Text

Forwards marked-up Proposed Changes to Draft Tech Specs. Changes Clarify Certain Statements & Correct Errors
	ML20092N620
Person / Time
Site:	Catawba
Issue date:	06/29/1984
From:	Tucker H; DUKE POWER CO.
To:	Adensam E, Harold Denton; Office of Nuclear Reactor Regulation
References
	NUDOCS 8407030348
	Download: ML20092N620 (9)
	v • d • e

.

-g, DUKE POWER GOMPANY 1

P.O. Box 33180 CHARLOTTE, N.C. 28242 HALH. TUCKER TELEPIf0NE (704) 373-6531 veo,a,mmessen June 29,1984

m.,-

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I 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 Draft Technical Specifications

Dear Mr. Denton:

Please. find attached proposed changes to the Draft Technica1 Specifications

.for Catawba Unit 1.

These changes clarify certain statements and make corrections to errors presently contained in the Specifications.

Very truly yours, stL r ecs Hal B. Tucker RWO/rhs

' Attachment t

cc: Mr. James P. O'Reilly' Mr. Jesse L. Riley l

Regional Administrator Carolina Environmental Study Group i

U. S. Nuclear Regulatory Commission 854Henley Place Region II Charlotte, N. C. 28207 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30303 Palmetto Alliance 21351 Devine Street NRC Resident Inspector Columbia, South Carolina 28207 l

Catawba Nuclear Station Mr. Robert Guild, Esq.

Attorney-at-Law

+

P. 0. Box 12097 Charleston, South Carolina 29205 00 8407030348 840629 PDR ADOCK 05000413 A

PDR

TABLE 3.3-4 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS s>

SENSOR e

TOTAL ERROR g

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETFOINT ALLOWABLE VALUE

[

10.

Loss of Power

a.

4 kV Bus Undervoltage-Loss N.A.

N.A.

1 3500 V 2 3200 V of Voltage 36BS 36//

b.

4 kV Bus Undervoltage-N.A.

N.A.

1 S M 4-1 Grid Degraded Voltage 11.

Control Room Area Ventilation Operation R

a.

Automatic Actuation Logic

[

and Actuation Relays N.A.

N.A.

e U

b.

Loss of-Offsite Power N.A.

N.A.

1 3500 V 3 3200 V c.

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

12.

Containment Air Return and Hydrogen Skimw;r Operation a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

N. A.

and Actuation Relays c.

Containment Pressure-12.7 0.71

1. 5 5 3 psig

< 3.2 psig g

High-High 2

4lD

-.,n

ui TABLE 3.3-12 I

n g

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION E

MINIMUM e

CHANNELS c-OPERABLE ACTION 5*

INSTRUMENT w

1.

Radioactivity Monitors' Providing Alarm and Automatic Termination of Release Waste Liquid Discharge Monitor (Low Range - EMF-49) 1 40 a.

l b.

Turbine Building Sump Monitor (Low Ran e M Mf 31 1

42 2

c.

Steam Generator Water Sample Montt

( MF-34) 1 43 Low R

2.

Continuous Composite Samplers and Samp1 F

tilto Conventional Waste Water Treatment Line 1

42 Y

g 3.

Flow Rate Measurement Devices a.

Waste Liquid Effluent Line 1

41 b.

Conventional Waste Water Treatment Line 1

41 c.

Low Pressure Service Water Minimum Flow Interlock.

'l 41

~

/

Z ao

.[

TABLE 4.3-8 h

RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS' E

ANALOG CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL E

INSTRUMENT CHECK CHECK CALIBRATION TEST

-1 s

1.

Radioactivity Monitors Providing

/

Alarm and Automatic. Termination of Release a.

Waste Liquid Discharge Monitor (Low Range -

D P

R(2)

Q(1)

EMF-49)

~

b.

Turbine Building Sump Monitor (Low Range -

D M

R(2)

Q(1)

EMF-31) w c.

Steam Generator Water Sample nitor (EMF-34)

D M

R(2)

Q(1)

A w

2.

Continuous Composite Samplers and mpler [ow k ogt,-

5 Flow Monitor u

N Conventional Waste Water Treatment Line D'

N.A.

R Q

3.

Flow Rate Measurement Devices a.

Waste Liquid Effluent Line D(3)

N.A.

R Q

b.

Conventional Waste Water Treatment Line D(3)

N.A.

R Q

c.

Low Pressure Service Water Minimum Flow D(3)

N.A.

R Q

Interlock c:z CD

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 accumulator shall be OPERABLE with:

The discharge isolation valve open, a.

b.

A contained borated water volume of between 78 4 and 817-

allons, c.

A boron concentration of between 19 100 pp d.

A nitrogen cover pressure of between 49fi ig, and e.

A water level and pressure channel OPERABLE.

APPLICABILITY: MODES 1, 2, and 3*.

ACTION:

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

-of a closed isolation valve, restore the-inoperable accumulatcr 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 followin'g 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , Withonecoldleginjectionaccumula[6p?inoperableduetothe-

e...

b.

isolation velve 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 SHUTDOWN 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 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)

Verifying that each cold leg injection accumulator isolation valve is open.

i

Pressurizer pressure above 1000 psig.

1 CATAWBA - UNIT 1 3/4 5-1 JUN 20 E64

CONTAINMENT SYSTEMS

~

CONTAINMENT LEAKAGE

~~ ~

~~

~

LIMITING CONDITION FOR OPERATION 3.6.1.2 Containment leakage rates shall be-limited to:

a.

An overall -integrated leakage rate of:

1)

Less than or equal to L,, 0.20,.

y weight of the containment air per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s-at P,, 14 psi, or

. /4 2.

2)

Less than or equal-to L, 0.Ja6% by weight of the containment t

air per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at a reduce ressure of P 7.3 psig.

t b.

A combined leakage rate of less than 0.60 L, f,or all penetrations and valves subject to Type B and C tests, when pressurized to P,,

and A combined bypass leakage rate of less than 0.07 L, for all c.

penetrations identified in Table 3.6-1 as secondary containment bypass leakage paths when pressurized to P,.

APPLICABILITY:

MODES 1,'2, 3, and 4.

ACTION:

With:

(a) the measured overall integrated containment leakage rate exceeding 0.75 L,or 0.75 L, as applicable, or (b) the measured combined leakage rate for all t

penetrations and valves subject to Types B and C tests exceeding 0.60 L,, or (c) the combined bypass leakage rate exceeding 0.07 L,, restore the overall integrated leakage rate to less than 0.75 L, or less than 0.75 L, as applicable, t

and the combined leakage rate for all penetrations and valves subject to Type B and C tests to less than 0.60 L,, and the combined bypass leakage rate to l

less than 0.07 L prior to increasing the Reactor Coolant System temperature 8

l above 200*F.

(

SURVEILLANCE REQUIREMENTS 4.6.1.2 The containment leakage rates shall be demonstrated at the following test schedule and shall be determined in conformance with the criteria speci-fled in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI l

N45.4-1972 or the mass plot method:

i l

t CATAWBA - UNIT 1 3/4 6-2 JUN 22 M L

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

Three Type A tests (Overall Integrated Containment Leakage Rate) a.

shall be conducted at 40 10 month intervals during shutdown at either P,14.7 psig, or at P, 7.3 psig, during each 10 year service a

t period.

The third test of each set shall be c.onducted during the shutdown for the 10 year plant inservice inspection; 1

b.

If any periodic Type A test fails to meet either 0.75 L, or 0.75 L '

t the t'est schedule for subsequent Type A tests shall be reviewed and s

\\3 approved by the Commission.

If two consecutive Type A tests fail to

,g meet either 0.75 L, or 0.75 L, a Type A test shall be performed at t

least every 18 months until two consecutive Type A tests meet either f

0. 75 L, or 0. 75 Lt at which time the above test schedule may be Q

resumed; yW c.

The accuracy of each Type A test shall be verified by a supplemental 4%$y test which:

N 1)

Confirms the accuracy of the test by verifying that the cupp h-

tal t :: t m'. ^., L' th

= Of th' TEO A t h*'

c

( f

{;

c d 1::E, L, i s e q"e! 1^ r t 2 ^ r r t + 0. 25 L, es--

g g,

h D

2) 8%

Has a duration sufficient to establish accurately the change in leakage rate between the Type A test and the supplemental test; CY d and 8 4 h

3)

Requires that the rate at which gas is injected into the 2

containment or bled from the containment during the supple-g4 O mental test is between 0.75 L, and 1.25 L, or 0.75 L and t

1.25 L

t d.

Type B and C tests shall be conducted with gas at a pressure not less than P,, 14.7 psig, at intervals no greater than 24 months except for tests involving:

1)

Air locks, 2)

Purge supply and exhaust isolation valves with resilient material seals, and 3)

Dual ply bellows assemblies on containment penetrations between the containment building and the annulus.

e.

The combined bypass leakage rate shall be determined to be less than 0.07 L,for penetrations which are not individually testable; per.etra-by applicable Type B and C tests at least once per 24 months except tions not individually testable shall be determined to have no detect-able leakage when tested with soap bubbles while the containment is pressurized to P,, 14.7 psig, or P, 7,3 psig, during each Type A test; t

CATAWBA - UNIT 1 3/4 6-3

PLANT SYSTEMS I

l l

SURVEILLANCE REQUIREMENTS (Continued) 3)

Verifying that ea&i:wn-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 whenever 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 and that the respective circuit breakers are padlocked.

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 u eceipt of an Auxiliary Feedwater Actuation test sign, and

<n.tur MrNen 2)

Verifying that each auxiliary fee ater pump starts as designed g

~

automatically u receipt Auxiliary Feedwater Actuatica test signal.

~

p Verifying that the valve in the suction line of each auxiliary feedwater pump from the Nuclear Service Water System automatically actuates to its full open position within less than or equal to 15 seconds

on a Loss-of-Suction test signal.

-4.7.1.2.2 An auxiliary feedwater flow path to each steam generator shall be demonstrated OPERABLE following each COLD SHUTDOWN of greater than 30 days prior to entering MODE 2 by verifying normal flow to each steam generator.

Includes 5 second time delay.

Ye yin O ad 4he du-rbine cd rNeo Rayi((cu-ceduX f~

P $r 5 b su d valves oren -ron rec 4f c# "n Aniham Teedwcler I< dud *n bs4 sia""l CATAWBA - UNIT 1 3/4 7-5 e-eee

DRA..-..

ADMINISTRATIVE CONTROLS r

6.1 RESPONSIBILITY I

6.1.1 The Station Manager shall be responsible for overall unit opera-tion and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor (or during his absence from the control room, a designated individual) shall be responsible for the control roo function. A management directive this effect% signed by th ":..,;r Nuclear Production shall be rei sued to all station )er onnel o an annual b uil car fndaMico h,g ce [re'EIca 6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for unit management and technical support l

shall be as shown in Figure 6.2-1.

UNIT STAFF 6.2.2 The unit organization shall be as shown in Figure 6.2-2 and:

Eac'h on-duty shift shall be composed of at least the minimum shift a.

crew composition shown in Table 6.2-1; b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room; c.

A Health Physics Technician

shall be on site when fuel is in the reactor; d.

All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation; e.

A site Fire Brigade of at least five members

shall be maintained on site 't all times. The Fire Brigade shall not include the three a

members of the minimum shift c'rew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and "The Health Physics Technician and Fire Brigade composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to accommodate unexpected absence, provided immediate action is'taken to fill the required positions.

l CATAWBA - UNIT 1 6-1

.,.

ML20092N620

Text

Dear Mr. Denton:

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