ML17331A287

Proposed TS 4.5.1.d Re SR for Accumulator Isolation Valves, 4.5.2.e Re SR for ECCS Automatic Valves,Centrifugal Charging Pump,Se Pump & RHR pump,4.6.2.1.c Re SR for Containment Spray Automatic Valves & 4.7.6.1.d.3 Re ESF Ventilation

ML17331A287
Person / Time
Site:	Cook
Issue date:	04/16/1993
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17331A288	List: ML17331A287 ML17334B471
References
AEP:NRC:1181, NUDOCS 9304220177
Download: ML17331A287 (71)

Text

Attachment 2 to AEP:NRC:1181 Proposed Technical Specifications for Donald C. Cook Nuclear Plant Unit 2 9304220177 9304f6 PDR ADQCK 05000316 P '

'PDR

3 4.0 APPLICABILITY SURVEILLANCE RE UIREMENTS 4.0.8 By specific reference to this section, those surveillances which must be performed on or before August 13, 1994, and are designated as 18-month or 36-month surveillances (or required as outage-related surveillances under the provisions of Specification 4.0.5) may be delayed until the end of the cycle 9-10 refueling outage.

For these specific surveillances under this section, the specified time intervals required by Specification 4.0.2 will be determined with the new initiation date established by the surveillance date during the Unit 2 1994 refueling outage.

COOK NUCLEAR PLANT - UNIT 2 3/4 0-4 AMENDMENT NO.

REACTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS-SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.3.3 At least one rod position indicator channel (excluding demand position indication) shall be OPERABLE for each shutdown or control rod not fully inserted.

APPLICABILITY: MODES 3*8, 4+/ and 5*jj ACTION:

With less than the above required position indicator channel(s) OPERABLE, immediately open the reactor trip system breakers.

SURVEILLANCE RE UIREMENTS 4,1.3.3 Each of the above required rod position indicator channel(s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.f

+With the reactor trip system breakers in the closed position.

jjSee Special Test Exception 3.10.5.

i The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 1-22 AMENDMENT NO. 40, QQ,

POWER DISTRIBUTION LIMITS DNB AND Tav OPERATING PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the following operational indicated limits:

a. DNB

1. Reactor Coolant System T Less than or equal to 578.7'W

2. Pressurizer Pressure g Greater than or equal to 2200 psig+/**

3. Reactor Coolant System Greater than or equal to 366,400 gpm~*

Total Flow Rate

b. T avg

1. Reactor Coolant System T Greater than or equal to 543.9'W avg APPLICABILITY: MODE 1 ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5X of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.5.1 Each of the above parameters shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The indicators used to determine RCS total flow shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.$

4.2.5.3 The RCS total flow rate shall be determined by a power balance around the steam generators at least once per 18 months.

4.2.5.4 The provisions of Specification 4.0.4 shall not apply to primary flow surveillances.

• Indicated average of at least three OPERABLE instrument loops.

• • Limit not applicable during either a THERMAL POWER ramp in excess of 5X of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10X of RTP

• • • Indicated value f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 2-15 AMENDMENT NO.

3 4.3 INSTRUMENTATION 3 4 3 1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3 '-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION's shown in Table 3.3-1.

SURVEILLANCE RE UIREMENTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1.f 4.3.1.1.2 The logic for the interlocks shall be demonstrated OPERABLE prior to each reactor startup unless performed during the preceding 92 days. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.i 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.$

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-1 AMENDMENT NO.

TABLE 4.3-1 O REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS O

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED

1. Manual Reactor Trip A. Shunt Trip Function N.A. N.A. S/U(1) (10) 3*, 4*, 5*

B. Undervoltage Trip N.A. N.A. S/U(1) (10) ], 2, 3*, 4*, 5*

I Function

2. Power Range, Neutron Flux D(2,8),M(3,8) M and S/U(1) 1, 2 and

• and Q(6,8)

3. Power Range, Neutron Flux, N.A. R(6) 1 2 High Positive Rate

4. Power Range, Neutron Flux, N.A. R(6) 1 2 High Negative Rate

5. Intermediate Range, R(6,8)f S/U(1) 1, 2 and*

Neutron Flux

6. Source Range, Neutron Flux R(6,14) M(14) and S/U(l) 2(7) 3(7) 4 and 5 W

7. Overtemperature hT R(9)f 1 2 2'I 2 8. Overpower dT R(9) f '1 2 R

9. Pressurizer Pressure--Low Rf 1 2 O

10. Pressurizer, Pressure--High R$ 1 2 ll. Pressurizer Water Level--High Rf 1 2

12. Loss of Flow<- Single Loop R(8) f f The provisions of Technical Specification 4.0.8 are applicable.

TABLE 4 3-1 Continued A

O REACTOR TRIP SYSTEM INSTRUMENTATION-SURVEILLANCERE UIREMENTS O

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST RE UIRED

13. Loss of Flow-Two Loops R(8)f N.A.

14. Steam Generator Water Level-- 1 2 I Low-Low

15. Steam/Feedwater Flow Mismatch and 1 2 Low Steam Generator Water Level

16. Undervoltage-Reactor Coolant N.A.

Pumps

17. Underfrequency-Reactor Coolant N.A.

Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1 2 B. Turbine Stop Valve Closure N.A. N.A. S/U(1) 1 2

19. Safety Injection Input from EFS N.A. N.A. M(4) 1 2

20. Reactor Coolant Pump Breaker N.A. N.A. N.A.

Position Trip

21. Reactor Trip Breaker A. Shunt Trip Function N.A. N.A. M(5) (11) and S/U(1) (11) 1,2,3*,4*,5*

B. Undexvoltage Trip Function N.A. N.A. M(5) (11) and S/U(l) (ll) 1,2,3*,4*,5*

22. Automatic Trip Logic N.A. N.A. M(5) ],2,3*,4*,5*

23. Reactor Trip Bypass Breaker N.A. N.A. M(12) and S/U(1) (13) ],2,3*,4*,5*

f The provisions of Technical Specification 4.0.8 are applicable.

INSTRUMENTATION 3 4 3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumenta-tion channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

APPLICABILITY: As shown in Table 3.3-3.

ACTION With an ESFAS instrumentation channel trip setpoint less conserva-tive than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3, SURVEILLANCE RE UIREMENTS 4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST and TRIP ACTUATING DEVICE OPERATIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2.f 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test. The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.f 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" Column of Table 3.3-3 t The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-14 AMENDMENT NO.

TABLE 4.3-2 ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST

1. SAFETY INJECTION; TURBINE TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Manual Initiation See Functional Unit 9

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

c. Containment Press- M(3) N.A. 1 2 3 ure-High

d. Pressurizer Press- Rf N.A. 1 2 3 ure-Low

e. Differential Press- N.A. 1 2 3 ure Between Steam Lines--High

f. Steam Line Pressure-- N.A. 1 2 3 Low CONTAINMENT SPRAY

a. Manual Initiation See Functional, Unit 9

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

c. Containment Press- M(3) N.A. 1, 2, 3 ure-High-High

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual See Functional Unit 9

2) From Safety N.A. N.A. M(2) N.A. 1, 2, 3, 4 Injection Automatic Actuation Logic

b. Phase "B" Isolation

1) Manual See Functional Unit 9

2) Automatic Actua- N.A. N.A. M(2) N.A. 1, 2, 3, 4 tion Logic

3) Containment Press- M(3) N.A. 1, 2, 3 ure-High-High The provisions of Technical Specification 4.0.8 are applicable.

I COOK NUCLEAR PLANT - UNIT 2 3/4 3-30 AMENDMENT NO. 34,

TABLE 4 3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST ml"

c. Purge and Exhaust Isolation

1) Manual See Functional Unit 9

2) Containment Radio- S M N.A. 1, 2,,3,4 activity-High

4. STEAM LINE ISOLATION

a. Manual - See Functional Unit 9

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

c. Containment Press- M(3) N.A. 1, 2, 3 ure--High-High

d. Steam Flow in N.A. 1 2 3 Two Steam Lines--

High Coincident with T --Low-Low

e. Steam Line Pressure-- N.A. 1 2 3 Low

5. TURBINE TRIP AND FEEDWATER ISOLATION

a. Steam Generator 1 2 3 Water Level--High-High "6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator N.A. 1 2 3 Water Level--Low-Low

b. 4 kV Bus N.A. 1, 2, 3 Loss of Voltage

c. Safety Inj ection N.A. N.A. M(2) N.A. 1, 2, 3

d. Loss of Main Feed N.A. N.A. Rf N.A. 1, 2 Pumps f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-31 AMENDMENT NO.

434, ~

8R, &7-, 43k,

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST "m'.

TURBINE DRIVEN AUXILIARY FEEDWATER PUMP

a. Steam Generator Water N.A. 1 2 3 Level--Low-low

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

8. LOSS OF POWER

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

b. 4 kv Bus N.A. 1,2,3,4 Degraded Voltage 9.'ANUAL

a. Safety Injection (ECCS) N.A. N.A. N.A. Rt 1,2,3,4.

Feedwater Isolation Reactor Trip (SI)

Containment Isolation-Phase "A" Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System

b. Containment Spray N.A. N.A. N.A. Rt 1,2,3,4 Containment Isolation-Phase "B" Containment Purge and Exhaust Isolation Containment Air Recirculation Fan C ~ Containment Isolation- N.A. N.A. N.A. Rt 1,2,3,4 Phase "A" Containment Purge and Exhaust Isolation

d. Steam Line Isolation N.A. N.A. M(lt) Rt 1 2,3 t The provisions of Technical Specification 4.0.8 are applicable.

'OOK NUCLEAR PLANT - UNIT 2 3/4 3-32 AMENDMENT NO. 8&, &7., 484,

TABLE 4 3-6A APPENDIX R REMOTE SHUTDOWN MONITORING INSTRUMENTATION O SURVEILLANCE RE UIREMENTS O

CHANNEL CHANNEL INSTRUMENT LOCATION CHECK CALIBRATION

1. Steam Generators 1 LSI Cabinet 1 and and 4 Level LSI Cabinet 4

2. Steam Generators 2 LSI Cabinet 2 and and 3 Level LSI Cabinet 4 I

3. Steam Generators 1 LSI Cabinet 4 and and 4 Pressure LSI Cabinet 5

4. Steam Generators 2 LSI Cabinet 4 and and 3 Pressure LSI Cabinet 6

5. Reactor Coolant Loop LSI Cabinet 4 and Rf 4 Temperature (Cold) LSI Cabinet 5

6. Reactor Coolant Loop LSI Cabinet 4 and Rf 4 Temperature (Hot) LSI Cabinet 5

7. Reactor Coolant Loop LSI Cabinet 4 and Rf 2 Temperature (Cold) LSI Cabinet 6

8. Reactor Coolant Loop LSI Cabinet 4 and 2 Temperature (Hot) LSI Cabinet 6 W 9. Pressurizer Level LSI Cabinet 3 ax O

10.

11.

Reactor Coolant System Pressure Charging Cross-Plow Between Units Source Range Neutron Detector (N-23)

Elev. 587'/a LSI Cabinet Corridor 3

LSI Cabinet 4 n/a R'2.

• Charging Cross-Flow between Units is an instrument common to both Unit 1 and 2. This surveillance will only be conducted on an interval consistent with Unit 1 refueling.

The provisions of Technical Specification 4.0.8 are applicable.

A TABLE 4 3-10 O POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS O

CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

1. Containment Pressure M R

2. Reactor Coolant Outlet Temperature - Tso~ (Wide Range) M Rf

3. Reactor Coolant Inlet Temperature - Tcpgg (Wide Range) M Rf Reactor Coolant Pressure - Wide Range M R

5. Pressurizer Water Level M R

6. Steam Line Pressure M R

7.

Steam Generator Water Level - Narrow Range M R

8. RWST Water Level M R

9. Boric Acid Tank Solution Level M R

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

12. PORV Position Indicator - Limit Switches M R

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

14. Safety Valve Position Indicator - Acoustic Monitor M R

15. Incore Thermocouples (Core Exit Thermocouples)(4) M R(1)

16. Reactor Coolant Inventory Tracking System M(2) R(3)t (Reactor Vessel Level Indication)

17. Containment Sump Level* M R

18. Containment Water Level* M Rf W

a (1) Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

W (2) With one train of Reactor Vessel Level Indication inoperable, Subcooling Margin Indication and Core'Exit Thermocouples may be used to perform a CHANNEL CHECK to verify the remaining o Reactor'essel Indication train OPERABLE.

(3) Completion of channel calibration for sensors to be performed below P-12 in MODE 3.

(4) The core exit thermocouples will not be installed until the 1988 refueling outage; therefore, surveillances will not be required until that time. See license amendment dated April 10, 1987.

• 'he requirements for these instruments will become effective after the level transmitters are modified or replaced and become operational. The schedule for modification or replacement of the transmitters is described in the Bases.

The provisions of Technical Specification 4.0.8 are applicable.

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued

2. 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 a total of at least two 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; and apply the portions of ACTION a.2 or a.3 above for inoperable PORVs, relating to OPERATIONAL MODE, as appropriate.

c. With PORVs and block valves not in the same line 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 valves to OPERABLE status or (2) close and de-energize the other valve in each line. Apply the portions of ACTION a.2 or a.3 above, relating to OPERATIONAL MODE, as appropriate for two or three lines unavailable.

d. The provisions of Specification 3.0.4 are not applicable.

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.) I 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 block valve(s) do not have to be tested when ACTION 3.4.1l.a or 3.4.1l.c is 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 a complete cycle of full travel while the emergency buses are energized by the onsite diesel generators and onsite plant batteries. This testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.e and 4.8.2.3.2.d I

• PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this leakage are not considered inoperable.

t The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 4-33 AMENDMENT NO.

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase greater than or equal to 1X of tank volume by

'verifying the boron concentration of the accumulator solution.

C. At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected by removal of the breaker from the circuit.

d. At lease once per 18 months by verifying that each accumulator isolation valve opens automatically upon receipt of a safety injection test signal.f I f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-2 AMENDMENT NO. M, 434

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

d. At least once per 18 months by:f

1. Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is above 600 psig.

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

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

1. Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal.

2. Verifying that each of the following pumps start auto-matically upon receipt of a safety injection test signal:

a) Centrifugal charging pump b) Safety injection pump c) Residual he'at removal pump By verifying that each of the following pumps develops the indicated discharge pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1. Centrifugal charging pump Greater than or equal to 2405 psig

2. 'afety Injection pump Greater than or equal to 1409 psig

3. Residual heat removal pump Greater than or equal to 190 psig g By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:

1. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS sub-systems are required to be OPERABLE.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-5 AMENDMENT NO. W-> 09, ~, 434

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.$

4.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 152'F as determined at least once per hour when any RCS cold leg temperature is between 152'F'and 200'F.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-8 AMENDMENT NO. Q0, 89, 434

CONTAINMENT SYSTEMS 3 4 6 2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWST and transferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION'ith one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 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 />; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be 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 4.6.2.1 Each containment spray syst: em shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked sealed, or otherwise secured in position, is in its correct position.

b. By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 255 psig at a flow of greater than or equal to 700 gpm, when tested pursuant to Specification 4.0.5.

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

1. Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure--High-High test signal.

2. Verifying that each spray pump starts automatically on a Containment Pressure--High-High test signal.

d. At least once per 5 years by performing an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT -'NIT 2 3/4 6-10 AMENDMENT NO.

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

c. At least, once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure--High-High test signal.i

d. At least once per 5 years by verifying a water flow rate of at least 20 gpm (greater than or equal to 20 gpm) but not to exceed 50 gpm (less than or equal to 50 gpm) from the spray additive tank test line to each containment spray system with the spray pump operating on recirculation with a pump discharge pressure greater than or equal to 255 psig.

$ The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 6-12 AMENDMENT NO. 4$ ,

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.6 '.1.2 Each isolation valve specified in Table 3.6-1 shall be demonstrated OPERABLE during the COLD SHUTDOWN or REFUELING MODE at least once per 18 months by:f a ~ Verifying that on a Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

c ~ Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation position.

4.6.3.1.3 The isolation time of each power operated or automatic valve of Table 3.6-1 shall be determined to be within its limit when tested pursuant to Specification 4.0.5 f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 6-14 AMENDMENT NO. W, 434

CONTAINMENT SYSTEMS DIVIDER BARRIER SEAL LIMITING CONDITION FOR OPERATION 3.6.5.9 The divider barrier seal shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

~CTION:

With the divider barrier seal inoperable, restore the seal to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200'F.

SURVEILLANCE RE UIREMENTS

(

4.6.5.9 The divider barrier seal shall be determined OPERABLE at least once per 18 months during shutdown by: f a ~ Removing two divider barrier seal test coupons and verifying that the physical properties of the test coupons are within the acceptable range of values shown in Table 3.6-2.

b. Visually inspecting at least 95 percent of the seal's entire length and:

Verifying that the seal and seal mounting bolts are pro-perly installed, and

2. Verifying that the seal material shows no visual evidence of deterioration due to holes, ruptures, chemical attack, abrasion,, radiation damage, or changes in physical appearances.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 6-47 AMENDMENT NO.

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE when tested pursuant to Specification 4.0.5 by:

a. Verifying that each motor driven pump develops an discharge pressure of greater than or equal to

'quivalent 1240 psig at 60'F in recirculation flow.

b. Verifying that the steam turbine driven pump develops an equivalent discharge pressure of greater than or equal to 1180 psig at 60'F and at a flow of greater than or equal to 700 gpm when the secondary steam supply pressure is greater than 310 psig. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

c. Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in posit'ion is in its correct position.

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 lOX RATED THERMAL POWER. This requirement is not applicable for those portions of the auxiliary feedwater system being used intermittently to maintain steam generator level.

Verifying at least once per 18 months during shutdown 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.f

f. Verifying at least once per 18 months during shutdown 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.f

g. Verifying at least once per 18 months during shutdown that the unit cross-tie. valves can cycle full travel.

Following cycling, the valves will be verified to be in their closed positions.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-6 AMENDMENT NO.

PLANT SYSTEMS 3 4 7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1

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

b. At least one .component cooling water flow path in support of Unit 1 shutdown functions shall be available.

APPLICABILITY: Specification 3.7.3.1.a. - MODES 1, 2, 3, 4.

Specification 3.7.3'.l.b. - At all times when Unit 1 is in MODES 1, 2, 3, or 4.

ACTION'hen Specification 3.7.3.1.a is applicable:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 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 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 />.

When Specification 3.7.3.1.b is applicable:

~

~ ~

With no flowpath in Unit 1 available, return at least one flowpath to available status within 7 days, or provide equivalent shutdown capability in Unit 1 and return at least

~ ~

one flow path to available status within the next 60 days, or have Unit 1 in HOT

~ ~

STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and HOT SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE:

At least o'ce per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked, sealed, or otherwise secured in position, is in its correct position.

b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.f I 4.7.3.2 At least 'once per 18 months during shutdown, verify that the unit cross-tie valves can cycle full travel. Following cycling, the valves will be verified to be in their closed positions.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-12 AMENDMENT NO.

3 4.7.4 ESSENTIAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1

a. At least two independent essential service water loops shall be OPERABLE.

b. At least one essential service water flowpath associated with support of Unit 1 shutdown functions shall be available.

APPLICABILITY: Specification 3.7.4.1.a. - MODES 1, 2, 3, and 4.

Specification 3.7.4.1.b. - At all times when Unit 1 is in MODES 1, 2, 3, or 4.

ACTION'hen Specification 3.7.4.1.a is applicable:

With only one essential service water loop OPERABLE, restore at least two loops to OPERABLE status within 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 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 />.

When Specification 3,7.4.1.b is applicable:

With no essential service water flow path available in support of Unit 1 shutdown functions, return at least one flow path to available status within 7 days or provide equivalent shutdown capability in Unit 1 and return the equipment to service within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and HOT SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.4.1 At least two essential service water loops shall be demonstrated OPERABLE:

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

b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.i I f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-13 AMENDMENT NO.

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. At least once per 18 months by:

1. Verifying'hat the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10X.

2. Verifying that on a Safety Injection Signal from either Unit 1 or Unit 2, or on a containment phase A isolation signal, the system automatically diverts its inlet flow through the HEPA filters and charcoal adsorber bank and that either fan can then be manually started in the re-circulation mode.f

3. Verifying that the system maintains the control room at a positive pressure of greater than or equal to 1/16 inch W. G.

relative to the outside atmosphere at a system flow rate of 6000 cfm plus or minus 10X.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99X of the DOP when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10X.

1 After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99X of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10X.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-16a AMENDMENT NO.

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrated OPERABLE by also verifying that the charcoal adsorbers remove greater than or equal to 99X of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10X.

d. At "least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Water Gauge while operating the ventilation syst: em at a flow rate of 25,000 cfm plus or minus 10X.

2. Deleted.

3. Verifying that the standby fan starts automatically on a Containment Pressure--High-High Signal and directs its exhaust flow through the HEPA filters and charcoal adsorber banks on a Containment Pressure--High-High Signal.f

e. After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99X of the DOP when they are tested in-place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10X.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99X of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10X.

i The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-19 AMENDMENT NO.

PLANT SYSTEMS 3 4.7.7 SNUBBERS LIMITING CONDITION FOR OPERATION 3.7.7.1 All snubbers listed in Table 3.7-9 shall be'OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES).

ACTION:

With one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.7.1.c on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE RE UIREMENTS 4.7.7.1 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a. Visual Ins ectionf The first inservice visual inspection of snubbers shall be per-formed after four months but with 10 months of commencing POWER OPERATION and shall include all snubbers listed in Table 3.7-9.

If less than two (2) snubbers are found inoperable duringvisual the first inservice visual inspection, the second inservice inspection shall be performed 12 months plus or minus 25X from the date of the first, inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:

No. Inoperable Snubbers Subsequent Visual er Ins ection Period Ins ection Period*

0 18 months plus or minus 25X 1 12 months plus or minus 25X 2 6 months plus or minus 25X 3,4 124 days plus or minus 25X 5,6,7 '62 days plus or minus 25X 8 or more 31 days plus or minus 25X The snubbers may be categorized into two groups: Those accessible and those inaccessible during reactor operation. Each group may be inspected independently in accordance with the above schedule.

• The inspection interval shall not be lengthened more than one step at a time.

$j The provisions of Specification 4.0.2 are not applicable.

i The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-20 AMENDMENT NO. &7-, 402i 434

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.7.9.2 Each of the above required water spray and/or sprimkler systems shall be demonstrated to be OPERABLE:

a. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full trav'el as provided by Technical Specification 4.7.9.1.1.e.

b. At least once per 18 months:

1. By performing a system functional test which includes simulated automatic actuation of the system, and:f a) Verifying that the automatic valves in the flow path actuate to their correct positions on a test signal, and*

b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

2. By visual inspection of deluge and preaction system piping (this is not required for systems supervised by air) to verify their integrity.

3. By visual inspection of each open head deluge nozzle to verify that there is no blockage.

c. At least once per 3 years by performing an air flow test through the piping of each open head deluge system and verifying each open head deluge nozzle is unobstructed.

• The fire proteciton water flow surveillance testing may be suspended until the completion of the fire protection water stroage tank and fire pump installations (May 31, 1993). The surveillance testing suspended as a result of this amendment will be initiated at its normal frequency within four months .

of the new fire protection water storage tanks and fire pumps being declared OPERABLE, with the exception of unit outage required testing which would be completed before the end of the next scheduled outage.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-40 AMENDMENT NO. M, ~p 43kf

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS Continued a) A kinematic viscosity of greater than or equal to 1.9 centistokes but less than or equal to 4.1 centistokes at 40'C (alternatively, Saybolt viscosity, SUS at 100'F of greater than or equal to 32.6 but less than or equal to 40.1), if gravity was not determined by comparison with supplier's certification.

b) A flash point equal to or greater than 125'F.

2) By verifying, in accordance with the test specified in ASTM D1298-80 and prior to adding the new fuel to the storage tanks, that the sample has either an API gravity of greater than or equal to 30 degrees but less than or equal to 40 degrees at 60'F or an absolute specific gravity at 60/60'F

. of greater than or equal to 0.82 but less than or equal to 0.88, or an API gravity of within 0.3 degrees at 60'F when compared to the supplier's certificate or a specific gravity of within 0.0016 at 60/60'F when compared to the supplier's certificate.

3) By verifying, in accordance with the test specified in ASTM D4176-82 and prior to adding new fuel to the storage tanks, that the sample has a clear and bright appearance with proper color.

4) By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are within the appropriate limits when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D2622-82.

d. At least once per 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-83, and verifying that total particulate contamination is less than 10 mg/liter when tested in accordance with ASTM D2276-83, Method A+.

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

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

• The actions to be taken should any of the properties be found outside of the specified limits are defined in the Bases.

t The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 8-4 AMENDMENT NO.

ELECTRICAL POWER SYSTEMS SHUTDOWN 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 Class lE distribution system, and

b. One diesel generator with:

1. A day fuel tank containing a minimum of 70 gallons of fuel,

2. A fuel storage system containing a minimum indicated volume of 46,000 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

ACTION'ith less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until the minimum required'.C. electrical power sources are restored to OPERABLE status.

SURVEILL'ANCE RE UIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except: for requirement 4.8.1.1.2.a.5.i

• For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 8-9 AMENDMENT NO.

Attachment 3 to AEP:NRC:1181 Existing Technical Specifications for Donald C. Cook Nuclear Plant Unit 2 Marked to Reflect the Proposed Changes

3 4.0 APPLICABILITY SURVEILLANCE RE UIREMEHTS 4.0.8 By specific reference to this section, those surveillances which must be performed on or before August 13, 1994,. and are designated as 18-month or 36-month surveillances (or required as outage-related surveillances under the provisions of Specification 4.0.5) may be delayed until the end of the cycle 9-10 refueling outage.

For these specific surveillances under this section, the specified time intervals required by Specification 4.0.2 will be determined with the new initiation date established by the surveillance date during the Unit 2 1994 refueling outage.

COOK NUCLEAR PLANT - UNIT 2 3/4 0-4 AMENDMENT NO.

~ ~

REAGTIVITY CONTROL SYSTEMS POSITION INDICATOR CHANNELS-SHUTDOWN LIMITING CONDITION FOR OPERATION r

3.1.3.3 Ae lease one.rod position indicator channel (excluding demand.

position indication) shall be OPERABLE for each shutdown'or control rod'not.

fully inserted. c c

APPLICABILITY: MODES 3+4, 4M and 5M ACTION:

With;.less than ehe above required position indicator channel(s). OPERABLE, I ~

'immediately open the reactor trip system breakers. s c

SURVEILLANCE UIREMENTS c I I 4.1.3.3: Each of ehe above required rod position indicator channel(s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST ae least once per 18 months. f I ~

I

~ ~

\~

~, s I 'r c.r ~ ~

+With. the reactor. trip syst: em breakers in the closed position.

~, 1 CSee Special Test Exception 3.10.5.

sgc'iclcsws cf'ccfcci(J 5icccific o:4'rr SiC,a ccc Sct(ilicc/4,.

COOK NUCLEAR PLANT - UNIT 2 3/4 1-22 AMENDMENT NO.gP,P),~

POWER DISTRIBUTION LIMITS DNB AND Tav OPERATINC PJJVAETERS LIMITINC CONDITION FOR OPERATION 3,2.5 The following DNB related parameters shall be maintained vithin the follo~ing operational indicated limits:

a. DNB 0

1. Reactor Coolant System T avg Less than or equal to 578.7 Fe

2. Pressurizer Pressure Creater than or equal to 2200 psig+/~

3. Reactor Coolant System Greater than or equal to 366,400 gpm***

Total Flov Rate

b. T avg 0

1. Reactor Coolant System T Greater than or equal to 543,9 F>>

avg APPLICABILITY'ODE 1 ACTION:

With any of the above parameters exceeding its limit, restore the parameter to vithin its limit vithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 54 of RATED THERMAL POWER vithin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE RE UIREMENTS 4.2.5.1 Each of the above parameters shall'e verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The indicators used to determine RCS total flov shall be sub]ected to a CHANNEL CALIBRATION at least once per 18 months.t 4.2.5.3 The RCS total flov rate shall be determined by a pover balance around the steam generators at lease once per 18 months.

4.2.5.4 The provisions of Specification 4.0.4 shall not apply to primary flov surveillances.

• Indicated average of at least three OPERABLE instrument loops.

Limit not applicable during either a THERMAL POWER ramp in excess of 54 of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of

~ 10% of RTP Indicated value COOK NUCLEAR PLANT - UNIT 2 3/4 2-15 AMENDMENT N0.82.~

0 3 4. 3 INSTRUMEIG'ATION 3 4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1.1 As a minimum, the reactor trip system instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY: As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE UIREMEHTS 4.3.1.1.1 Each reactor trip system instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations during the MODES and at the frequencies shown in Table 4.3-1. f 4.3.1.1.2

~ ~ ~ The logic f'r the interlocks shall be demonstrated OPERABLE prior to

~

each reactor startup unless performed during the preceding 92

~

days. The total interlock function shall be demonstrated OPERABLE at

~

least once per 18 months during CHANHEL CALIBRATION testing of each channe affected by interlock operation.t 4.3.1.1.3 The REACTOR TRIP SYSTEM RESPONSE TIME of each reactor trip function shall be demonstrated to be within its limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at, least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.t COOK NUCLEAR PLANT - UNIT 2 3/4 3-1 AMENDMENT NO.7$ ,97,~

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS V.

CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT 4 ~ 'I CHECK CALIBRATION TEST RE(EUIRED 70

1. Manual Reactor Trip t

'h A. Shunt Trip Function N.A. N.A. S/U(l)(lo) 2, 3*, 4*, 5*

B. Undervoltage Trip N.A. N.A. S/U(l)(10) 3* 4*, 5*

L Function

2. Power Range, Neutron Flux D(2,8),M(3,8) M and S/U(1) 1, 2 and *

<<) and Q(6,8)

3. Power Range, Neutron Flux, N.A. R(6) 1 2 High Positive Rate

4. Power Range, Neutron Flux,; N;A. R(6) 1, 2 High Negative Rate

5. Intermediate Range.;=: R(6,8)T 1, 2 and

• Neutron Flux

6. Source Range, Neutron Flux 7, Overtemperature AT R(9)0 t'/U(1)

R(6,14) M(14) and S/U(1) 2(7), 3(7),

4 and 5 1, 2

8. Overpower AT S R(9)% 1, 2

9. Pressurizer Pressure--Low RT'... 1, 2

10. Pressurizer Pressure--High RV 1, 2 ll. Pressurizer Water L'evel--High,: 1, 2 0

12. Loss oE Flow - Single Loop R(8)

~ 0 p~v15)4~ crf Rc4'hl c J Spic l6c a+~~ 4iOa~ o-~ cLpP') >c C~~4r

TABLE 4.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODES IN WHICH CHANNEL CHANNEL -FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST ~RE UIRED 13.. Loss of Flow-Two Loops R(8)$ N.A.

14. Steam Generator Water Level S 1, 2 Low-Low

15. Steam/Feedwater Flow Mismatch and S 1/ 2 Low Steam Generator Water Level

16. Undervoltage-Reactor Coolant N.A.

Pumps

17. Underfrequency-Reactor Coolant N.A.

Pumps

18. Turbine Trip A. Low Fluid Oil Pressure N.A. N.A. S/U(1) 1, 2 B. Turbine Stop Valve Closure N;A. N.A. S/U (1) 1, 2

19. Safety In)ection Input from ESF N.A. N.A. M (4). 1/ 2

20. Reactor Coolant Puqp Breaker N.A. N.A. N.A.

Position Trip '!

21. Reactor Trip Breaker A. Shunt Trip Function N.A. H(5) (11) and S/U(1) (l l) 3* 4* 5*

B. Undervoltage Trip Function N.A. N.A. M(5) (ll) and S/U(1) (11) ] 2 3* 4* 5*

22. Automatic Trip Logic N.A. N.A. M(5) 2 3* 4* 5*

23. Reactor Trip Bypass Breaker N.A. M(12) and S/U(1) (13) 3* 4* 5*

~)~ p~g>s>o~ ~ jgd.~ ~ ( %~~ii~ 9'~ 4O ~ a.p~ dp ply'c. 44.,

INSTRUMENTATION el' 4.3.2 ENGINEERED SAFETY FEATURE ACTUATION'YSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumenta-tion channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown the'rip in Table 3.3-5:

APPLICABILITY: As shown in Table 3.3-3.

ACTION a ~ With an ESFAS instrumentation channel trip setpoint less conserva-tive than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint .value.

b. With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

SURVEILLANCE RE UIREMENTS 4.3.2.1'.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION, CHANNEL FUNCTIONAL TEST and TRIP ACTUATING DEVICE OPERATIONAL TEST operations for the MODES and at the frequencies. shown in Table 4.3-2.f 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test, The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.C 4.3.2.1.3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months. Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times'8 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" column of Table 3 3;3. r

~;ae 'im. a4"TecLnimJ Spec r'cgkim 60'0 ~ ~~pl ~~ +.',, l'f,

- UNIT 2 COOK NUCLEAR PLANT 3/4 3-14

~

AMENDMENT NO. 70 97, $ 81,

TABLE 4.3-2 ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIRE".ANTS TRIP l ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE (-,

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST m"'.

SAFETY INJECTION, TURBINE TRIP, FEEDVATER ISOLATION, AND MOTOR DRIVEN AUXILIARY.

FEEDVATER PUMPS

a. Manual Initiation See Functional Unit

b. Automatic Actuati.on Logic N.A. N.A. M(2) N'. A':

9

'1 2,: '3,'4

c. Contai.nment Press- M(3) N.A. 1, 2, 3 ure-High

d. Pressurizer Press- N.A.

ure-Lov .

e, Differential Press- N.A.

N.A'.'---

1, 2, 3 ure Between Steam Lines--High f.~ Steam Line Pressure-- M Lov 2.~ CONTAINMENT SPRAY

a. Manual Initiation - See Functional Unit 9

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

c. Containment Press- M(3) N.A. 1, 2, 3 ure-High-High

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1) Manual ear --------- See Functional Unit 9

2) From Saf'ecy N.A. . N.A. M(2) ' N.A. 1, 2, 3, 4 In] ection Automatic Actuation Logic

b. Phase "B" Isolation

1) Manual See Functional Unit

2) Automatic Actua-tion Logic N.A. N.A. M(2) '.A. 9 1, 2, 3, 4

3) Containment Press- S R M(3) N.A. 1, 2, 3 ure-High-Hi.gh c.pr~vist~s ~

~4 }echriicQ 5pc.c.i'Sic,c ft~ 40 2 c.~ ~pc,'c,e.44,, ~

COOK NUCLEAR PLANT - UNIT 2 3/4 3-30 AMENDMENT NO. 34 fQ i

~ >>

TABLE 4. 3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UZREMENTS TRIP ACTUA.TING MODES IN

%RNNEL, DEVICE WHICH QKHNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST M

c. Purge and Exhaust Isolation ')

Manual Seo Functional Unit 9

2) Containment Radio- M N.h. 1, 2, 3, 4 activity-High

4. STEAM LINE ISOLATION a>> Manual See Functional Unit 9

b. Automatic Actuation N.h. N.h. M(2) N A 1, 2, 3 Logic c Containment Press-

~

M(3) N.h. 1,2;3 ure--High-High Steam Flow in N.h. 1,2,3 Two Steam Lines--

High Coincident wi.thl

- T ave Steam Low

- -Low-Low Line Pressure-- N.h. 1, 2, 3

5. TURBINE TRIP AND FEEDWATER ISOLATION

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

6. MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Steam Generator N.h. 1, 2, 3 Water Level--Ler-Low

b. 4 kV Bus N.h. 1, 2, 3 Loss of Voltage

c. Safety Infection N.h. N.h. M(2) N.h. 2, 3

d. Loss of Main Pood N.h. N.h. Rt N.h. 1, 2 Pumps

~ Prov)5levU K Qc,c4AiGJ 5P<<~<"

NUCLEAR PLANT UNIT, 2. 3/4 3<<31 AMENDMENT NO. 82, 97, ggf, fgg, AT

TABLE 4. 3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM ZNSTRUMENTATZON SURVEILLANCE RE UIREMENTS TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHAHNEL KKNNEL FUHCTZONAL OPERATIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CBZTBRILTIOH TEST TEST m

7. TURBINE DRIVEN AUXILIARY FEEDWATER PUMP

a. Steam Generator Water S. N.h. 1, 2, 3 Level--Lov-los

b. Reactor Coolant Pump N.h.. H.h. 1, 2, 3 Bus Undervoltage

8. LOSS OF POWER

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

b. 4 kv Bus H.h. 1, 2, 3, Degraded Voltage.

4',

9. Manual Safety Injection (ECCS) N.h. N.h. N.h. 2, 3, 4

'eedvater Isolation Reactor Trip (SZ)

Containment Zsolation-Phase "h" Containment Purge and Exhaust Isolation Auxi.liary Feedvater Pumps Essential Service Water System b, Containment Spray N.h. H.h. H.h; 1, 2, 3, 4 Containment Isolation-Phase "B" Containment Purge and Exhaust Isolation Containment hir Recirculation Fan

c. Containment Isolation- N.h. N.h. N.h.

Phase "h" Containment Purge and Exhaust: Isolation

d. Steam Line Isolation H.h. H.h. H(lgt ) 1, 2, 3 c pre,Jism~ of ~hcLAi c.cJ 5pcc.qlicm4(4M f Oil~ a rE 4'-t I C .%CLEAR PLANT - UNIT 2 3/4 3-32 AMENDMENT HO. 82 97, fNB

t) 0 8

TA13I.E.4. 3-6A AI PEHnlX R REMOTE SNurOOWH MONITORING INSTRUMENTATION SURVEII.LANCE Rl'. UIREMI'.HTS CllANHEL CllANHEL

) ) +~ IHSTRUMEHT LOCATION CllECK CALIBRATION

l. Steam Generators 1 LSI Cabinet 1 and and 4 Level I.SI Cabinet 4 I

2 Steam Generators 2 LSI Cabinet 2 and and 3 Level LSI Cabinet: 4 Steam Generators 1 LSI Cabinet 4 and a>>d 4 Pressure LSI Cabinet 5

4. Steam Generators 2 LSI Cabinet 4 and and 3 Pressure LSI Cabinet 6

5. Reactor Coolant Loop LSI Cabinet 4 and Rt 1 4 Temperature (Cold) LSI Cabinet 5

6. Reactor Coolant Loop LSI Cabinet 4 and 4 Temperature (Hot) IXI Cabinet 5

7. Reactor Coolant Loop ISI Cabinet 4 and Rt 2 Temperature (Cold) LSI C:)bi>>et 6 Reactor Coolant Loop LSI Cabinet 4 and R4 2 Temperature (llot) LSI Cabinet 6

9. Pressurizer Level LSX Cabinet 3

10. Reactor Coolant System Pressure LSI Cabinet 3 R

11. Charging Cross-Plow Corridor n/a

'4 0 i Between Un t:s. Elev.

587'SI

12. Source Range Neutron Detector (N-23) Cabit)et 4 n/a Charging Cross-Plow between Units is a>> instru)nent connno>> to both Unit 1 and 2. This surveillance will onlv

~

be co>>d)~<:d o>> ')>> i>>terval consiste>>t: wi t:h Unit 1 regueli~

"prov st~ g$ gk >ecAn c. l gp.c W)c.c%)'~ 46,8 o.m....c),ppl cQle .

TABI.E 4.3-1O POST-ACCIDENT MONIT RING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALlBRATION

l. Containment Pressure M R

2. Reactor Coolant Outlet Temperature - T (Wide Range) M RT

3. Reactor Coolant Inlet Temperature - T HOT (Wide Range) M R4

4. Reactor Coolant Pressure - Wide Range COLD M R

5. Pressurizer Water Level M R Steam Line Pressure M R

7. Steam Ge'nerator Water Level - Narrow Range M R

8. RWST Water Level M

9. Boric Acid Tank Solution Level M

10. Auxiliary Feedwater Flow Rate M R

11. Reactor Coolant System Subcooling Margin Monitor M- Rf

12. PORV Position Indicator - Limit Switches M R

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

14. Safety Valve Position Indicator - Acoustic Monitor M R

15. Incore Thermocouples (Core Exit Thermocouples)(4) M R(1)

16. Reactor Coolant Inventory 'Tracking System M(2) R(3)4 (Reactor Vessel Level Indication)

17. Containment Sump Level* M R

18. Containm'ent Water Level* M R t 1

(1) Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

(2) With one train of Reactor Vessel Level Indication inoperable, Subcooling Margin Indication and Core Exit Thermocouples may be used to perform a CHANNEL CHECK to'verify the remaining Reactor Vessel Indication train OPERABLE.

(3) Completion of channel calibration for sensors to be performed below P-12 in MODE 3.

(4) The core exit thermocouples will not be installed until the 1988 refueling outage; therefore, surveillances will not be required until that time. See license amendment dated April 10, 1987.

The requirements for these instruments will become effective after the level transmitters are modi.fied or replaced and become operational. The schedule for modification or replace>nent of the transmitters is described in the Bases.I her gt 0 i>s>cion < + <c, Ch>q> cat 'SpecIS> c~+>~ Q.Oi8'JAM +pt> l'~

COOK NUCLEAR PLANT - UNIT 2 "3/4 "3.-4 7 AMENDMENT NO. gg, gg~f /'

l CTOR COOLANT SYST M LIMXTINC CONDITION POR OPTATION Continued

2. Vith tvo or mo e, block valves inoperable, Within 1 hou either (1) restore a total of at least tvo block valves to OPERABLE status, or (2) close the, Mock valves and remove pover from the. block valves, or (3) close the associated PORVs and remove pover fzom their associated solenoid valves and apply the portions of ACTION a.2 or a.3 above for inoperable PORVs," relating to OPERATIONAL MODE, as appropriate.

c. Vi.th PORVs and block valves not in the same line inoperable,+

vithin 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 valves to OPHMLE status or (2) close and de-enezgi=e the other valve in each line. Apply the portions of ACTION a.2 or a.3'above, relating to OPEBATIONAL MODE, as appropriate or tvo or three lines unavailable.

The provisions of Specificati.on 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4. 11.1 Each of .the three PORVs shall be demonstrated OP~LE:

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

't least once per 18 months by performance of a CHANNEL CALIBRATION.Y 4.4.11.2 Each of the. three block valves shall be demonstrated OPEBABLE at least once per 92 days by operating the valve through one complete cycle of full travel. The block valve(s) do not have to be tested vhen ACTION 3.4.11.a or 3.4.ll.c is applied.

4 4.11.3 The emergency paver supply'or 'the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a. complete cycle of full travel vhile the emergency buses are energized by the onsite diesel generators and onsite plant batteries. This testing can be w.

performed in con)unction vith the requirements of Specifications 4.8.3..1.2.e and 4.8.2.3.2.d.g I

• PORVs isolated to limit BCS leakage through their "eats and the block valves shut to isolat~ this leakage are not considered inoperable.

COOK NUCLEAR PLANT - UNIT 2 3/4 4-33 AMENDMENT NO'. PA

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

b. At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after each solution volume increase greater than or equal to 18 of tank volume by verifying the boron concentration of the accumulator solution.

c ~ At least once per 31 days when the RCS pressure is above 2000 psig by verifying that power to the isolation valve operator is disconnected by removal of the breaker from the circuit.

d. At lease once per 18 months by verifying that each'accumulator isolation valve opens automatically upon receipt of a safety injection test signal. t p~Vis~an,s of'c J ni'c~l

~ ~

Spcct'Fice.s+l~ O'OiR'~ c )plI c.chal.,

COOK NUCLEAR PLANT - UNIT 2 5-2 AMENDMENT NO.g ~'/4

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued

d. At least 'once per 18 months by:t Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System when the Reacto'r Coolant Sy'tem pressure is above 600 psig. . ~> /

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

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

1. Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal.

2. Verifying that each of the following pumps start auto-matically upon receipt of a safety injection test signal:

a) Centrifugal charging pump .

b) Safety injection pump c) Residual heat removal pump

f. By verifying that each A

of the following pumps develops the indicated

'discharge pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1. Centrifugal charging pump Greater than or equal to 2405 psig

2. Safety Injection pump Greater than or equal to 1409 psig

3. Residual heat removal pump Greater than or equal to 190 psig By verifying. the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:

1. Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> following completion of each valve stroking operation or maintenance on the valve when the ECCS sub-systems are required to be"OPERABLE.

f, (({ PcM)s)~ <<+ gQ~~, c~( +peg ~g ~c+i~ Q,oa g o-~ ckPP sccc44 COOK NUCLEAR PLANT - UNIT 2 3/4 5-5 AMENDMENT

~

N0.97>> 99~

f31,

EHERGENCY CORE COOLING SYSTEHS SURVEILLANCE RE UIREHENTS 4.5.3 1 The ECCS subsystem shall be demonstrated

~ OPERABLE per the applicable Surveillance Requirements of 4.5.2.f'.5.3.2 All charging pumps and safety injection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whenever the temperature of one or more of the RCS cold legs is less than or equal to 152 F as determined at least once 0 0 per hour when any RCS cold leg temperature is between 152 F and 200 F.

props>o>s oF WachnlccJ ~~ill(.cd>~ 40.$ ~~ happ l< c a-44.,

COOK NUCLEAR PLANT - UNIT 2 3/4 5-8 AMENDMENT NO. PP,Pg,~

CONTAINMENT SYSTEMS 3 4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent containment spray systems shall be OPERABLE with each spray system capable of taking suction from the RWST and transferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With one containment spray system inoperable, restore the inoperable spray system to OPERABLE status within 72 hours8.333333e-4 days <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 />; restore the inoperable spray system to OPERABLE status within the next 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be 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 />.

SURVEILIANCE REQUIREMENTS 4.6.2.1 Each containment spray system shall be demonstrated OPERABLE:

a ~ At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked sealed, or otherwise secured in position, is in its correct position.

b. ~

By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 255 psig at a flow of greater than or equal to 700 gpm, when tested pursuant to Specification 4.0.5.

c ~ At least once per 18 months during shutdown, by:f Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure--High-High test signal.

2. Verifying that each spray pump starts automatically on a Containment Pressure--High-High test signal.

At least once per 5 years by performing an air or'smoke flow test "

through each spray header and verifying each spray nozzle is unobs true ted.

)Pov l5/II)',s a5 Ec4A lc'.J SPcc I'FIc&loA Foa 8 b.~ cN PP!i c> e ~

COOK NUCLEAR PLANT - UNIT 2 3/4 6-10 AMENDMENT NO. '$7~'

f CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

c. At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure--High-High test signal.<

I vl

d. "

At least once per 5 years by verifying a water flow rate of at least byline 20 gpm (greater than or equal to 20 gpm) but not to exceed 50 gpm (less than or equal to 50 gpm) from the spray additive tank test to each containment spray system with the spray pump operating on recirculation with a pump discharge pressure greater than or equal to 255 psig.

Ac frovl5twM csF Ic~nlcg Spec > Pic'r+J~ WiOiS a-rc mpp(i r c 44..

COOK NUCLEAR PLANT - UNIT 2 3/Li 6-12 AMENDMENT NO.PP.,)7,~

CON1'AINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued 4.6.3.1.2 Each isolation valve specified in Table 3.6-1 sh~ll be demonstrated OPERABLE during the COLD SHUTDOWN or. REFUELING MODE 'at least once per 18 months by: t

a. Verifying that on a 'Phase A containment isolation test signal, each Phase A isolation valve actuates to its isolation position.

b. Verifying that on a Phase B containment isolation test signal, each Phase B isolation valve actuates to its isolation position.

C. Verifying that on a Containment Purge and Exhaust isolation signal, each Purge and Exhaust valve actuates to its isolation position.

4.6.3.1.3 isolation time of each power operated or automatic valve of Table 3 '-1Theshall be determined to be within its limit when tested pursuant to Specification 4.0.5 g~ ppogigg ~ f Qc.cQ cq.l 5(cc.kcchm 4 o - A~ eppl~c a~+

COOK NUCLEAR PLANT - UNIT 2 3/4 6-14 AMENDMENT NO.).7 ~

CONTA'INMENT SYSTEMS DIVIDER BARRIER SEAL LIMITING CONDITION FOR OPERATION 3.6.5.9 The divider barrier seal shall be, OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTION:

With the divider barrier seal inoperable, restore the seal to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200 F.

SURVEILLANCE UIREMENTS 4.6.5.9 The divider barrier seal shall be determined OPERABLE at least once per 18 months during shutdown by:+

a. Removing two divider barrier seal test coupons and verifying that the physical properties of the test coupons are within the acceptable range of values shown in Table 3.6-2.

. b. Visually inspecting at least length and:

95 percent of the seal's entire

1. Verifying that the seal and seal mounting bolts are pro-perly installed, and

2. Verifying that the seal material shows no visual evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearances.

PI ~V>~iAnS ~< ~C~LniCJ 5PCC,'iR'~i~ 40 8 0-r O-PPli I-+4~ i COOK NUCLEAR PLANT - UNIT 2 3/4 6-47 AMENDMENT N0.7$ ~

C RE 0 REMENTS

,7.1.2 Each auxili.ary feedwater pump shall be demonstrated OPERA3LE when tested pursuant to Specification 4.0.5 by".

a. Verifying that each motor driven pump develops an equivalent di.scharge pressure of greater than or equal to 1240 psig at 60 P in recirculation flow.

b. Vezifying that the steam turbine driven pump develops an equivalent di.scharge pressure of greater than or equal to 1180 psi.g at 60 P and at a flow of greater than or equal to 700 gpm when the secondary steam supply pressure i.s greater than 310 psig. The provisions of Specification 4.0.4 are not appli.cable for entry into MODE 3.

Co Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position i.s in its correct position.

d. 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 10t RATED THERMAL POVER.

This requi.rement is not applicable for those portions of the auxiliary feedwater system being used intermittently to maintain steam generator water level.

Verifying at least once'er 18 months during shutdown that each g ~

automatic valve in the flow path actuates to its correct position upon recei.pt of the appropriate engineered safety .features actuation test signal required by Specification 3/4.3.2.+

eri.fying at least once per 18 months during shutdown that each auxiliary feedwater pump starts as designed automatically upon receipt of the appropriate engineered safety features actuation test signal required by Specifi.cation 3/4.3.2. t go Vezifying at least once per 18 months during shutdown that the unit cross-tie valves can cycle full travel. Pollowtng cycling, the valves will be verified to be in their closed positions.

>4 pe~tsi~~~ ~ Ec4nicJ 5pcci4ic~~~ il:o.~ etc Qpii&4..

NUCLEAR PLANT - UNIT 2 3/4 7-6 AMENDMENT NO. Mg M ~~ j

)

LANT SYSTEMS 3 4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3.1

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

b. At least one component cooling water flow path in support of Unit 1 shutdown functions shall be available.

APPLICABILITY: Specification 3.7.3.1.a. - MODES 1, 2, 3, 4.

Specification 3.7.3.1.b. - At all times when Unit 1 is in MODES 1, 2, 3, or 4.,

ACTION:

When Specification 3.7.3.l.a is applicable:

With only one component cooling water loop OPERABLE, restore at least two loops to OPERABLE status within 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 /> 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 />. 'nd When Specification 3.7.3.1.b is applicable:

th no flowpath in Unit 1 available, return at least one flowpath to available status

~ ~

within 7 days, or provide equivalent shutdown capability in Unit 1 and return at least

~ ~ ~

one flow path to available status within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and HOT SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.3.1 At least two component cooling water loops shall be demonstrated OPERABLE:

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

b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.f 4.7-3.2 At least once per 18 months during shutdown, verify that the unit cross-tie va1ves can cycle full travel. Following cycling, the valves will be verified to be in their closed positions.

~i~s ~f'~~s cpJ ~jf>e e.hem 40 F ~~ ~PP~l ow Qwc~Atz. P~N~ - uu> c 3/4 7-12 Amendment No. 97, /f8, /

/4.7.4

~ ~ ESSENTIAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1

a. At least two independent essential service water loops shall be OPERABLE.

b. At least one essential service water flowpath associated with support of Unit 1 shutdown functions shall'e available.

APPL'ICABILITY: Specification 3.7.4.1.a. - MODES 1, 2, 3, and 4.

Specification 3.7.4.1.b. - At all times when Unit 1 is in MODES 1, 2, 3 or 4.

ACTION:

When Specification 3.7.4.l.a is applicable:

With only, one essential service water loop OPERABLE, restore, at least two loops to OPERABLE status within 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 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 />.

When specification 3.7.4.l.b is applicable:

'th no essential service water flow path available in support of Unit 1 shutdown ctions, return at least one flow path to available status within 7 days or provide equivalent shutdown capability in Unit 1 and return the equipment to service within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and HOT SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The requirements of Specifications 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.4.1 At least two essential service water loops shall be demonstrated OPERABLE:

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

b. At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.j.

pn> si~ W IccJini~y I ~c i%i c.o+i<h g,c. 9 n.~ n.pp] c dole,.

oryJQG(EAa. /'LAND - uuii Z 3/4 7-13 Amendment No.97, fN, ]

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS Continued

e. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsozber banks is less than 6 inches Water Gauge while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10$ .

2. Verifying that on a Safety Injection Signal from either Unit 1 or Unit, 2, or on a containment phase A isolation signal, the system automatically diverts its inlet flow through the HEPA filters and charcoal adsorber bank and that either fan can then be manually started in the re-circulation mode. 0

3. Verifying that the system maintains the control room at, a posfckve pressure of greaser than or equal ro 1/16 inch W. G.

relative to the outside atmosphere at a system flow rate of 6000 cfm plus or'inus 10%.

After each complete or partial replacement of a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in-place in accordance with ANSI N510'-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10%.

g ~ After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant'est gas when they are tested in-place in accordance with ANSI N510-1975 while operating the ventilation system at a flow rate of 6000 cfm plus or minus 10$ .

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COOK NUCLEAR PLANT - UNIT 2 3/4 7-16a AMENDMENT NO. P7 ~+

PLAN1, SYSTEMS SURVEILLANCE RE UIREMENTS Continued b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at, least two inches in diameter and with a length equal to the thickness of the bed.

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrated OPERABLE by also verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10$ .

d. At least once per 18 months by:

1. Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than 6 inches Mater Gauge while operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10%.

2. Deleted.

I'.

Verifying that the standby fan starts automatically on a Containment Pressure--High-High Signal and directs its exhaust flow through the HEPA filters and charcoal adsorber banks on a Containment Pressure--High-High Signal. g

e. After each complete or partial replacement o'f a HEPA filter bank by verifying that the HEPA filter banks remove greater than or equal to 99% of the DOP when they are tested in-place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 25,000 cfm plus or minus 10%.

After each complete or partial replacement of a charcoal adsorber bank by verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant. test gas when they are tested in-place in accordance with ANSI N510-1980 while operating .the ventilation system at a flow rate of 25,000 cfm plus or minus 10%.

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COOK NUCLEAR PLANT - UNIT 2 3/L4 7-19 AMENDMENT NO. ETEs~~

PLANT SYSTEMS LIMITING CONDITION FOR OPERATION 3.7.7.1 All snubbers listed in Table 3.7-9 shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3 and 4. (MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES).

ACTION:

With one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.7.1.c on the supported component or declare the supported..

system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE RE UIREMENTS 4.7.7.1 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a. Visual Ins ection'he first inservice visual inspection of snubbers shall be per-formed after four months but with 10 months of commencing POWER OPERATION and shall include all snubbers listed in Table 3.7-9.

If less than two (2) snubbers are found inoperable during the first inservice visual inspection, the second inservice visual inspection shall be performed 12 months plus or minus 25%

from the date of the first inspection. Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:

No. Inoperable Snubbers Subsequent Visual er Ins ection Period Ins ection Period+4 0 18 months plus or minus 25%

1 12 months plus or minus 25%

2 6 months plus or minus 25%

3,4 124 days plus or minus 25%

5,6,7 62 days plus or minus 25%

8 or more 31 days plus or minus 25%

The snubbers may be categorized into two groups: Those accessible and those inaccessible during reactor operation. Each group may be inspected independently in accordance with the above schedule.

• The inspection interval shall not be lengthened more than one step at a time.

0 The provisions of ~S ecigicption 4.0.2 pre not applicable.

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COOK NUCLEAR PLANT - UNIT 2 . 3/4 7-20 .AMENDMENT NO.97~797~

FIANT SYSTEMS SURVE LLANCE RE UTREMENTS Continued 4.7.9.2 Each of the above required water spray and/or sprinkler'ystems shall be demonstrated to be OPERABLE:

At least once per 12 months by cycli.ng each testable valve in the flow path through at least one complete cycle of full travel as provided by Technical Specification 4. 7. 9. l.l. e.

b. At least once per 18 months:

By performing a syseem functional test whichI includes simulated automatic actuation of the system, and: 7 a) Verifying that the automatic valves in the flow path actuate to their correct positions on a test signal, and+

b) Cycling each valve in the flow path that is noe testable during plant operation through at least one complete cycle of full travel.

2. By visual inspection of deluge and preaction system piping (this is not required for systems supervised by air) to verify their integrity.

3. By visual inspection of each open head deluge nozzle eo verify that there is no blockage.

c ~ At least once per 3 years by performing an air flow test through ehe piping of each open head deluge system and verifying each open head deluge nozzle is unobstructed.

• The fire protection water flow surveillance teseing may be suspended uneil the completion of the fire protection waeer storage eank and fire pump installations (May. 31, 1993). The surveillance testing suspended as a result of this amendment will be initiated at its normal frequency within four monehs of the new fire proeection water seorage tanks and fire pumps being declared OPERABLE, with the exception of unit outage required testing which would be completed before the end of the next scheduled unit outage.

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3/4 7-40 AMENDMENT NO. W', 434) ~) ~)

ELECTRICAL POMER SYSTEMS

<'VRVEILLANCE RE VIREHENTS Continued a} A kinematic viscosity of greater than or equal to 1.9 centistokes bIIt less than or equal to 4. 1 centistokes at 40 C (alternatively, Saybolt viscosity, SUS at 100 F of greater than or equal to 32.6 but less than or equal to 40.1), if gravity was not determined by comparison with supplier's certification.

b) A flash point equal to or greater then 125 F.

2) By verifying, in accordance with the test specified in ASTH D1298-80 and prior to adding the new fuel to the storage tanks, that the sample has either an API gravity of greater than or equal to 30 degrees but less than or equal to 40, degrees at 60 F or an absolute specific gravity at 60/60 F of greater than or equal to 0.82 but less than or e8ual to

, 0.88, or an API gravity of within 0.3 degrees at 60 F when compared to the supplier's certificate or a specific gravity of within 0. 0016 at 60/60 F when compared to the

'-supplier's certificate. I

3) By verifying, in accordance with the test specified in ASTH D4176-82 and prior to adding new fuel 'to the storage tanks, that the sample has a clear and bright appearance with proper color.

4) By verifying within 31 days of obtaining the sample that the other properties specified in'able 1 of ASTH 0975-81 are within the appropriate limits when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D2622-82.

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d. At least once per 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTH 02276-83, and verifying that total particulate contamination is less than 10 mg/liter when tested in accordance with ASTM D2276-83, Hethod A".

e. At least once per 18 months, during shutdown, by: 4

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

" he act)ons to be taken should any of the propert>es be tound outs>de of the specified limits are defined in the Bases.

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3/4 8-4 AMENDMENT NO. ~

ELECTRICAL POWER SYSTEMS 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 Class 1E distribution system, and

b. One diesel generator with:

A day fuel tank containing a minimum of 70 gallons of fuel,

2. A fuel storage system containing a minimum indicated volume of 46,000 gallons of fuel, and

3. A fuel transfer pump.

APPLICABILITY: MODES 5 and 6.

ACTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until the minimum required A.C. electrical power sources are restored to OPERABLE status.

SURVEILLANCE RE UIREMENTS 4.8.1.2 The above required A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 and 4.8.1.1.2 except for requirement 4.8.1.1.2.a.5. i )

• For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

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COOK NUCLEAR PLANT - UNIT 2 AKM)MENT NO 27(~<

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Page 2 of 2 FORM NIS-2 (Back)

Remarks UNIT 2 REPLA ED DISC IN VALVE 2-CS- W Applicable Manufacturer's Data Reports to be attached DISC PUR HASED PER SPEC. D'CCPV802 CN AND P .00746-041-2X Ref. B007224 File:ME-VAL - W ISI la s 2 CERTIFICATE OF COMPLIANCE We certify t.hat t.he statements made in the report are correct and this REPLACEMENT conforms to the rules of the ASME Code,Section XI.

repair or replacement, Type Code Symbol Stamp

~H Expiration Date N A Signed .R.Pisars aint. En . Su ervisor Date 19 'PM Owner or Owner's Designee, Title CERTIFICATE OF INSERVICE INSPECTION I , the undersigned, holding a valid commission issued by the National Board of Boiler by ARKWRIGHT MUTUAL INS. CO;* of NORWOOD MASS ~

have inspected the components described in this Owner's Report during the period 9-rs" t:o and state that to the best of my knowledge and belief, the Owner has performed examination's and taken corrective measures described in this Owner's Report in accordance with the requirements of the ASME Code,Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the examinations and corrective measures described'n this Owner's Report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal in)ury or property damage or a loss of any kind arising from or connected with this inspection.

Commissions Inspec or's Si at e National Board, State, Province, Endorsements

• FACTORY MUTUAL ENGINEERING ASSOCIATION Date 19 SW