ML17331A519

Proposed Changes to Tech Specs 3/4 for Unit 1,incorporating Category a Lessons Learned Automatic Initiation Requirements Into Engineered Safety Feature Actuation Sys

ML17331A519
Person / Time
Site:	Cook
Issue date:	12/10/1980
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17331A518	List: ML17331A517 ML17331A519 ML17331A520
References
AEP:NRC:00449, AEP:NRC:449, NUDOCS 8012180317
Download: ML17331A519 (29)

Text

ATTACHMENT 2 TO AEP:NRC:00449 Proposed Technical Specifications Changes for Unit 1

PLANT SYSTEMS

~

'%URVEILLANCE RE UIREMENTS b.

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 10Ã RATED THERMAL POWER.

At least once per 18 months during shutdown by:

l.

Verifying that each automatic valve in the flow path actuates to its correct position upon receipt of the appropriate Engineered Safety Features actuation test signal required by Specification 3/4.3.2.

2.

Verifying that each auxiliary feedwater pump starts as designed automatically upon receipt of the appropriate Engineered Safety Features actuation test signal required by Specification 3/4.3.2.

D. C.

COOK - UNIT 1 3/4 7-6

n TABLE 3.3-3 ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUHENTATION

~

~

CD CD7' FUNCTIONAL UNIT SAFETY INJECTION,FEEDWATER ISOLATION AND MOTOR DRIVEN AUXILIARY FEEDWATER PUHPS a.

Hanual Initiation b.

Automatic Actuation Logic c.

Containment'ressure-High

.d.

Pressurizer Pressure

- Low e.

Differential Pressure Between Steam Lines - High Four Loops Operatinq Three Loops Operating TOTAL NO.

CHANNELS OF CHANNELS TO TRIP HINIMUH CHANNELS OPERABLE 2

2 2

2.

3 n

4e 2

2, 2

3/steam line 3/operating, steam line 2/steam line 2/steam line any steam line 1

/steam 2/operating line, any steam line operating steam line'PPLICABLE HODES ACTION I

)*

14 1,2',34-i 14 1,2,3 1, 2, 3N 15

\\

.,1,2,3,4 I

1&

1,2N 3N4 '3

TABLE 3.3-3 Cont'd.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 6.

MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS TOTAL NO.

CHANNELS OF CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES ACTION a.

Steam Generator Water Level-Low-Low 2/Stm.

Gen.

3/Stm.

Gen.

Any Stm.

Gen.

2/Stm.

Gen.

1, 2, 3 14*

b.

4 kv Bus Loss of Voltage c.

Safety Injection d.

Loss of Main Feedwater Pumps 2/Bus 2/Bus 2

2/Bus 1, 2, 3

1, 2, 3 1, 2, 3 le~

~

14'k 7.

TURBINE DRIVEN AUXILIARYFEEDWATER PUMPS.

a.

Steam Generator Water Leyel-Low-l ow 2/Stm.

Gen.

2/Stm.

Gen.

3/Stm.

Gen.

Any 2 Stm. Gen..

1, 2, 3

. b.'. Reactor Coolant Pump Bus Undervoltage 4-1/Bus 2

1, 2, 3

14 8.

LOSS OF POWER a.

4 kv Bus Loss of Voltage b.

4 kv Bus Degraded Voltage 3/Bus 3/Bus 2/Bus 2/Bus 2/Bus 2/Bus 1, 2, 3, 4

1, 2, 3, 4 14*

0

~

~

4 n

TABLE 3.3-4 ENGIHEEREO SAFETY FEATURE ACTUATIO SYSTEM INSTRUMENTATION TRIP SETPOINTS n

CD CD I

FUNCTIONAL UNIT 1.

SAFETY ItWECTION, FEEDMATER ISOLATION AND MOTOR DRiVEN AUXILIARY FEEDWATER PUMPS a.

Manual Initiation TRIP SETPOINT Not Applicable ALLOWABLE VALUE Not Applicable b.

Automatic Actuation Logic c.

Containment Pressure lligh d.

Pressuri zer Pressure--Low e.

Differenti a'l Pressure Between Steam Lineslligh f.

Steam Flow in Two Steam Lines-lligh Coinc~dent with Tav

-Low-Low

.. or Steam Line Pressure-3ow Not Applicable

\\

< 1.1 psig

> 1815 psig

< 100 psi Not Applicable

< 1.2 psig

~'

1805 psig

~ a,

'.i

< 112 psi

< 1.42 x 10 lbs/hr from OX load to 20$

load.

Li~ear from i

1.42 x 10 lbs/hr at6205 load to 3.88 x.

10 lbs/hr at lOOX load T

> 541'F

> MO psig steam line pressure a

- ~

< 1.56 x 10 lbs/hr from OX load to 20K load.

Li~ear from

).56 x 10 lbs/hr at620X load to 3.93 x 10 lba/hr at 100C load.

T

> 539'F

> 3IIO psig steam line pressure

~ I a/

TABLE 3.3-4 Cont'd.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS n'D CD FUNCTIONAL UNIT 6.

MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.

Steam Generator Water Level-Low-Low

b. '4 kv Bus Loss of Voltage c.

Safety Injection d.

Loss of Main Feedwater Pumps TRIP SETPOINT

> 105 of narrow range Tnstrument span each steam generator 3196 volts with a 2-second delay Not Applicable Not Applicable ALLOWABLE VALUES

> 9C of harrow range Tnstrument span each steam generator 3196 + 18 volts with a

2 a.2 second delay Not Applicable Not Applicable 7.

TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS a.

Steam Generator Water Level-Low-Low b.

Reactor Coolant Pump Bus Undervoltage

> lOX of narrow range instrument span each

'team generator

> 2750 Volts-each bus

> 9X of narrow range Tnstrument span each steam generator

~ >2725 Volts-each bus 8.

LOSS OF POWER a..

4 kv Bus Loss of Voltage b.

4 kv Bus Degraded Voltage 3196 volts with a 2-second delay 3596 volts with a 2.0 min. time delay 3196 a 18 volts with a

2 a.2 second delay 3596 a 18 volts with a 2.0 minute a 6 second time delay

TABLE 3.3-5 Continued ENGINFERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION 6.

Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low

RESPONSE

TIME IN SECONDS a 0 b.

Ce d.

e.

9 ~

h.

Safety Injection (ECCS)

Reactor Trip {from SI)

Fe'edwater Isolation Con ta inmen t Iso 1 a tion-Pha se "A"

Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System

.Steam Line Isolation

< 13.0Pr/23. Pg"

< 3.0

< 8.0

< 18.0f/28.0ÃP Not Applicable Hot Applicable

< 14.0$ /48.04' 8.0 7.

Containment Pressure Hi h-Hi h

a.

Containment Spray b.

Containment Isolation-phase "8"

c.

Steam Line Isolation d.

Containment Air Recirculation Fan

< 45.0 Not Applicable

< 7.0

< 660.0 8.

Steam Generato~

Water Level--Hi h-Hi h

a.

Turbine Trip-Reactor Trip b.'eedwater Isolation

~ r 9.

Steam Generator Water LevelLow-Low

< 2.5

< 11.0 a.

- Motor Driven Auxiliary Feedwater Pumps

< 60.0 b.

Turbine Driven Auxiliary Feedwater Pumps

< &0.0 10.

4160 volt Emergency Bus Loss of Voltage a.

Motor Driven Auxiliary Feedwater Pumps

< 60.0 11.

Loss Of Main Feedwater Pum s

a.

Motor Driven Auxiliary Feedwater Pumps c 60.0 12.

Reactor Coolant Pum Bus Undervolta'e a.

Turbine Driven Auxi'liary Feedwater Pumps

< 60.0 D.C.

COOK - UNIT 1 3/4 3-29

TABLE 4.3-2

'L ENGINEERED SAFETY FEATURE ACTUATION SYSTEtt INSTRUttENTATION UR E

L NCE RE UIRENE lT CD C)

I FUNCTIONAL UNIT l

1.

SAFETY INECTION,FEEDlQTER ISOLATION AND MOTOR DRIVEN I AUXIL'IARY FEEDWATER PUt1PS a.

Hanual Initiation b.

Automatic Actuation Logic c.

Containment Pressure-tligh d.

Pressurizer Pressure--Low e.

Differential Pressure Between Steam Lines--High f.

Steam Flow in Two Steam LinesHigh Coincident with T

Low or Steam Line PQksur eLow 2.

CONTAINt1ENT SPRAY CHANNEL CHECK tt.A.

N.A.

CHANNEL CALIORATION N.A.

tl.A.

CHANNEL FUNCTIONAL TEST

~ t1 {l )

tl(2)

N(3)

HODES IN HHICH SURVEILLANCE RE ltIRED 1,2,3,4 1,2,3,4 1,2.

3 1,.2, 3

1,2,3

'I 1,2,3 a.

manual Initiation b.

Automatic Actuation.Logic c.

Containment Pressure High-High N.A..

N,A.

S N.A M{1) tl.A.'l{2)

- 1, 2, 3, 4 1, 2, 3, 4 1, 2, 3

TABLE 4. 3 Continued EHGIHEEREO SAFETY FEATURE ACTUATIOH SYSTEM INSTRUMENTATION R EIELA~E~fflLMNT n

C)

I C

I FUNCTIONAI UNIT STEAN LINE ISOLATION a.

Manual b.

C ~

Automatic Actuation Logic.

Containment Pressure High-High CHANNEL CHECK H.A.

H.A.

CHANNEL CALIBRATION N.A.

H.A.

CHANNEL FUHCTIOHAL TEST H(2)

H(3)

NODES IN WHICH-.-

SURVEILLANCE

'E UIRED 1,2;3 1,2,3 1,2,3

. ~

d.

Steam Floe in Two Steam;-

S LinesHigh Coincident with T

-- Low or Steam Line PAksure Low 5.'URBINE TRIP ANO FEEOWATER ISOLATION a.

Steam Generator Water Level--High-High 6.

MOTOR DRIVEN AUXILIARY FEEOWATER PUMPS a.

Steam Generator Mater S

Level--Low-Low,

b.

4 kv Bus Loss of Voltage

~

y 1,2,3, 1,.2, 3

1, 2; 3

1,2,3 g

g ~

c.

Safety Injection,~

d.

Loss of Hain Feed Pumps Aa N.A.

N.A.

N.A.

H(2)

N.A.

1, 2, 3

1, 2, 3

TABLE 4.3-2 Continued I

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 7.

TURBINE DRIVFN AUXILIARY FEEDWATER PUMPS CHANNEL CHECK CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST MODES IN WHICH SURVEILLANCE

~RE IIIRER a ~

b.

Steam Generator Water Level--Low-Low Reactor Coolant Pump Bus Undervoltage N.A.

R M

1,2,3 1,2,3 8.

LOSS OF POWER a.

4 kv Bus Loss of Voltage b.

4 kv Bus Degraded Voltage 1,2,3,4 1, 2, 3, 4

I I

INSTRUMENTATION POST-ACCIDENT INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3. 8 The post-accident monitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.

APPLICABILITY:

MODES l, 2 and 3.

ACTION:

a.

With the number of OPERABLE post-accident monitoring channels less than required by Table 3.3-11, either restore the inoperable channel to OPERABLE status within 30 days, or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

b.

The provisions of Specification 3.0.4 are not applicable.

SURVEIlLANCE RE UIREMENTS 4.3.3. 8 Each post-accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.

D. C.'OOK - UNIT 1

TABLE 3.3-11 POST-ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE-1.

Containment Pressure 2.

Reactor Coolant Outlet Temperature - THOT (Wide Range) 3.

Reactor Coolant Inlet Temperature -

TCOLD (Wide Range) 4.

Reactor Coolant Pressure

- Wide Range 5.

Pressurizer Water Level 6.

Steam Line Pressure 7.

Steam Generator Water Level - Narrow Range 8.

Refueling Water Storage Tank Water Level 9.

Boric Acid Tank Solution Level 10.

Auxiliary Feedwater Flow Rate 11..

Reactor Coolant System Subcooling Margin Monit6r 12.

PORV Position Indicator - Limit Switches***

13.

PORV Block Valve Position Indicator - Limit Switches 14.

Safety Valve Position Indicator - Acoustic Monitor 2/Steam Generator 1/Steam Generator 2

1/Steam Generator" 1/Valve 1/Valve 1/Valve

• Steam Generator Water Level Channels can be used as a substitute for the corresponding auxiliary feedwater flow rate channel instrument.

• • PRODAC 250 subcooling margin readout can be used as a substitute for the subcooling monitor instrument.

• • • Acoustic monitoring of PORV position (1 channel per three valves - headered discharge) can be used as a substitute for the PORV Position Indicator-L>mitSwitches instruments.

~ A A

C)

~ C)

~ 7C INSTRUMENT TABLE 4.3-7 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHECK CHANNEL CALIBRATION Ca)I CJl 1.

Containment Pressure 2.

Reactor Coolant Outlet Temperature - THOT (Wide Range) 3.

Reactor Coolant Inlet Temperature -

TCOLD (Wide Range) 4.

Reactor Coolant Pressure

- Wide Range 5.

Pressurizer Water Level 6.

Steam Line Pressure 7.

Steam Generator Water Level - Narrow Range 8.

RWST Watei Level 9.

Boric Acid Tank Solution Level 10.

Auxiliary Feedwater Flow Rate ll.

Reactor Coolant System Subcooling Margin Monitor 12.

PORV Position Indicator - Limit Switches 13.

PORV Block Valve Position Indicator - Limit Switches 14.

Safety Valve Position Indicator - Acoustic Monitor M

M R

~

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume less than or equal to 624 of span and at least 150 kW of pressurizer heaters.

APPLICABILITY:

MODES 1 and 2

ACTION' With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supplywithin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.4.4.1 Not applicable.

4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency poser supply and energizing the required capacity of heaters.

D. C.

COOK - UNIT 1 3/4 4-6

REACTOR COOLANT SYSTEM ~

RELIEF VALVES - OPERATING LIMITING CONDITION FOR OPERATION 3.4.11 Three Power Operated Relief Valves (PORVs) and their associated

,block valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2 and 3

ACTION:

a ~

b.

C.

d.

With one PORV inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV to OPERABLE status or close the associated block valve and remove power from the'block valve; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The provisions of Specifications 6.9.1:9, 3.0.3 and 3.0.4 are not applicable.

With two or more PORVs inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORVs to OPERABLE status or close the associated block valves and remove power from the block valves; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

With one block valve inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either (1) restore the block valve to OPERABLE status or (2) close the block valve and remove power from the block valve or (3) close the associated PORV and remove power from its associated Solenoid valve;otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

The provisions of Specifications 6.9.1.9, 3.0.3 and 3.0.4 are not applicable.

With two or more block valves inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either (1) restore the block valves to OPERABLE status or (2) close the block valves and remove power from the block valves or (3) close the associated PORVs and remove power from their associated Solenoid valves; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE RE UIREMENTS 4.4. 11.1 Each of the three PORVs shall be 'demonstrated OPERABLE:

a.

At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and b.

At least once per 18 months by performance of a CHANNEL CALIBRATION.

D. C.

COOK UNIT l 3/4 4-41

SURVEILLANCE RE UIREMENTS Cont'd

'4.4.11.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.

The provisions of Specification 4.0.4 are not applicable when Actions 3.4.ll.a or 3.4.ll.c are applied.

4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through' complete cycle of full travel while the emergency buses are energized by the on-site diesel generators and on-site plant batteries.

This

'testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.b and 4.8.2.3.2.c.

D. C.

COOK UNIT 3 3/4 4-42

3 4.0 APPLICABILITY SURVEILLANCE REQUIRB"ENTS Continued b.

A total maximum ccmbined interval time for any 3 consecutive survei 1 1 ance interva1 s not to exceed

3. 25 times the speci fi ed surveillance interval.

4.0.3 Perfor,ance of a Surveillance Requirer ent within the sp cified time interval shall constitute compliance with OPERABILITY requirements for a Limiting Condition for Operation and associated ACTION statements unless otherwise required by the specification.Surveil'lance Requil.ements do not have to be ~elformed on inoperable equipment.

4.0.4 Entry in.o an OPERATIONAL MQOE or other specified applicability condition shall not be rade unless ihe Surveillance Re.!ire-:.,en.(s) associated with the Limiting Condition for Operation have been perfor.-..ed within the stated surveillance interval or as otherwise specified.

The provisions of Speci ication 4.0.4 are not applicable to the per-formance of surveillance a tivities associated with fire protection technical speci-ications, 4.3.3.7, 4.7.9 and 4.7.1Q, until the completion of the initial surveillance interval associated with each specification.

0. C.

COOK - Uib?T 1

3(4 0-2

i

~

~

~

VALVE NUMBER

. A.

PHASE "A" ISOLATION FUNCTION Continued TABLE 3.6-1 Continued I

TESTABLE DURING ISOLATION TIHE:

PLANT OPERATION'H SECONDS

(

l 57.

58.

59.

60.

61.

62.

63.

64.

65.

66.

67.

68.

69.

70.

71.

72.

73.

74, 75.

l.

2.

3.

5.

6.

7.

t/CR-107 tlCR-108 tlCR-109 tlCR-110 HCR-252 QCN-250 QCH-350 OCR-300 QCR-301 RCR-100 RCR-101 VCR-10 VCR-11 VCR-20 VCR-21 XCR-100 XCR-101 XCR-102 XCR-103 PllASE "B" ISOLATION CCH-451 CCt1-452 CCH-453 CCfl-454 CCH-458 CCH-459 ECR-31 ECR-32 PRZ Liquid 'Sample PRZ Liquid Sample PRZ Steam Sample PRZ Steam Sample Primary Water to Pressurizer Relief Tank'CP Seal Hater Discharge RCP Seal Water Discharge LeMo<<n to Letdown Nx.

Letdown to Letdown Hx..

PRZ Relief Tank to Gas Anal.

PRZ Relief Tank.to Gas Anal.

Glycol Supply to Fah Cooler Glycol Supply to Fan Cooler Cilycol Supply from Fan.Cooler Glycol Supply from Fan Cooler.

Control Air to Containment Control Air to Containment Isolation Control Air to Containment Isolation Control Air to Containment CCW from PCP Oil Coolers CCW from RCP Oil Coolers CCW from RCP Thermal Barrier CCW from RCP Tliermal Barrier CCH to RCP Oil Coolers 8 Thermal Barrier CCH to RCP Oil Coolers L Thermal Barrier.

Containment Air Particle Radio Gas Detector Containment Air Particle Radio Gas Detector Yes Yes Yes Yes Yes Ho Ho Ho Ho Yes Yes Yes Yes Yes Yes tlo tlo

'o

-Ho Iw Ho Ho Ho tlo Ho

~ Ho

'o Ho 10 10 10 10 10 15 15 10 10 10 10 10 10 10 10 10 10 10 10 60 60 30 30 60 60 10 10 I

I I

i ~

Ij 4p (x

)

I i

~

1 l

'i I

ljI

TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITIONS LICENSE CATEGORY

'OL OL NON-Licensed Shift Technical Advisor 1, 2, 3 & 4 5&6 None Re uired APPLICABLE MODES Woes,not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS after the initial fuel loading.

fShift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accoraradate

'nexpected absence of on duty shift crew members provided ioxradiate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

",*Shared With D.C.

COOK - UNIT 2.

O.,C.

COOK - UNIT 1 6-4 Amendment No.

~ g)

AOMINISTRATIVE CONTROLS

6. 3 FACILITY STAFF UAL IF ICATIONS

~-3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI NQ.1-1971 for comparable positiorrs, except fur (1) the Radiation Protection Supervisor.who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975 and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and analysis of the plant for transients and accidents.~

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the r equirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55, 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the require-ments of Section 27 of the NFPA Code-1976.

6.5 REVID ANO AUOIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION 6.5.1.1 The PNSM shall function to advise the Plant Manager on all

'atters related to nuclear safety.

Ful compliance by January 1,1981 O. C.

COOK - UNIT 1 6-5 Amendment No.

INSTR VMiENTATION BASES 3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation. ensures thai adequate warning capability is available for the prcmpt detection of fires.

This capability is required in order to detect and locate fires in their e rlv stages.

Prompt detection of fires will reduce th. poten-tia'1 for da.age to safety related equipment and is an integral element in the overall,acility fire protection program.

In the event that a portion of the fire detection instru entation is inoperable, ti e establisn.-.,ent of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY.

3 4.3.3.8 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information. is available on selected plant parameters to monitor and assess'hese variables during and following an accident.

D.

C.

COOK-UNIT 1 B 3/4 3-4

REACTOR COOLANT SYSTEM BASES 3/4.4.4 PRESSURIZER' steam bubble in the pressurizer ensures that the RCS is not a hydraulically solid system and is capable of accommodating pressure surges during operation.

The steam bubble also protects the pressurizer code safety valves and power operated relief valves against water relief.

The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.

Operation of the power operated

'elief valves minimizes the undesirable opening of the spring-loaded pressurizer code safety valves.

The requirement that 150 Klrl of pressurizer heater s and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of offsite power condition to maintain natural circulation conditions.

3 4.4.5 STEAM GENERATORS e

urves ance equirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained.

The program for inservice inspection of steam generator tubes is based on a modification of Regulatory Guide 1;83, Revision l.

Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to

design, manufacturing errors, or in'service conditions that lead to corrosion.

Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those parameter limits found to result in negligible corrosion of the steam generator tubes.

If the secondary coolant chemistry is not maintained within these parameter limits, localized corrosion may likely result in stress corrosion cracking.

The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage

=

500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an 0.

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REACTOR COOLANT SYSTEM BASES the ASIDE Boiler and Pressure Vessel Code "Inservice Inspection of Nuclear Reactor Coolant Systems",

1971 Edition and Addenda through Minter 1972.

All areas scheduled for volumetric examination have been pre-service mapped using equipment, techniques and procedures anticipated for use during post-operation examinations.

To assure that consideration is given to the use of new or improved inspection equipment, techniques and procedures, the Inservice Inspection Program will be periodically reviewed on a 5 year basis.

The use of conventional nondestructive, direct visual and remote visual test techniques can be applied to the inspection of most reactor coolant loop components except the reactor vessel.

The reactor vessel requires special consideration because of the radiation levels and the requirement for remote underwater accessibility.

The techniques anticipated for inservice inspection include visual inspections, ultrasonic, radiographic, magnetic particle and dye penetrant testing of selected parts.

The nondestructive testing f'r repairs on components greater than 2 inches diameter gives a high degree of confidence in the integrity of the system, and will detect any significant defects in and near the new welds.

Repairs on components 2 inches in diameter or smaller receive a

surface examination which assures a similar standard of integrity.

In each case, the leak test will ensure leak tightness during normal oper ation.

For normal opening and reclosing, the structural integrity of the Reactor Coolant System is unchanged.

Therefore, satisfactory performance of a system leak test at 2235 psig following each opening and subsequent reclosing is acceptable demonstration of the system's structural integrity.

These leak tests will be conducted within the pressure-temperature limita-tions for Inservice Leak and Hydrostatic Testing and Figure 3.4-1.

3 4.4.11 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure be'low the setting of the pressurizer code safety valves.

These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable.

The electrica'l power for both the relief valves and the block valves is supplied from an emergency power source to ensure the ability to seal this possible-RCS leakage path.

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