-1975 for a halogenated hydro-carbon refrigerant test gas while operating the system at a flow l rate of 30,000 cfm i 10%.

CATAWBA - UNIT 1 3/4 7-17

3/4.8 ELECTRICAL POWER SYSTEMS PR00F & REM COPY 3/4.8.1 A.C. SOURCES OPERATING -- -

LIMITING CONDITION FOR OPERATION 3.8.1.1 As a minimum, the following A.C. electrical power sources shall*be OPERABLE:

a. Two physically independent circuits between the offsite transmission network and the Onsite Essential Auxiliary Power System, and-

b. Two separate and independent diesel generators, each with:

1) A separate day tank containing a minimum volume of 518.5 gallons of fuel,

2) A separate Fuel Storage System containing a minimum volume of 82,056 gallons of fuel, and

3) A separate fuel transfer valve.

l APPLICABILITY: MODES 1,2,3Iand4k ACTION:

( a. With either an offsite circuit or diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica--

tions 4.8.1.1.la. and 4.8.1.1.2a.4) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 'and at least once l

per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least two offsite circuits and two diesel generators to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With one offsite circuit and one diesel generator of the above required A.C. electrical power sources inoperable, demonstrate the OPERABILITY of the remaining A.C. sources by performing Specifica-tions 4.8.1.1.la. and 4.8.1.1.2a.4) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter; restore at least one of the -

inoperable sources to OPERABLE status within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HDT 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 . . Restore at least two offsite circuits and two diesel generators to OPERA 8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from the time of initial loss or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . "

c. With one diesel generator inoperable in addition to ACTION a. or b.

j above, verify that: ,

1. All required systems. subsystems, trains, components and devices that depend on the remaining OPERABLE diesel generator as a source of emergency power are also OPERABLE, and

. m ,we y.m,w muay.

CATAWBA - UNIT 1 3/4 8-1

ELECTRICAL POWER SYSTEMS C A.C. SOURCES Fi100F & REViEV1 COPY SHUT 00WN LIMITING CONDITION FOR OPERATION 3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

a. One circuit between the offsite transmission network and the Onsite Essential Auxiliary Power System, and

b. One diesel generator with:

1) A day tank containing a minimum volume of 518.5 gallons of fuel,

2) A fuel storage system containing a minimum volume of 82,056 gallons of fuel, and

3) A fuel transfer valve.

APPLICABILITY: MODES and6f ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System through a greater than or equal to 4.5 square inch vent. In addition, when in MODE 5 with the Reactor Coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS .

4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requirements of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.8.1.1.2a.5), 4.8.1.1.3, and 4.8.1.1.4.

N g N4 mt,cau, pec h inthaluikidy.

CATAWBA - UNIT 1 3/4 8-10

~

ELECTRICAL POWER SYSTEMS ;

PROOF m & gay nnny

.an

. 1

!. 3/4.8.3 ONSITE POWER DISTRIBUTION OPERATING LIMITING CONDITION FOR OPERATION t

3.8.3.1 The following A.C. electrical busses and inverters shall be OPERABLE

, end energized with tie breakers open between redundant busses: -

a. 4160-Volt Essential Bus #1 ETA,

b. 4160-Volt Essential Bus #1ETB,

c. 600-Volt Essential Bus #1ELXA, t

d. 600-Volt Essentia1' Bus #1ELXB, 3 e. 600-Volt Essential Bus #1ELXC, '

f. 600-Volt Essential Bus #1ELXD, l

g. 120-Volt A.C. Vital Bus # IERPA energized from Inverter # 1EIA connected to D.C. Channel 1,*

h. 120-Volt A.C. Vital Bus # 1ERPB energized from Inverter # 1EIB connected to D.C Channel 2,*

1. 120-Volt A.C. Vital Bus # IERPC energized from Inverter # 1EIC connected to D.C. Channel 3,*

( j. 120-Volt A.C. Vital Bus # 1ERP0 energized from Inverter # 1EID connected to D.C. Channel 4.*

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

ACTION:

a. Witpie I than Dtheove complement of A.C. busses OPERA 8LE and e rgized restore 4he inoperable busses to OPERABLE and energized tu 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months k be in at least H0T STAN08Y within the next hours and in CQLD within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

lw%;s

b. Wi ane inverter i erable, energize the associated A.C. vital bus within4 , restore the inoperable inverter to OPERABLE and -

energized status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in at least H0T STAND 8Y 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 .

l SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified A.C. busses and inverters shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

r .

An inverter may be disconnected from its D.C. source for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for f the purpose of performing an equalizing charge on its associated battery bank

( provided: (1) its vital bus-is 0PERABLE and energized, and (2) the vital busses associated with the other battery banks are OPERABLE and energized.

An inverter may be disconnected from its 0.C. source for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the conditions of ACTION c. of Specification 3.8.2.1 are satisfied.

CATAWBA - UNIT 1 3/4 4-16

f O '

h TABLE 3.8-1 CONTAlleqENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES j -

1 RIP

--e SETPOINT OR RESPONSE w CONT. RATING TIME SYSTEM DEVICE NUMBER & LOCATION (AMPERES) (SECONOS) POWERED

1. 6900~VAC Swgr .

Primary Skr RCPIA 5.0 42 + 4.2 9 ISA Reactor Coolant Pump 1A Backup Skr ITA-3 6.0 27 + 2.7 9 18A '

y

23 Primary 8kr RCPIB 5.0 42 + 4.2 9 ISA Backup Skr 1T8-3 6.0 27 + 2.7 9 18A Reactor Coolant Pump IB Q

mi t

Primary SKR RCPIC 5.0 42 + 4.2 9 15A Reactor Coolant Pump IC e Backup Skr ITC-3 6.0 27 + 2.7 9 18A M

" l

, Primary 8KR RCPID 5.0 42 + 4.2 9 15A Reactor Coolant Pump ID I-U Backup Skr ITD-3 6.0 27 + 2.7 9 18A G

2. 600 VAC MCC a

Sy IEMMC-F02C -<

Primary Skr 20 45 9 60A Cont Isol at 134 Deg "le 5ER.T Backup Fuse 20 N..A. Annulus Area Viv IVI312A t

N 1EMMC-F03A gg Primary Skr Backup Fuse 20 20 45 9 60A NC Pump IC Thermal Barrier Outlet i N.A. Isol Viv IKC345A Q B-Q 1EMXC-F038 Primary 8kr 20 45 9 60A N, to Prt Cont Isol Inside i

Backup Fuse i 20 N.A. Viv INC54A i l

O O O .

g , TABLE 3.8-1 (Continued) 5 CONTAllSENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES E

q TRIP y SETPOINT OR RESPONSE CONT. RATING TIME SYSTEM DEVICE laster & LOCATION (AWERES) (SECONOS) POWERED

2. 600 VAC MCC (Continued) 1EMNL-F10C Primary Skr 20 45 9 60A Reactor Vessel Head Vent Viv Backup Fuse 20 N.A. INC252B !

l IEMNL-Fila . l y Primary Skr 125 110 0 375A Containment Air Return

Backup Fuse 125 as N. A. Fan Motor 18 0

1EMNL-F118 Primary Skr l '

Backup Fuse 125 125 110 9 375A N.A.

Hydrogen Skimmer Fan Motor IB

{%c:)

IEMNS-F018

'g ic Primary Okr 20 45 0 60A NC Pumps Seal Rtn #

3,ckup Fuse 20 N.A. Inside Cont Isol Viv INV89A $

6 IEMNS-F02A Primary Skr

~ 'r+3 20 45 9 60A NO Pump 18 $; fon from NC 2 Backup Fuse 20 N.A. Loop C V1v IN037A cm 1EMNS-F028 Primary Skr 20 45 9 60A Reactor Vessel Head Vent Viv l

i Backup Fuse 20 N.A. INC250A #

IEMNS-F03C Primary 8kr 20 45 9 60A NO Pump 1A fon from NC Backup Fuse 20 N.A. Loop B Viv 2A

h O ~

9 TABLE 3.8-1 (Continued) g CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES E TRIP Q SETPOINT OR RESPONSE g CONT. RATING TIME SYSTEM DEVICE NUMBER & LOCATION (AMPERES) (SECONDS) POWERED

2. 600 VAC MCC (Continued)

IMXZ-F07D Primary Bkr 30 45 9 90A Reactor Cavity Manipulator Backup Fuse 30 N.A. Crane No. R007 & R027 INXZ-F08A t* Primary Bkr 20 45 9 60A Steam Generator Drain Pump

Backup Fuse 20 N.A. Motor 1

?

g IMXZ-F08C _

Primary Bkr 30 45 @ 90A 15 Ton Equipment Access flatch I Backup Fuse 30 N.A. Holst Crane No. R009 m CD IMXZ-F080 Primary Bkr 20 45 9 60A Control Rod Drive 2 Ton Jib S

Backup Fuse 20 g

N.A. Holst Crane No. R017 r -1 IMXZ-F08E =C Primary Bkr 20 45 @ 60A Reactor Side fuel llandling Backup Fuse 20 N.A. Control Console Q

4 CD SMXG-F01C Cf3 Primary Bkr m3 20 45 9 60A Standby Hakeup Pump Drain Isol -.< l Backup Fuse 20 N.A. Viv INV876 SMXG-F05C Primary Bkr 100 1I0 0 300A Pressurizer lleaters 28, 55 & 56 Backup Fuse 100 H.A. -

/ -'t /

SMXG-F06A Primary Bkr 20 45 @ 60A Standby Hakeup Putnp to' Seal Backup Fuse -

20 N.A. Water Line Isol Viv IV877

~

S TABLE 3.8-1 (Continued)

Y f CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES E TRIP U SETPOINT.OR RESPONSE N CONT. RATING TIME SYSTEM DEVICE NUMBER & LOCATION (AMPERES) (SECONOS) POWERED

2. 600 vat MCC (Continued) 1EMXC-F01B Primary Bkr 50 110 9 150 Accumulator IC Discharge Backup Fuse 50 N.A. Isoly Vivg INI76A w IEMXC-F01C c3 D Primary Bkr 20 45 e 60 Check Valve Test Ileader y e Backup Fuse 20 N.A. Cont Isol Viv INI95A a a

" r1 1EMXC-F02A po Primary Bkr 20 45960 Train A Alternate Power oc,e Backup Fuse 20 N.A. To ND LTDN Vivy INDIB h on IEMXC-F028 Primary Bkr 20 45960 Q

p Backup Fuse 20 N.A.

f - ^- INI153A 3M g.20 3. 600 VAC Pressurizer Heater Power Panels N 5 Tak '

1

-Tist Isol Vlv PHPIA-F01A Primary Bkr 90 110 9 270A Pressurizer lleaters Backup Fuse 90 N.A. 1, 2, & 22

9 TABLE 3.8-1 (Continued)

Y

> CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES e

I i

E TRIP Q SETPOINT OR RESPONSE p

CONT. RATING' TIME SYSTEM DEVICE NUMBER & LOCATION (AMPERES) (SECONOS) POWERED

3. 600 V,AC Pressurizer Heater Power Panels (Continued)

PHP1A-F018 Primary Bkr 90 110 0 270A Bac se Pressurizer Heaters ,

90 N.A. 5, 6, & 27

[-SIC w Pri1 fiery Bkr 90 110 9 270A Pressurizer Heaters c3

) ackup Fuse 90 N.A.

e 9, 10, & 32

{$

CD 1

PHPIA-F02C ~M Primary Bkr 90 110 9 270A Backup Fuse Pressurizer Heaters f20 90 N.A. 11, 12, & 35 m PHPIA-F020 Primary Bkr 90 110 9 270A Backup Fuse Pressurizer Heaters q 90 N.A. 13, 14, & 37 C7 PHPIA-F02E Primary Skr 90 110 9 270A ks M

Backup Fuse Pressurizer Heaters 90 N.A. 17, 18, & 42 PHP18-F01A Primary Bkr 90 110 @ 270A Backup Fuse Pressurizer Heaters S9 N.A. 21, 47 & 48 9

f D D '

9 TABLE 3.8-1 (Continued)

CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES E TRIP Q SETPOINT OR RESPONSE w CONT. RATING TIME SYSTEM DEVICE NUMBER & LOCATION (AMPERES) (SECONDS) POWERED 3.

600 VAC Pressurizer Heater Power Panels (Continued)

PHP18-F018 Primary Bkr 90 110 @ 270A Backup Fuse Pressurizer Heaters 90 N.A. 26, 53 & 54 PHP18-F01C ~n

cJ w Primary Bkr 90 110 @ 270A l Backup Fuse Pressurizer Heaters @

90 N.A. 31, 59 & 60 -

en

.A PHPIB-F02C .

f2o Primary Bkr 90 110 @ 270A Backup Fuse Pressurizer Heaters F].-

90 N.A. 36, 65 & 66 f:

PHP18-F020 C9 Primary Bkr 90 110 @ 270A Backup Fuse Pressurizer Heaters C~J 90 N.A. 41, 71 & 72 @

PHP18-F02E -<:

Primary Bkr 90 110 0 270A Backup Fuse Pressurizer Heaters 90 N.A. 46, 77 & 78 PHPIC-F01A Primary Bkr 90 110 @ 270A Backup Fuse Pressurizer Heaters 90 N.A. 7, 8 & 30 rose.RT -+

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1,4l25 W

i REFUELING OPERATIONS 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS r- 3 LIMITING CONDITION FOR OPERATION 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, immediately 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 penetrations 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 irradiated fuel in the containment _ building by:

a. Verifyingthepenetragionsareintheirclosed/isolatedcondition-or )

b.

?"B" Verifying the containmentg isol tion valves close upon a High Relative

Humidity test signal, i -

CATAWBA - UNIT 1 3/4 9-4 l

I REFUELING OPERATIONS SURVEILLANCE REQUIREMENTS (Continued) 4.9.4.2 The Reactor Building Containment Purge System shall be demonstrated OPERABLE:

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 less than 1% and uses the test procedures guidance in Regula-tory Positions C.5.a. C.S.c, and C.S.d of Regulatory Guide 1.52, Revision 2, March 1978, and the system flow rate is 28,000 cfm i 10% (both exhaust fans operating);

Verifying within 31 days after removal, that a laboratory

(

2) 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 testing criteria of Regulatory Position C.6.a of Regulatory Guide 1.5.2 Revision 2, March 1978, for a methyl iodide penetration of less than 6%;

and -

3) Verifying a system flow rate of 28,000 cfm i 10% (both exhaust fans operating) during system operation when tested in accordance with ANSI N510-1975.

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%; y

d. At least once per 18 months by: g p=I p ea Sef*

1) Verifying that the pressure dro across the combined EPA filters W charcoal adsorber ban s is less than nches Water i

Gauge while operating the system at a flow rate of 28,000 cfm

10% (both exhaust fans operating);

2) Verifying that the filter cooling bypass valves can be opened j by operator action; and CATAWBA - UNIT 1 3/4 9-5 i

1

( REFUELING OPERATIONS i PR00F & H.T.',i C 3 i l

SURVEILLANCE REOUIREMENTS (Continued) i

2) Verifying, within 31 days after removal, that a laboratory 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 Regulatory Guide 1.52, Revision 2, i

March 1978, for a methyl iodide penetration of less than 1%;

and ,

3) Verifying a system flow rate of 18,000 cfm 10% during system operation when tested in accordance with ANSI N510-1975.

! 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 18 months by: g gets p s48'J,r5 g

1) Verifying that the pressure dro cross the combined HEPA filters M charcoal adsorber banks is less than inches Water Gauge while operating the system at a flow rate of jq 18,000 cfm 2 10%.

2) Verifying that the system maintains the spent fuel storage pool e

area at a negative pressure of greater than or equal to dd> inch

Water Gauge relative to the outside anosphere during system *

. operation, i

i

3) Verifying that the filter cooling bypass valves can be manually l opened, and

4) Verifying that the heaters dissipate 80 t 8 kW when tested in accordance with ANSI N510-1975.

e. After each complete or partial replacement of a HEPA filter bank, by verifying that the cleanup system satisfies the in place penetration and bypass leakage testing acceptance criteria of less than 1% in i accordance with ANSI N510-1975 for a D0P test aerosol while operating

[

the system,at a flow rate of 18,000 cfm i 10%; and i

f. After'each complete or partial replacement of a charcoal adsorbee bank, by verifying that the cleanup system satisfies the

/ in place penetration and bypass leakage testing acceptance criteria

( of less than 15 in accordance with ANSI N510-1975 for a halogenated hydrocarbon refrigerant test gas while operating the system at a flow !

rate of 18,000 cfm i 105.

CATAWBA - UNIT 1 3/4 9-15

a O REACTIVITY CONTROL SYSTEMS gu d EnG m.g.- v u q Jde \,

.e BASES BORATION SYSTEMS (Continued)

MARGIN from expected operating conditions of 1.3% ak/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 16,321 gallons of 7000 ppm borated water from the boric acid storage tanks or 75,000 gallons of 2000 ppm borated water from the refueling water storage tank.

With the coolant temperature below 200*F, one Baron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 300 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The baron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200*F to

( 140 F. This condition requires either 906 gallons of 7000 ppm borated water from the boric acid storage tanks or 3170 gallons of 2000 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the 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 minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Baron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that: .(1) accept'able power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is main-tained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is i required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital l Rod Position Indicator agrees with the demanded position within i 12 steps

at 24, 48,120 and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurances that the Digital r

Rod Position Indicator is operating correctly over the full range of '~- indication.

I c'- t': ^2 p t ' a:r ^: ' '-- ( - "--- - ' ' "' _ __ c ::^ -_7' .__

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CATAWBA - UNIT 1 8 3/4 1-3

I E

POWER DISTRIBUTION LIMITS Pr"(M & Ei,irgij g, BASES HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT i- CHANNEL FACTOR (Continued)

When Reactor Coolan IEhflowrateandFharemeasured,noadditional

{

allowances are necessar o comparison with the limits of NM g m 3.2-3.

Measurement errors of for otal flow rate and 4% for F# have been allowed for in determination of th design DNBR value.

Rea<h C.. lad Sydemi l

^

The measurement error for Reactor Coolant System total flow rate is based Lyon performing a precision heat balance and using the result to calibrate the Reactor Coolant System flow rate indicators. Potential fouling of the feedwater 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 i 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 i performance parameters. If detected, action shall_be taken before performing subsequent precision heat balance measurements, i.e., either the effect of the fouling shall be quantified and compensated fee 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 Coolant System flow is sufficient to detect only flow degradation which could lead to opera-

, tion outside the acceptable regior of operation shown on Figure 3.2-3.

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 periodically during power operation.

The limit of 1.02, at which a : tion is required, provides DN8 and linear heat generation rate pt 'h x y plane power tilts. A limit of 1.02 was selected to provt ance for.the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for a tilt condition greater than 1.02 but less than 1.09 is p- .. , identification and correction of a dropped or misaligned control . In ter svent such action does not correct the tilt, the margin for unce ainty on F isq reinstated by reducing the maximum allowed power by 35 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 the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore

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POWER DISTRIBUTION LIMITS hl100F & Ey,Ey my BASES QUADRANT POWER TILT RATIO (Continued) flux map or two sets of four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. Th g locations are C-8, E-5, E-11 H-3, H-13, L-5, L-11, N-8. n%gg 3/4.2.5 DNB PARAMETERS *** ** 8 " * *V*

norm loc. Mons are knavaildle .

")

The limits on tne DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the j initial FSAR assumptions and have been analytically demonstrated adequate to maintain a design limit DNBR throughout each analyzed transient.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation. Measure-ment uncertainties must be accounted for during the periodic surveillance.

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I .

CATAWBA - UNIT 1 8 3/4 2-6 L ._. _ _ _ . ~ . _ , _ __ _

REACTOR COOLANT SYSTEM {}Q]{ { }{y]y {}((

BASES 3/4.4.9 PRESSdRE/ TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

1. The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figuros 3.4-2 and 3.4-3 for the service period specified thereon:

a. Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown. Limit lines for cooldown rates between those presented may be obtained by interpolation; and

b. Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only. For normal operation, other inherent plant characteristics, e.g., pump heat aedition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

2. These limit lines shall be calculated periodically using methods provided below, .

3. The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F,

4. The pressurizer heatup and cooldown rates shall not exceed 100'F/h and 200'F/h, respectively, and

5. System'preservice hydrotests and in-service leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler -

and Pressure Vess 1 C e Section XI. 4 gg g;g.,n MT(6 s-2, .

The fracture toughness properties o the vessel are determined in accordance with the ee80=6ausser Addenda o Section III of the ASME Boiler and Pressure Vessel Code and the NRCg_- __. . ;.. .., 7...., J.; ~ 7 ^~ 7. and ,

in accordance with additional reactor vessel requirements. These properties are then evaluated in accordance with Appendix G of the 1971 Winter Addenda to Section III of the ASME Boiler and Pressure Vessel Code.

w CATAWBA - UNIT 1 8 3/4 4-7 .

REACTOR COOLANT SYSTEM PR00F & RB.!.H.' CON BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of nonductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs or a Reacter Coolant System vent opening of at least 4.5 square inches ensures that the Reactor Coolant System will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the cold legs are less than or equal to 300*F. Either PORV has adequate relieving capability to protect the Reactor Coolant System from overpressurization when the transient is limited to either:

(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.

3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection 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 accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 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).

Components of the Reactor Coolar.t System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Cod y ;;7; :_._:... .... i___..__ _.__,.. :.__. 1 %

as).npgUsdleA.LL**ula as te pire A by IO OR S0SGSCfexcefl*dert ,

spuhe ur$en retlef % s Leen gru W Q 4ke com arn u puer J g to cra. so.ss 3 gi(.(,) u k ,

CATAWBA - UNIT 1 B 3/4 4-15

~

F PLANT SYSTEMS -PR00F& RlV!EW CDPv] 8 j BASES STANDBY NUCLEAR SERVICE WATER POND (Continued) _

ThelimitationsonAinimumwaterlevelandmaximumtemperaturearebased on providing a 30-day cooling water supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommend-ations of Regulatory Guide 1.27, " Ultimate Heat Sink for Nuclear Plants,"

4 March 1974.

3/4.7.6 CONTROL ROOM AREA VENTILATION SYSTEM The OPERABILITY of the Control Room Area Ventilation System en'sures that:

4

' (1) the ambient air temperature does not exceed the allowable temperature for continuous-duty rating for the equipment and instrumentation cooled by this j system, and (2) the control room will remain habitable for operations personnel during and following all credible accident conditions. Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day i period is sufficient to reduce the buildup of moisture on the adsorbers and i

I HEPA filters. The OPERABILITY of this system in conjunctinn with control room design provisions is based on limiting the radiation exposure to personnel i

( occupying the control room to 5 rems or less whole body, or its equivalent.

This limitation is consistent with.the requirements of General Design Criteria 19 of Appendix A, 10 CFR Part 50. ANSI N510-1975 will be used as a procedural guide for surveillance testing.

gW pdtM 3/4.7.7 AUXILIARY BUILDING. VENTILATION. SYSTEM n n g;lyd _ desef l The OPERABILITY of the Auxiliary BuildingV4 entilationAystem ensures that radioactive materials leaking from the ECCS equipment within the auxiliary building following a LOCA are filtered prior to reaching the environment.

Operation of the system with the heaters operating for at least 10 continuous hours in a 31-day period is sufficient to reduce the buildup of moisture on the adsorbers and HEPA filters. The operation of this system and the resultant l effect on offsite dosage calculations was assumed in the safety analyses.

l ANSI N510-1975 will be used as a procedural guide for surveillance testing.

3/4.7.8 SNUB 8ERS i

All snubbers are required OPERA 8LE 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. ' -

CATAWBA - UNIT 1 8 3/4 7-4

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ML20083J988

Text

Dear Mr. Denton:

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