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

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.

b. 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

TABLE 3.3-3 ~ ~

n ENGINEERED SAFETY FEATURE ACTUATION SYSTEH INSTRUHENTATION CD CD 7'

HINIMUH TOTAL NO. CHANNELS CHANNELS FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE HODES ACTION SAFETY INJECTION,FEEDWATER ISOLATION AND MOTOR DRIVEN AUXILIARY FEEDWATER PUHPS Initiation .,1,2,3,4 a.

b.

Hanual Automatic Actuation Logic 2

2 2

2.

line'PPLICABLE 1,2N 3N4 I

'3 1&

I

c. Containment'ressure-High 3 n 4e 2 1,2,3 )* 14

.d. Pressurizer 2, 2 1,2',34- i 14 Pressure - Low

e. Differential 1, 2, 3N Pressure Between Steam Lines - High Four Loops 3/steam line 2/steam line 2/steam line Operatinq any steam line Three Loops 3/operating, 1 /steam 2/operating 15 Operating steam line line, any steam line \

operating steam

TABLE 3.3-3 Cont'd.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water Level- 2/Stm. Gen.

Low-Low 3/Stm. Gen. Any Stm. Gen. 2/Stm. Gen. 1, 2, 3 14*

b. 4 kv Bus Loss of Voltage 2/Bus 2/Bus 2/Bus 1, 2, 3 le~

4'k

~

c. Safety Injection 1, 2, 3 1

d. Loss of Main Feedwater Pumps 2 1, 2, 3

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS.

a. Steam Generator Water Leyel- 2/Stm. Gen. 2/Stm. Gen.

Low-l ow 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 3/Bus 2/Bus 2/Bus 1, 2, 3, 4

b. 4 kv Bus Degraded Voltage 3/Bus 2/Bus 2/Bus 1, 2, 3, 4 14*

0 ~ ~

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

CD CD I FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUE

1. SAFETY ItWECT ION, FEEDMATER ISOLATION AND MOTOR DRiVEN AUXILIARY FEEDWATER PUMPS

a. Manual Initiation Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable

\

c. Containment Pressure lligh ,

< 1.1 psig .. < 1.2 psig

d. Pressuri zer Pressure--Low > 1815 psig

~

' 1805 psig

~ a,

e. Di fferenti a'l Pressure < 100 psi '.i < 112 psi Between Steam Lines lligh a

f. Steam Flow in Two Steam Lines- < 1.42 x 10 lbs/hr - ~ < 1.56 x 10 lbs/hr lligh Coinc~dent with Tav -Low-Low from OX load to 20$ from OX load to 20K

.. or Steam Line Pressure-3ow load. Li~ear from i load. Li~ear from 1.42 x 10 lbs/hr ).56 x 10 lbs/hr at6205 load to 3.88 x. at620X load to 3.93 x '

10 lbs/hr at lOOX load 10 lba/hr at 100C load.

T > 541'F T > 539'F

> MO psig steam line > 3IIO psig steam line pressure pressure

~ I a/

TABLE 3.3-4 Cont'd.

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES n'D

6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS CD

a. Steam Generator Water Level- > 105 of narrow range > 9C of harrow range Low-Low Tnstrument span each Tnstrument span each steam generator steam generator

b. '4 kv Bus 3196 volts with a 3196 + 18 volts with Loss of Voltage 2-second delay a 2 a.2 second delay

c. Safety Injection Not Applicable Not Applicable

d. Loss of Main Feedwater Pumps Not Applicable Not Applicable

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator Water Level- > lOX of narrow range > 9X of narrow range Low-Low instrument span each Tnstrument span each generator 'team steam generator

b. Reactor Coolant Pump Bus Undervoltage > 2750 Volts-each bus ~ >2725 Volts-each bus

8. LOSS OF POWER a.. 4 kv Bus 3196 volts with a 3196 a 18 volts with Loss of Voltage 2-second delay a 2 a .2 second delay

b. 4 kv Bus Degraded Voltage 3596 volts with a 3596 a 18 volts with 2.0 min. time delay a 2.0 minute a 6 second time delay

TABLE 3.3-5 Continued ENGINFERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

6. Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low a 0 Safety Injection (ECCS) < 13.0Pr/23. Pg"

b. Reactor Trip {from SI) < 3.0 Ce Fe'edwater Isolation < 8.0

d. Con ta inmen t Iso 1 a ti on-Pha se "A" < 18.0f/28.0ÃP

e. Containment Purge and Exhaust Isolation Not Applicable Auxiliary Feedwater Pumps Hot Applicable 9 ~ Essential Service Water System < 14.0$ /48.04'

h. .Steam Line Isolation 8.0

7. Containment Pressure Hi h-Hi h

a. Containment Spray < 45.0

b. Containment Isolation-phase "8" Not Applicable

c. Steam Line Isolation < 7.0

d. Containment Air Recirculation Fan < 660.0

8. Steam Generato~ Water Level--Hi h-Hi h

a. Turbine Trip-Reactor Trip < 2.5 b.'eedwater Isolation < 11.0

~ r

9. Steam Generator Water Level Low-Low

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 CHANNEL HODES IN HHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIORATION TEST RE ltIRED l

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

a. Hanual Initiation tt.A. N.A. ~

t1 { l) 1,2,3,4

b. Automatic Actuation Logic N.A. tl.A. tl(2) 1,2,3,4

c. Containment Pressure-tligh N(3) 1,2. 3

d. Pressurizer Pressure--Low 1,.2, 3

e. Di fferential Pressure 1,2,3 Between Steam Lines--High

'I

f. Steam Flow in Two Steam 1,2,3 Lines High Coincident with T Low or Steam Line PQksur e Low

2. CONTAINt1ENT SPRAY

a. manual Initiation N.A.. N.A M{1) 1, 2, 3, 4

b. Automatic Actuation .Logic N,A. tl.A.'l{2) 1, 2, 3, 4

c. Containment Pressure High- S 1, 2, 3 High

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

I CHANNEL NODES IN WHICH-.-

I C CHANNEL CHANNEL FUHCTIOHAL SURVEILLANCE

'E F UNCT IONAI UNIT CHECK CALIBRATION TEST UIRED STEAN LINE ISOLATION

a. Manual H.A. N.A. 1,2;3

b. Automatic Actuation Logic. H.A. H.A. H(2) 1,2,3 . ~

C ~ Containment Pressure H(3) 1,2,3 High-High

d. Steam Floe in Two Steam;- S 1,2,3, Lines High Coincident with T -- Low or Steam Line PAksure Low 5.'URBINE TRIP ANO FEEOWATER ISOLATION

a. Steam Generator Water 1,.2, 3 Level--High-High

6. MOTOR DRIVEN AUXILIARY FEEOWATER PUMPS g

g ~

a. Steam Generator Mater S ~

y 1, 2; 3 Level--Low-Low ,

b. 4 kv Bus Loss of Voltage 1,2,3

c. Safety Injection,~ N.A. N.A. . H(2) 1, 2, 3

d. Loss of Hain Feed Pumps N.A. N.A. 1, 2, 3 Aa

TABLE 4.3-2 Continued I

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST ~RE IIIRER

7. TURBINE DRIVFN AUXILIARY FEEDWATER PUMPS a ~ Steam Generator Water Level--Low-Low 1,2,3

b. Reactor Coolant Pump Bus Undervoltage N.A. R M 1,2,3

8. LOSS OF POWER

a. 4 kv Bus Loss of Voltage 1,2,3,4

b. 4 kv Bus Degraded Voltage 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 2/Steam Generator

7. Steam Generator Water Level - Narrow Range 1/Steam Generator

8. Refueling Water Storage Tank Water Level 2

9. Boric Acid Tank Solution Level

10. Auxiliary Feedwater Flow Rate 1/Steam Generator" 11.. Reactor Coolant System Subcooling Margin Monit6r

12. PORV Position Indicator - Limit Switches*** 1/Valve

13. PORV Block Valve Position Indicator - Limit Switches 1/Valve

14. Safety Valve Position Indicator - Acoustic Monitor 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>mit Switches instruments.

TABLE 4.3-7 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS

~ A A

C)

CHANNEL CHANNEL

~ C) INSTRUMENT CHECK CALIBRATION

~ 7C

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 Ca)

I CJl 8. RWST Watei Level M

9. Boric Acid Tank Solution Level

10. Auxiliary Feedwater Flow Rate ll. Reactor Coolant System Subcooling Margin Monitor M

12. PORV Position Indicator - Limit Switches

13. PORV Block Valve Position Indicator - Limit Switches R ~

14. Safety Valve Position Indicator - Acoustic Monitor

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 ~ 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.

b. With two or more PORVs inoperable, within 1 hour 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 />.

C. 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.

d. 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 in terva1 s f not to exceed 3. 25 times the speci i 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 ~

~ ~

TABLE 3.6-1 Continued

'H I

TESTABLE DURING ISOLATION TIHE:

VALVE NUMBER FUNCTION PLANT OPERATION SECONDS

(

. A. PHASE "A" ISOLATION Continued l

57. t/CR-107 PRZ Liquid 'Sample Yes 10

58. tlCR-108 PRZ Liquid Sample Yes 10

59. tlCR-109 PRZ Steam Sample Yes 10

'i

60. tlCR-110 PRZ Steam Sample Yes 10 I

61. HCR-252 Primary Water to Pressurizer Relief Tank'CP Yes 10 I

62. QCN-250 Seal Hater Discharge Ho 15 I I

63. QCH-350 RCP Seal Water Discharge Ho 15 i ~

64. OCR-300 LeMo<<n to Letdown Nx. Ho 10

65. QCR-301 Letdown to Letdown Hx.. Ho 10

66. RCR-100 PRZ Relief Tank to Gas Anal. Yes 10 I j

67. RCR-101 PRZ Relief Tank.to Gas Anal. Yes 10

68. VCR-10 Glycol Supply to Fah Cooler Yes 10

69. VCR-11 Glycol Supply to Fan Cooler Yes 10

70. VCR-20 Cilycol Supply from Fan .Cooler Yes . 10 4p

71. VCR-21 Glycol Supply from Fan Cooler. Yes 10

72. XCR-100 Control Air to Containment tlo 10

73. XCR-101 Control Air to Containment Isolation tlo 10 (x Air to Isolation

'o 74, XCR-102 Control Containment 10 )

75. XCR-103 Control Air to Containment -Ho 10 I Iw PllASE "B" ISOLATION

l. CCH-451 CCW from PCP Oil Coolers Ho 60

2. CCt1-452 CCW from RCP Oil Coolers Ho 60

3. CCH-453 CCW from RCP Thermal Barrier Ho 30 CCfl-454 CCW from RCP Tliermal Barrier tlo 30

5. CCH-458 CCH to RCP Oil Coolers 8 Thermal Barrier Ho 60

6. CCH-459 CCH to RCP Oil Coolers L Thermal Barrier. ~

Ho 60 i ~

Air Particle

'o

7. ECR-31 Containment Radio Gas Detector 10 ECR-32 Containment Air Particle Radio Gas Detector Ho 10 1

l

lj I TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITIONS LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6

'OL OL NON-Licensed Shift Technical Advisor None Re uired 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)

AOMINI STRATI V E 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 related to nuclear safety. 'atters Ful compliance by January 1,1981

• O. C. COOK - UNIT 1 6-5 Amendment No.

INSTR VMiENTATI ON 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 valves minimizes the undesirable opening of the spring-loaded 'elief 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. C. COOK-UNIT 1 B 3/4 4-2

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.

0. C. COOK-UNIT 1 B 3/4 4-12