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Proposed Tech Specs Reflecting Adminstrative Changes
	ML17328A923
Person / Time
Site:	Cook
Issue date:	02/15/1991
From:	; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	;
Shared Package
ML17328A924	List: ML17328A923; ML17334B393;
References
	NUDOCS 9102220144
	Download: ML17328A923 (140)
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{{#Wiki_filter:ATTACHMENT 2 to AEP:NRC:1137 PROPOSED, REVISED TECHNICAL SPECIFICATIONS PAGES FOR DONALD C. COOK NUCLEAR PLANT UNITS 1 AND 2

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant average temperature (Tavg ) shall not exceed the limits shown in Figure 2.1-1 for 4 loop operation. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. COOK NUCLEAR PLANT - UNIT 1 2-1 AMENDMENT NO. jUP

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658 UNACCEPTABLE ZOO Ps)'a OP ERAT ION 648

                                       ~100 Ps fa
                                      ~000 Ps 1'a le~0

~ 618 Ps)'a ACCEPTABLE OPERATION 578

                 .2             .4        .5         .6       .7      .8    . 1     l. 1.1   1.2 POSER    t frect,ian of         noe<cldl  )

PRESSURE 8REAKPOINTS (PS IA) (FRACTION RATED THERMAL POWER, T-AVG IN DEGREES F) 1840 (0.0, 622.1), (1.13, 587.3), (1.20, 577.5) 2000 (0.0, 633.8), (1.08, 601.4), (1 2o, 586.0). 2100 (0.0, 640.8), (1.06, 609.8), (1.20, 591.3) 2250 (0.0, 650.7), (1.02, 621.9), (1.20, 598.9) 2400 (0.0, 660".1), (0.98, 633.7), (1.20, 606.2) FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS COOK i%'CLEAR PLANT - UNIT I 2-2 AH~HDilENT NO. 7g. 7N rsz

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TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 13.Steam Generator Greater than or equal to 17'%f Greater than or equal to Water Level-Low- narrow range instrument 16% of narrow range Low span - each steam generator instrument span - each steam generator 14.Steam/Feedwater Less than6or equal to Less than or equal to Flow Mismatch and 0.71 x 10 lb/hr of steam 0.73 x 10 lbs/hr of Low Steam Generator flow at RATED THERMAL POWER steam flow at RATED Water Level coincident with steam THERMAL POWER coincident generator water level with steam generator water greater than or equal to 25'%f level greater than or equal narrow range instrument to 24% of narrow range span - each steam generator instrument span - each steam generator 15.Undervoltage Greater than or equal to Greater than or equal to Reactor Coolant 2750 volts - each bus 2725 volts - each bus Pumps 16.Underfrequency- Greater than or equal to Greater than or equal to Reactor Coolant 57.5 -Hz - each bus 57.4 Hz - each bus Pumps 17.Turbine Trip A. Low Fluid Oil Greater than or equal to Greater than or equal to Pressure 800 psig 750 psig B. Turbine Stop Greater than or equal to Greater than or equal to Valve Closure 1% open 1%. open 18.Safety Infection Not Applicable Not Applicable Input from ESF 19.Reactor Coolant Not Applicable Not Applicable Pump Breaker Position Trip COOK NUCLEAR PLANT - UNIT 1 2-6 AMENDMENT NO.

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

6. Steam Flow in Two Steam Lines-Hi h Coincident With Steam Line Pressure-Low

a. Safety In]ection (ECCS) Less than or equal to 13.0¹/23.0¹¹

b. Reactor Trip (from SI) Less than or equal to 3.0

c. Feedwater Isolation Less than or equal to 8 '

d. Containment Isolation-Phase "A" Less than or equal to 18.0¹/28.0¹¹

e. Containment Purge and Exhaust Isolation Not Applicable

f. Auxiliary Feedwater Pumps Not Applicable

g. Essential Service Water System Less than or equal to 14.0¹/48.0¹¹

h. Steam Line Isolation Less than or equal

7. Containment Pressure--Hi h-Hi h r

a. Containment Spray Less than or equal to 45.0

b. Containment Isolation-Phase "B" Not Applicable Ce Steam Line Isolation Less than or equal to 10.0 Containment Air Recirculation Fan Less than or equal to 600.0

8. Steam Generator Water Level--Hi h-Hi h

a. Turbine Trip Less than or equal to 2.5

b. Feedwater Isolation Less than or equal to 11.0

9. Steam Generator Water Level--Low-Low

a. Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0

b. Turbine Driven Auxiliary Feedwater Pumps Less than or equal to 60.0

10. 4160 volt Emer enc Bus Loss of Volta e

a. Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0

11. Loss of Main Feedwater Pum s

a. Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0 12: Reactor Coolant Pum Bus Undervolta e

a. Turbine Driven Auxiliary Feedwater Pumps Less than or equal to 60.0 COOK NUCLEAR PLANT - UNIT 1 3/4 3-29 AMENDMENT NO. APi l/Hi le

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST 4.STEAM LINE ISOLATION a.Manual N.A. N.A. M(1) 1,2,3 b.Automatic Actuation N.A. " N.A. M(2) 1,2,3 Logic c.Containment Press-ure--High-High M(3) 1,2,3 d.Steam Flow in 1,2,3 Two Steam Lines-- High Coincident with Tavg--Low-Low e.Steam Line Pressure--Low S 1,2,3 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 1,2,3 Water Level--Low-Low b.4 kv Bus 1,2,3 Loss of Voltage c.Safety Injection N.A. N.AD M(2) 1,2,3 d.Loss of Main Feed N.A. N.A. 1,2 Pumps COOK NUCLEAR PLANT - UNIT 1 3/4 3-33 AMENDMENT NO. APE, Z/P N2

TABLE 3.3-6 (Continued) TABLE NOTATION ACTION 20 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, comply with the ACTION requirements of Specification 3.4.6.1. ACTION 21 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per day. ACTION 22 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirements, comply with the ACTION requirements of Specification 3.9.9. This ACTION is not required during the performance of containment integrated leak rate test. ACTION 22A- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements:

1. either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3. Technical Specification Sections 3.0.3 and 3.0.4 Not Applicable.

ACTION 22B- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements'. either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a Special Report to the Commission pursuant to Specification 6.9 ' within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3. In the event of an accident involving radiological releases initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours.

4. Technical Specification Sections 3.0.3 and 3.0.4 Not Applicable.

W COOK NUCLEAR PLANT - UNIT 1 3/4 3-37 AMENDMENT NO. W

TABLE 3.3-10 Unit 1 and Common Area Fire Detection S stems Total Number Detector S stem Location of Detectors Heat Flame Smoke (x/y)+ (x/y)" (x/y)" Auxiliary Building Elevation

')

573 23/OC'5/OC b) Elevation'87 c) Elevation 609 41/OC d) Elevation 633 41/OC e) Elevation 650 34/OC f) New Fuel STGE Area 4/OC g) RP Access Control & Chem Labs 25/0 Ul East Main Steam Valve Enclosure 28/Om Ul Main Steam Line Area El. 612 (Around Containment) 13/Oi'm Ul NESW Valve Area El. 612 2/0 Ul 4KV Switchgear (AB) 0/3 0/2 Ul 4KV Switchgear (CD) 0/3 0/2 Ul Engr. Safety System Switchgear & XFMR. Rm. 0/5 0/9 Ul CRD, XFMR. & Switchgear Rm, Inverter & Bttry. Rms. 0/5 0/8 Ul Pressurizer Heater XFMR. Rm. 12/0 Ul Diesel Fuel Oil Transfer Pump Rm. 0/1 Ul Diesel Generator Rm. 1AB 0/2 Ul Diesel Generator Rm. 1CD 0/2 Ul Diesel Generator Ramp Corr. 4/0 Ul&2 AFWP Vestibule 2/OC Ul Control Room 45/0 Ul Switchgear Cable Vault 0/10~ 0/13 Ul Control Room Cable Vault 0/6 5nw'ov Ul Aux. Cable Vault 0/6 U162 ESW Basement Area 4/OC Ul ESW Pump & MCC Rms. 9/0 C System protects area common to both Units 1 and 2 >'<(x/y) x is number of Function A (early warning fire detection and notification only) instruments. y is number of Function B (actuation of fire suppression systems and early warning and notification) instruments'ircuit contains both smoke and flame detectors two circuits of five detectors each two circuits of 32 and 33 detectors each COOK NUCLEAR PLANT - UNIT 1 3/4 3-53 AMENDMENT NO. 79'8P

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TABLE 3.3-11 POST-ACCIDENT MONITORING INTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE 1 ~ Containment Pressure

2. Reactor Coolant Outlet Temperature-( ide Range)

3. Reactor Coolant Inlet Temperature-T (Wide Range) 2

4. Reactor Coolant Pressure-Vide Range 2

5. Pressurizer Water Level 2

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 1 10.Auxiliary Feedwater Flow Rate 1/steam generator~

11.Reactor Coolant System Subcooling Margin Monitor ].%% 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 15 Incore Thermocouples (Core Exit Thermocouples)

     ~                                                                2/Core Quadrant 16.Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)                             One  Train

(~annels/Train) 17.Containment Sump Le~el 18.Containment Water Level 2 Steam Generator Water Level Channels can be used as a substitute for the corresponding auxiliary feedwater flow rate channel instrument.

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PPC 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-Limit Switches instruments. ~'~~The 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. COOK NUCLEAR PLANT - UNIT 1 3/4 3-55 AMENDMENT NO. ggg

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REACTOR COOLANT SYSTEM LIMITING CONDITION. FOR OPERATION (Continued)

2. With two or more block valves inoperable, within 1 hour 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 appropriately above for inoperable PORVs, relating to OPERATIONAL MODE, as

c. With PORVs and block valves not in the same line inoperable,~

within 1 hour 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: At least once per 31 days by performance of CHANNEL FUNCTIONAL TEST, excluding valve operation, and At least once per 18 months by performance of a CHANNEL CALIBRATION. 4.4.11.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days'y operating the valve through one complete cycle of full travel. The block valve(s) do not have to be tested when ACTION 3.4.11.a or 3.4.11 ' 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 4.8.2.3.2.d and 4.8.1.1.2.e. PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this leakage are not considered inoperable. COOK NUCLEAR PLANT - UNIT 1 3/4 4-36 AMENDMENT NO. NH, X2'8,

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment ai'r lock shall be OPERABLE with:

a. Both doors .closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.05 L a

at P , 12 psig ~ MODES a'PPLICABILITY: 1, 2, 3 and 4. ACTION: With an air lock inoperable, restore the air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE UIREMENTS 4.6.1 ~ 3 Each containment air lock shall be demonstrated OPERABLE: a ~ By visual inspection after each opening to verify that the seal has not been damaged.

            *Within 72 hours following each closing, perform an air leakage test without a simulated pressure force on the door by pressurizing the gap between the seals to 12 psig and verifying a seal leakage of no greater than 0.5 L

~'<Exemption to Appendix "J" of,10 CFR 50, COOK NUCLEAR PLANT - UNIT 1 3/4 6-4

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued C. At least once per 6 months, perform an air leakage test without a simulated pressure force on the door per 4.6 '.3.b., then perform an air leakage test with a simulated pressure force on the door, by pressurizing the volume between the seals to 12 psig and verifying a seal leakage of no greater than 0.0005 L . a't

d. least once per 6 months by conducting an overall air lock leakage test at P a

(12 psig) and by verifying that the overall air lock leakage rate is within its limits

e. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

COOK NUCLEAR PLANT - UNIT 1 3/4 6-5 AMENDMENT NO.

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CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 4.6.3.1.2 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: a ~ Verifying that on a Phase A containment isolation test signal, L 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. COOK NUCLEAR PLANT - UNIT 1 3/4 6-15 AMENDMENT NO. gP7,

C' TABLE 3.6-1 Continued ISOLATION TIME VALVE NUMBER FUNCTION IN SECONDS CONTAINMENT PURGE EXHAUST~ Continued

12. VCR-205 COMP. PURGE AIR INLET 14.

VCR-207%'PPER

13. VCR-206 UPPER CONT.

COMP. PRESS PURGE AIR OUTLET RELIEF FAN ISOLATION MANUAL ISOLATION VALVES

1. ICM-111 RHR TO RC COLD LEGS NA

2. ICM-129 RHR INLET TO PUMPS NA

3. ICM-250 BORON INJECTION OUTLET NA

4. ICM-251 BORON INJECTION OUTLET NA

5. ICM-260 SAFETY INJECTION OUTLET NA

6. ICM-265 SAFETY INJECTION OUTLET NA

7. ICM-305 RHR/CTS SUCTION FROM SUMP NA

8. ICM-306 RHR/CTS SUCTION FROM SUMP NA

9. ICM-311 RHR TO RC HOT LEGS NA

10. ICM-321 RHR TO RC HOT LEGS NA

11. NPX 151 VI DEAD WEIGHT TESTER NA

12. PA 343 CONTAINMENT SERVICE AIR NA

13. SF-151 REFUELING WATER SUPPLY NA

14. SF-153 REFUELING WATER SUPPLY NA

15. SF-159 REFUELING CAVITY DRAIN TO NA PURIFICATION SYSTEM

16. SF-160 REFUELING CAVITY DRAIN TO NA PURIFICATION SYSTEM

17. SI-171 SAFETY INJECTION TEST LINE NA

18. SI-172 ACCUMULATOR TEST LINE NA COOK NUCLEAR PLANT UNIT 1 3/4 6-21 AMENDMENT NO.

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PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2

a. At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

1. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and

2. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

b. At least one auxiliary feedwater flowpath in support of Unit 2 shutdown functions shall be available.

APPLICABILITY: Specification 3.7 '.2.a - MODES 1, 2, 3. Specification 3.7.1.2.b - At all times when Unit 2 is in MODES 1, 2, or 3. ACTIONS'hen Specification 3.7.1.2.a is applicable: a ~ With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours'.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. When Specification 3.7.1.2.b is applicable: With no flow path to Unit 2 available, return at least one flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 2 and return at least one flow path to available status within the next 60 days, or have Unit 2 in HOT STANDBY within the next 12 hours and HOT SHUTDOWN within the following 24 hours. The requirements of Specification 3.0.4 are not applicable. COOK NUCLEAR PLANT - UNIT 1 3/4 7-5 AMENDMENT NO. g7, NP,

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REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING+ LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2,500 pounds shall be prohibited from travel over fuel assemblies in the storage pool. Loads carried over the spent fuel pool and the heights at which they may be carried over racks containing fuel shall be limited in such a way as to preclude impact energies over 24,240 in.-lbs., if the loads are dropped from the crane. APPLICABILITY: Vith fuel assemblies in the storage pool. ACTION: Llith the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE UIREMENTS 4.9.7.1 Crane interlocks which prevent crane travel with loads in excess of 2,500 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. 4.9.7.2 The potential impact energy due to dropping the crane's load shall be determined to be less than or equal to 24,240 in.-lbs. prior to moving each load over racks containing fuel.

~'i Shared system   with Cook Nuclear  Plant - Unit 2.

"COOK NUCLEAR PLANT - UNIT 1 3/4 9-8 AMENDMENT NO. NX, ZAP

5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1-1. LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1-2. Site Boundar For Gaseous and Li uid Effluents 5.1.3 The SITE BOUNDARY for gaseous and liquid effluents shall be as shown in Figure 5.1-3. 5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter 115 feet.

b. Nominal inside height 160 feet.+

c. Minimum thickness of concrete walls 3'6".

d. Minimum thickness of concrete roof 2'6".

e. Minimum thickness of concrete floor pad 10 feet.

Nominal thickness of steel liner, side and dome 3/8 inches.

g. Nominal thickness of steel liner, bottom 1/4 inch.

6

h. Net free volume 1.24 x 10 cubic feet.

From grade (Elev. 608') to inside of dome. COOK NUCLEAR PLANT - UNIT 1 5-1 AMENDMENT NO, H9

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Docket No. 316 Page 5 of 11 (1) Deleted by Amendment 63. (m) Deleted by Amendment 19 Deleted by Amendment 28.

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(o) Fire Protection Amendment The licensee may proceed with and is required to No. 12 complete the modifications identified in Table 1 of the Fire Protection Safety Evaluation Report for the Donald C. Cook Nuclear Plant dated June 4, 1979 'hese modifications shall be completed in accordance with the dates contained in Table 1 of that SER or Supplements thereto. Administrative controls for fire protection as described in the licensee's submittals dated January 31, 1977 and October 27, 1977 shall be implemented and maintained. Amendment (p) Deleted by Amendment 121 No. 64, 121

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DEFINITIONS SOLIDIFICATION 1.29 SOLIDIFICATION shall be the conversion of radioactive liquid, resin and sludge wastes from liquid systems into a form that meets shipping and burial site requirements. OFFSITE DOSE CALCULATION MANUAL ODCM 1.30 The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints and the conduct of environmental radiological monitoring program. GASEOUS RADWASTE TREATMENT SYSTEM 1.31 A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. VENTILATION EXHAUST TREATMENT SYSTEM , 1 32

   ~         A VENTILATION EXHAUST TREATMENT SYSTEM      is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate. form in effluents by passing ventilation or vent exhaust gases through charcoal absorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components'URGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging              air or gas from a confinement to maintain temperature, pressure, humidi.ty, concentration or other operating condition, in such a manner that replacement           air or gas is required to purify the confinement.

VENTING 1.34 VENTING is the controlled process of discharging air or gas from,a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a VENTING process. COOK NUCLEAR PLANT - UNIT 2 1-7 AMENDMENT NO.

Pl r, 0 t'.f 4 P.

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and the highest operating loop coolant average temperature (Tavg ) shall not exceed the limits shown in Figure 2.1-1 for 4 loop operation. APPLICABILITY: MODES 1 and 2. ACTION: Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour. REACTOR COOLANT SYSTEM PRESSURE 2' ~ 2 The Reactor Coolant System pressure shall not exceed 2735 psig. APPLICABILITY: MODES 1, 2, 3, 4 and 5. ACTION: MODES 1 and 2 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limit within 1 hour. MODES 3, 4 and 5 Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes. COOK NUCLEAR PLANT - UNIT 2 2-1 AMENDMENT NO. N

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3.4.2 POWER DISTRIBUTION LIMITS AXIAL FLUX DIFFERENCE AFD LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the target'band about a targe flux difference. The target band is specified in the COLR. APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER+ ACTION:

a. With the indicated AXIAL FLUX DIFFERENCE outside of the target band about the target flux difference and with THERMAL POWER:

l. Above 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER, within 15 minutes:

a) Either restore the indicated AFD to within the target band limits, or b) Reduce THERMAL POWER to less than 90'%r 0.9 x APL (whichever is less) of RATED THERMAL POWER.

2. Between 50% and 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER; a) POWER OPERATION may continue provided:

1) The indicated AFD has not been outside of the target band for more than 1 hour penalty deviation cumulative during the previous 24 hours, and

2) The indicated AFD is within the limits specified in the COLE. Otherwise, reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip Setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours.

b) Surveillance testing of the Power Range'eutron Flux Channels may be performed pursuant to Specification 4.3.1.1.1 provided the indicated AFD is maintained within the limit specified in the COLR. A total of 16 hours operation may be accumulated with the AFD outside of the target band during this testing without penalty deviation. See Special Test Exception 3 '0.2 COOK NUCLEAR PLANT - UNIT 2 3/4 2-1 AMENDMENT NO. Ag, gg7, Xg/

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TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION TOTAL NO. OF MINIMUM CHAN- CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT NELS TO TRIP OPERABLE MODES ACTION

1. Manual Reactor Trip 1,2 and + 12

2. Power Range, Neutron 1, 2 and

2¹ Flux

3. Power Range, Neutron 1 2 2¹ Flux, High Positive Rate

4. Power Range, Neutron 1, 2 2¹ Flux, High Negative Rate

5. Intermediate Range, 1, 2 and + 3 Neutron Flux

6. Source Range, Neutron Flux A. Startup 2¹¹ and + 4 B. Shutdown 3, 4 and 5 5

7. Overtemperature Delta T Four Loop Operation 1, 2 6¹

8. Overpower Delta T Four. Loop Operation 1, 2 6¹ COOK NUCLEAR PLANT - UNIT 2 3/4 3-2 AMENDMENT NO. N

~ I )r3 0 C~

    ~ ".Iy C

TABLE 3.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TOTAL NO. OF MINIMUM CHAN- CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT NELS TO TRI OPERABLE MODES ACTION

9. Pressurizer Pressure Low 1, 2 6¹

'10.Pressurizer Pressure

    - High                                                           1, 2          6¹ 11.Pressurizer Water                                                 1, 2          7¹ Level -- High 12.Loss of Flow - Single Loop (Above P-8)        3/loop     2/loop in      2/loop in       1 any opera-     each operating ting loop      loop 13.Loss of Flow - Two      3/loop     2/loop in      2/loop in                     7¹ Loops (Above P-7                   two opera-     each opera-and below P-8)                    ting loops     ting loop 14.Steam Generator         3/loop     2/loop in      2/loop          1, 2          7¹ Water Level-Low-Low                any opera-     each operating ting loop      loop 15.Steam/Feedwater Flow 2/loop        1/loop         1/loop-level    1, 2          7¹ Mismatch and Low        level      level          and Steam Generator        and        coincident     2/loop-flow Water                   2/loop-    with          mismatch or flow       1/loop-        2/loop-level mismatch   flow mis-      and in  same   match in       1/loop-flow loop       same  loop    mismatch COOK NUCLEAR PLANT   - UNIT 2                 3/4 3-3             AMENDMENT NO. Ni lP7

t>> E

TABLE 3.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TOTAL NO. OF MINIMUM CHAN- CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT NELS TO TRI OPERABLE MODES ACTION 16.Undervoltage-Reactor 4-1/bus 2 6¹ Coolant Pumps 17.Underfrequency- 4-1/bus 6¹ Reactor Coolant Pumps 18.Turbine Trip; A.Low Fluid Oil 7¹ Pressure B.Turbine Stop Valve 6¹ Closure 19.Safety Injection 2 1, 2 Input from ESF 20.Reactor Coolant Pump Breaker Position Trip Above P-7 1/ 1/breaker breaker per operating loop 21.Reactor Trip 1,2, 1,13, Breakers 3%,4%,5i'r 14 22.Automatic Trip Logic 1,2, 1 3>'r, 4>'c, 5~'< 14 COOK NUCLEAR PLANT - UNIT 2 3/4 3-4 AMENDMENT NO. PP, gPj, ggj

4 fi gt

TABLE 3.3-6 (Continued) TABLE NOTATION'CTION 20 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, comply with the ACTION requirements of Specification 3.4.6 '. ACTION 21 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement, perform area surveys of the monitored area with portable monitoring instrumentation at least once per day. ACTION 22 - With the number of channels OPERABLE less than required by the Minimum Channels Operable requirement," comply with the ACTION requirements of Specification 3.9.9. This ACTION is not required during the performance of containment integrated leak rate test. ACTION 22A- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements: 1 ~ either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3. Technical Specification Sections 3.0.3 and 3.0.4 Not Applicable.

ACTION 22B- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements.

l. either restore the inoperable Channel(s) to OPERABLE status within 7 days of the event, or

2. prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and'he plans and schedule for restoring the system to OPERABLE status.

3. In the event of an accident involving radiological releases initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours.

4. Technical Specification Sections 3.0.3 and 3.0.4 Not Applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-36 AMENDMENT NO W,

v J C

TABLE 3.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE

1. Pressure 2

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

3. 'ontainment Reactor Coolant Inlet Temperature - T (Wide Range) 2

4. Reactor Coolant Pressure - Wide Range 2

5. Pressurizer Water Level 2

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 1

10. Auxiliary Feedwater Flow Rate 1/steam generator*

11. Reactor Coolant System Subcooling Margin Monitor ] %%'

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

15. Incore Thermocouples (Core Exit Thermocouples) 2/core quadrant

16. Reactor Coolant Inventory Tracking System one-train(3 channels/train) 17.

18 (Reactor Vessel Level Indication) Containment Sump Level Containment Water Level

]~

2%9hlPk Steam Generator Water Level Channels can be used as a substitute for the corresponding auxiliary feedwater flow rate channel instrument. PPC 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-Limit Switches instruments'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. COOK NUCLEAR PLANT - UNIT 2 3/4 3-46 AMENDMENT NO. Pg, gt

     )fan t

I,

  "~  l'%

TABLE 3.3-11 Unit 2 and Common Area Fire Detection S stems Total Number Detection S stem Location of Detectors Heat Flame Smoke (x/y)* (x/y)* (x/y)% Auxiliary Building a) Elevation 573 23/OC b) Elevation 587 55/OC c) Elevation 609 41/OC d) Elevation 633 41/OC e) Elevation 650 34/OC f) New Fuel STGE Area 4/OC U2 East Main Steam Valve Enclosure 28/P~ U2 Main Steam Line Area El. 612 (Around Containment) ] 3/OMY U2 NESW Valve Area El. 612 2/0 U2 4KV Switchgear (AB) 0/3 0/2 U2 4KV Switchgear (CD) 0/3 0/2 U2 Engr. Safety System Switchgear & XFMR. Rm. 0/5 0/14 U2 CRD, XFMR & Switchgear Rm. Inverter & AB Bttry. Rms. 0/5 0/17 U2 Pressurizer Heater XFMR Rm. ~ 12/0 U2 Diesel Fuel Oil XFMR. Rm. 0/1 U2 Diesel Generator Rm. 2AB 0/2 U2 Diesel Generator Rm. 2CD 0/2 U2 Diesel Generator Ramp Corr. 4/0 U1&2 AFWP Vestibule 2/OC U2 Control Room 42/0 U2 Switchgear Cable Vault 0/] PM'<O'/13 U2 Control Rm. Cable Vault 0/76~'nw U2 Aux. Cable Vault 0/6 Ul&2 ESW Basement Area 4/OC U2 ESW Pump & MCC Rms. 9/0 C System protects area common to both Units 1 and 2 +(x/y) is number of Function A (early warning fire detection and x notification only) instruments. y is, number of Function B (actuation of fire suppression systems and early warning and notifi.cation) instruments. circuit contains both smoke and flame detectors two circuits of five detectors each two circuits of 38 detectors each COOK NUCLEAR PLANT - UNIT 2 3/4 3-52 AMENDMENT NO. N, X2$

   <<ji 0

ir VF 0

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued

2. With two or more block ~alves inoperable, Within 1 hour 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'f ACTION a.2 or a 3 above for inoperable

PORVs, relating to OPERATIONAL MODE, as appropriate. C. PORVs 'and block valves not in the same line inoperable,* k'ith within 1 hour 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.

4 4 11.2 Each of the three block valves shall be demonstrated OPERABLE at least full

~ ~

once per 92 days by operating the valve through one complete cycle of travel. The block valve(s) do not have to be tested'hen ACTION 3.4 'l.a or 3.4.11.c is applied. 4.F 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.

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

COOK NUCLEAR PLANT - UNIT 2 3/4 4-33 AMENDMENT NO. Ni 97( Z8Z

1~

~ I

>h 1 P

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE: a ~ At least once per 12 hours by verifying that the following valves are in the indicated positions with the control power locked out: Valve Number Valve Function Valve Position

a. IMO-390 a. RVST to RHR a. Open

b. IMO-315 b. Low head SI b. Closed to Hot Leg

c. IMO-325 c. Low head SI c. Closed to Hot Leg

d. IMP-262+ d, Mini flow line d. Open

e. IMO-263* e. Mini flow line e. Open f . IMO-261* f. SI Suction f. Open

g. ICM-305+ g. Sump Line g. Closed

h. ICM-306+ h. Sump Line h. Closed

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

ce By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

These valves must change position during the switchover from injection to recirculation flow following LOCA.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-4 AMENDMENT NO. 7H,

f

 Ã
 )<

t ( , ~

CONTAINMENT SYSTEMS CONTAINMENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containment air lock shall be OPERABLE with:

a. Both doors closed except when the air lock is being used for normal transit entry and exit through the containment, then at least one air lock door shall be closed, and

b. An overall air lock leakage rate of less than or equal to 0.05 L a

at P , 12 psig. MODES a'PPLICABILITY: 1, 2, 3 and 4 ~ ACTION'ith an air lock inoperable, maintain at least one door closed; restore the air lock to OPERABLE status within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE RE UIREMENTS 4.6 '.3 Each containment air lock shall be demonstrated OPERABLE:

a. +After each opening, except when the air lock is being used for multiple entries, then at least once per 72 hours, by performing an air leakage test without a simulated pressure force on the door by pressurizing the volume between the door seals to 12 psig and verifying a seal leakage rate of no greater than 0.5 L

b. +Within 72 hours following each closing,,perform an air leakage test without a simulated pressure force on the door per Specification 4.6 '.3.a.; then by performing an air leakage with a simulated pressure force on the door by pressurizing the volume between the door seals to 12 psig and verifying a seal leakage rate of no greater than 0.0005 La ~

+Exemption to Appendix "J" of 10 CFR 50. COOK NUCLEAR PLANT - UNIT 2 3/4 6-4 AMENDMENT NO.

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS Continued ce At least once per 6 months by conducting an overall air lock leakage test at P a (12 psig) and by verifying that the'" overall air lock leakage rate is within its limit.

d. At least once per 6 months by verifying that only one door in each air lock can be opened at a time.

COOK NUCLEAR PLANT - UNIT 2 3/4 6-5 AMENDMENT NO.

Cl t I

TABLE 3.6-1 (Cont'd) CONTAINMENT ISOLATION VALVES ISOLATION TIME VALVE NUMBER FUNCTION IN SECONDS D. MANUAL ISOLATION VALVES (1) (Cont'd)

3. ICM-250 BORON INJECTION OUTLET NA

4. ICM-251 BORON INJECTION OUTLET NA

5. ICM-260 SAFETY INJECTION OUTLET NA

6. ICM-265 SAFETY INJECTION OUTLET NA

7. ICM-305 RHR/CTS SUCTION FROM SUMP NA

8. ICM-306 RHR/CTS SUCTION FROM SUMP NA

9. ICM-311¹ RHR TO RC HOT LEGS NA

10. ICM-321¹ RHR TO RC HOT LEGS NA E. OTHER

1. CS-442-1 SEAL WTR. TO RCP ¹1 NA

2. CS-442-2 SEAL WTR. TO RCP ¹2 NA

3. CS-442-3 SEAL WTR. TO RCP ¹3 NA

4. CS-442-4 SEAL WTR. TO RCP ¹4 NA COOK NUCLEAR PLANT - UNIT 2 3/4 6-27 AMENDMENT NO.

>,E ~4

PLANT SYSTEMS AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7,1.2

a. At least three independent steam generator auxiliary feedwater pumps and associated flow paths shall be OPERABLE with:

1. Two motor-driven auxiliary feedwater pumps, each capable of being powered from separate emergency busses, and

2. One steam turbine-driven auxiliary feedwater pump capable of being powered from an OPERABLE steam supply system.

b. At least one auxiliary feedwater flow path in support of Unit 1 shutdown function shall be available.

APPLICABILITY: Specification 3.7.1 2.a - MODES 1, 2, 3.

Specification 3.7.1.2.b - At all times when Unit 1 is in MODES 1, 2, or 3. ACTIONS'hen Specification 3.7.1.2.a is applicable:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours.

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT STANDBY within the following 6 hours.

c. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible.

When bourse'he Specification 3.7.1.2.b is applicable: With no flow path to Unit 1 available, return at least one flow path 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 hours and HOT SHUTDOWN within the following 24 requirements of Specification 3.0.4 are not applicable, COOK NUCLEAR PLANT - UNIT 2 3/4 7-5 AMENDMENT NO. N

4 T

ELECTRICAL POWER SYSTEMS 3 4.8.3 Alternative A.C. Power Sources LIMITING CONDITION FOR OPERATION 3.8.3.1 The steady state bus voltage for the manual alternate reserve source~ shall be greater than or equal to 90% of the nominal bus voltage. APPLICABILITY: Whenever the manual alternate reserve source (69 kV) is connected to more than two buses. ACTION: With bus voltage less than 90% nominal, adjust load on the remaining buses to maintain steady state bus voltage greater than or equal to 90'4 limit. SURVEILLANCE RE UIREMENTS 4.8.3.1 No additional surveillance requirements other than those required by Specifications 4.8.1.1.1 and 4.8.1.2. >>Shared with Cook Nuclear Plant Unit 1, COOK NUCLEAR PLANT - UNIT 2 3/4 8-20 AMENDMENT NO. Xfg

0 IA I I ~ 4 4 ig ~l, Qb F 4g

REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING* LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2,500 pounds shall be prohibited from travel over fuel assemblies in the storage pool. Loads carried over the spent fuel pool and the heights at which they may be carried over racks containing fuel shall be limited in such a way as to preclude impact energies over 24,240 in.-lbs., if the loads are dropped from the crane. APPLICABILITY: With fuel assemblies in the storage pool. ACTION: With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE UIREMENTS 4.9.7.1 Crane interlocks which prevent crane travel with loads in excess of 2,500 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. 4.9.7.2 The potential impact energy due to dropping the crane's load shall be determined to be less than or equal to 24,240 in.-lbs. prior to moving each load over racks containing fuel.

Shared system with Cook Nuclear Plant - Unit l.

COOK NUCLEAR PLANT - UNIT 2 3/4 9-7 AMENDMENT NO. gg gg

fE '4

REFUELING OPERATIONS CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust isolation system shall be OPERABLE. APPLICABILITY: During Core Alterations or movement of irradiated fuel within the containment. ACTION: With the Containment Purge and Exhaust isolation system inoperable, close each of the Purge and Exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE RE UIREMENTS 4.9.9 The Containment Purge and Exhaust isolation system shall be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment Purge and Exhaust isolation occurs on manual initiation and on a high radiation test signal from each of the containment radiation monitoring instrumentation channels. COOK NUCLEAR PLANT - UNIT 2 3/4 9-9 AMENDMENT NO. lN

4 I'4, 5.0 DESIGN FEATURES 5.1 SITE Exclusion Area 5.1.1 The exclus'ion area shall be as shown in Figure 5.1-1. Low Po ulation Zone 5.1.2 The low population zone shall be as shown in. Figure 5.1-2. Site Boundar For Gaseous and Li uid Effluents 5.1 '3

   ~    The SITE BOUNDARY    for gaseous   and liquid effluents shall    be as shown in Figure 5.1-3.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The reactor containment building is a steel lined, reinforced concrete building of cylindrical shape, with a dome roof and having the following design features:

a. Nominal inside diameter 115 feet.

b, Nominal inside height 160 feet.

c. Minimum thickness of concrete walls 3'6".

d. Minimum thickness of concrete roof 2'6".

e. Minimum thickness of concrete floor pad 10 feet.

f, Nominal thickness of steel liner '3/8 inches.

g. Net free volume 1.24 x 10 6 cubic feet.

DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained in accordance with the original design provisions contained in Section 5.2.2 of the FSAR. COOK NUCLEAR PLANT - UNIT 2 5-1 AMENDMENT NO ~ Pl

ATTACHMENT 3 TO AEP:NRC:1137 EXISTING T/S PAGES MARKED TO REFLECT PROPOSED CHANGES

Of ~f

rave p mvmps'g The combination of a pOQER, pressurizer pressure ance rbe hXghesc opera@'ng loop coolanc emperature (T ave ) shall noc exceecL che 1&its shovn in Pipce 2.1-1 for 4 loop operas'5n. Q~OV: "henevcr rhe po'.'n" ref'.".ed bv rhe combinacion o j -he highesc ooe ac'..g average cempera"'e and .:":"=R~AL PO4KR has exceeCecf "he appropra"e pressuri er pressu e 1'ne. "e in HOT STANDBY ~ichin 1 hour. OT lpga 5 l5 5'0 c 5'.+ 2.'.2 ...e Beaccor Coo'.an" 5'rscem "ressure sha' noc exceed 2735 "s >

                            .COLS     r., 2.      '       "8   5 JP>

vv

    '"cene:er -he 2.eac=o                     4c a.. 5 vs rem p essu        .as excee~ed     . i%- s '-r.

in ';.O S.A'iD3v ":".e ?.eaccor Coolan= Sys-. h AQgssu a ~ ~

                     ~   ~~     ~  oL
    'ene"er         ='.".e   ?eacror oo an= Sysrem "ressure                   .-.-s ex" ceca"               L ~

red ce =';.e B.eaccor Coolan" Syscem pressure

       ..u~es   .

D. C. COOK - WIT 1 2-1 ~mme SO. 120

                                                     ~4~O                    UHACCE?TABLE Ps l g            OPERAi'IOH SO Ps /~

lOo Ps) oooo Ps 1g P 4 4O ala Ps p'~ see ACCEPTABLE OP c RAT I GH 578

                             ,5       .4      .5        .6        .7   .8,    . l    l. l.l   I.Q POUTER   lf'rcgtlOh 0( hO+>>Ill
     ,PRESSURE                                   BREAKPO IHTS (PS IA)       (FRACTION RATED THERMAL POMER, T-.'VG IH DEGREES F) 1840                   (0.0,  622.1),     (1.13,     587.3). (1.20, 577.5) 2000                   (0.0,  633.8),     (1.08,    601.4).  (1.20, 586.0) 2100                   (0.0,  640.8),     (1.06,    609.8).  (1.20, 591.3) 2250                   (0.0,  650.7),     (1.02,    621.9),  (1.20, 598.9) 2400                   (0.0,  660.1),     (0.98,     633.7), (1.20, 606  ')
                          =REHLTHERMAL-POMER- ~.=34QPAR=

FIGURE 2.1-1 REACTOR CORE SAFETY LIMITS COOK R:CL~c PLAVZ - UNIT 1 2-2 az~vDs~vr yo. 7g,,fgp 152

IIIIII< 2.~2-I ~Cunt Inured HEAC1DR TRIP SYSTEH INSTRtNENTAT IOH TRIP SETPOIIITS FUHCT IAHAL UNIT TRIP SETPOltlT Al LOMABLE VALUES firn l3. Stcaa Generator Hater >- llX of narra rapgc Instr~nt > 16'l of narrott ringo Inttrt&4nt', level - Lo~-Lo~ span - each stcam generator span - each steae generator b &

14. Steam/Feed~atcr Floe ~ 0.71 x lO lb/hr of stean fin>f < 0.73 x lO lbs/hr of steae Hlseatch and to~ Steaa at. RAIL.O TIIEfNAL f'OMEH coincident at RATED TIIERHAL POHER coincident Generator Mater Level arith steafw generator water level with steaa generator sratcr level

                                              > 251 of narc m range l<<stru-              > 24K of narra range     lnstru-eent span - eacli steam generator         ient span - each stela gcncrator l5. Undcrvoltage     - Reactor                    2750   volts -     each bus             >  2725   volts - each bus Coolant  Pumps

16. Underfrequency - Reactor > 57.5 Ill - each bus > 57.4 Ilz - each bus Coolant Puaps l7. Turbine 1rip aoo pslg > 750 pslg memmr lu~F/oid01 ~ < ~0~ > 11 open

     $. Turbine Stop Valve                       )X OPtft Closure

18. Safety injection Inliut llot hppl lcablc Ifot hppl lcablc froa ESf lg. Reactor Coolant Puwp Hot hppl lcablc Hot Appl icablc Sreaker Pos I t lon Tr lp Q~~'i'm p atd WP~

Q<~

C

~ f

~'4

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

6. Steam Flov in Tvo Steam Lines'-Hi h Coincident Vith Stcam Linc Pressure-Lov

a. Safety Injection (ECCS) < 13.08/23.0~

b. Reactor Trip (from SI) < 3.0 C ~ Feedvater Isolation < 8.0

d. Containment Isolation-Phase "h" < 18.08/28.0~

c ~ Containment Purge and Exhaust Isolation Not hpplicable

f. Auxiliary Fecdvater Pumps Not Applicable g ~ Essential Service Vater System < 14.0e/48.0e>>

Steam Line Isolation < 11.0

7. Containment Pressure--Hi h-Hi h

a. Containment Spray < 45.0

b. Containment Isolation-Phase "B" Not Applicable Linc Isolation < 10.0 C

d.

     ~   Steam Containment  hir Recirculation    Fan           < see~    b~

8. Steam Generator Water Level--Hi h-Hi h

a. Turbine Trip < 2.5

b. Fecdvater Isolation < 11.0

9. Steam Generator Vater Level--Lov-Lov

a. Motor Driven Auxiliary Feedvater Pumps 60.0

b. Turbine Driven Auxiliary Fcedvatcr Pumps' < 60.0

10. 4160 volt Emer enc Bus Loss of Volta e

a. Motor Driven Auxiliary Feedvater Pumps < 60.0

11. Loss of Main Feedvater Pum s

a. Motor Driven Auxiliary Feedvater Pumps < 60.0

12. Reactor Coolant Pum Bus Undervolta e

a. Turbine Driven Auxiliary Feedvater Pumps < 60.0 COOK NUCLEAR PLANT - UNIT 1 3/4 3-29 AMENDMENT NO. gg'ag

n

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS I CHANNEL MODE IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT " CHECK CALIBRATION TEST 4.STEAM LINE ISOLATION a.Manual N.A. N.A. M(1) 1,2,3 b.Automatic Actuation N.A. N.A. M(2) 1,2,3 Logic c.Containment Press-ure--High-High M(3) 1,2,3 d.Steam Flow in S 1,2,3

        'wo   Steam Lines--

High Coincident with Tavg--Low-Low

         ~~~TRIPLqn~

TURBINE sure-Lum S

                    & FEEDWATER ISOLATION a.Steam Generator                                                1,2,3 Water Level--High-High 6.MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.Steam Generator                                                1,2,3 Water Level--Low-Low b.4 kv    Bus                                                    1,2,3 Loss   of Voltage c.Safety Injection            N.A.         N.A.         M(2)     1,2,3 d.Loss of Main Feed           N.A.         N.A.                   1,2 Pumps COOK NUCLEAR PLANT -                      3/4 3-33           AHENDMEvT No.         129, 121,

'I UNIT 1 f f],

ft

 ~I 'I rt I

                                      ~ gv V  <<3    3  +  (Cow<<i aeied)

~ CQTQwj 1Q 'i~'n'mr "h "he -u."be of char'eis OP"=HML"- Charne's Operable re~uiremerc. less chan requ'red oy "he comp y with "he AC. O. z rec " remencs of Specificac'on 3.4.5.1. ACT:ON 21. <<ich che n ..ber of charnels OPKPJBL= "less cnan required by che minimum Channe's Operable requi emenc, perform area su obeys or che monitored area with porcab1e mon'cor ..g instr zencacion at

               'east once per dav.

AC i.ON 22 'ich "he number of channe's OP"=ML"= less chan requ'red bv che minimum Channe 1 s Operable equ'remen=s, comply with che AC; ON

               -ecu "o'ne" s      o- Spec'='cat'on 3.9.9. ;h's AC.TON is not requ red dur ng che pe.=or.-.,ance         of concainmenc 'ncegraced       eak ace    test.

AC, 2 'i=h che ".. mber of O?""RABL"- Channels less chan required by che

              'Ainimum Channels           0?".RA3L". requiremencs:

either restore che inooerab'e Channel(s) co OPKRABLi stacus within 7 davs of che evenc, or prepare and submit a Specia 1'epo r" "o che Commis s ion pursuanc co Soec'ficacion 'o .9.2 w;chin '4 days Eo1.lowing

                       ..e ever.c ou '.'..z che ac ;on =='::en, che cause of che inoper'abii'='nd;he plars and --hedule fo. rescor ng che s'stem =o O.:-BAB'= scat s.
                                                                           ~p,h.

3. .echnica Spec'='cac'on Sec ='crs 3.0.3 ~3.0.4 Noc Aoolicabie.

2"3 <<ich t'h e number of OP="~~BL:- Chanre1.s 'ess chan rea ired b" che

              .".'n'mum Channels          OPKRA3L:- requi emencs.

i. either restore the inoperable Channel(s) co OPKRABLZ status within 7 days of che evenc, or

2. prepare and submi" a Special Report co che Commission pursuanc co Specif'cac'on 6.9.2 with'n 14 days foo~ing che event outl'ning the action taken. che cause of che inoperability and che plans and schedule for restoring che sy'stem co OP"-BABE status.

ln che evenc of an acciden" irvoiv rg radioiog cal releases inic ace che preplanned alternate method of monitoring c..e appropriate paramecer(s) wichin 72 hours. No c App 1. i cable . D. C. COOK - L".lid 1 3/4 3-37 Amendment No. 94.134

k r, P f 'I N sw l~

TABLE 3.3-10 Unit 1 and Common Area Fire Detection S stems Total Number Detector S stem Locatio'n of Detectors Haae Plans Saa'ka (x/7>* (x/7>" <*/7>* Auxiliary Building a) Elevation 38%c ~~K 23/OC 55/OC b) Elevation 587 c) Elevation 609 41/OC d) Elevation 633 41/OC e) Elevation 650 34/OC f) Nev Fuel STGE Area 4/OC g) RP Access Control & Chem Labs 25/0 Ul East Main Steam Valve Enclosure 28/0~ Ul Main Steam Line Area El. 612 (Around Containment) ],3/0~ Ul NESTS Valve Area El. 612 2/0 Ul 4KV Svitchgear (AB) 0/3 0/2 Ul 4KV Svitchgear (CD) 0/3 0/2 Ul Engr. Safety System Svitchgeaz & XFMR. Rm. 0/5 0/9 Ul CRD, XFMR. & Svitchgear Rm. Invartar & Bttry'. Rms. 0/5 0/8 Ul Pressure.ter Heater XFMR. Rm. 12/0 Ul Diesel Fuel Oil Transfer Pump Rm. 0/1 Ul Diesel Generator Rm. 1AB 0/2 Ul Diesel Generator Rm. 1CD 0/2 Ul Diesel Generator Ramp Corr. 4/0 U162 hEVP Vestibule 2/OC Ul Control Room 45/0 Ul Svitchgear Cable Vault 0/10~ 0/13 Ul Ul Control Room Cable Vault hux. Cable Vault 0/6 0/6 5~ Ul&2 ESV Basement hzea 4/OC Ul ESV Pump & MCC Rms. 9/0 C System protects area common to both Units 1 and 2

 *(x/y) x is     number of Functi.on h (early varning fi.re detecti.on and notification only) instruments.

y is number of Function B (actuation of fire suppression systems and early varning and notification) instruments. ( ~ circuit contains both smoke and flame detectors tvo circuits of five detectors each tvo circuits of 32 and 33 detectors each COOK NUCLEAR PLANT - UNIT 1 3/4 3-53 AMENDMENT NO. 79,>3o

  .I 4

1 l-i

ThBL . 3-11 POST-ACCIDENT HONITORIHC INSTRUHENThTION n INSTBUHEIIY HIHIHUH CllhHNELS OPERhhIZ n

1. Contalnacnt Pressure R actor Coo l nt O tlat T peraturo - T T

(Mlde Rang M

3. Rosctox'oolant Inlet Teaperature -

TCO~ (Mide Range)

~ j        Reactor Coolant Pressuro       - Mide Range

5. Pressurizer Mater Level

    &. Stoaa Linc Prcssure                                                               2/Steaa Cencrator

l. St.cua C<<nuxuxor Muxer Level - Harrow Range 1/Stcaa Cencrator 8, kcl>><<11>>g Muti:r Storage Tank Mater Level Bur li: held Tunk Solution Level lo

10. huxt1lary Fceduater Flou Rate 1/Steaa Generator*

11. Reactor Coolant Systca Subcoollng Hargin Honitor 1*a

12. PORV Position Indicator - Liait Sultches*** 1/Valve

13. PORV Block Valve Position Indicator - Llalt Svitches 1/Valve

14. Safety Valve Position Indicator - hcoustic Honitor 1/Va lve

15. Incore Theraocouples (Cote Exit Theraocouples) 2/Core Quadrant 1&. Reactor Coolant Inventory Tracking Systea One Train (3 channels/Train),

(Reactor Vassal Level Indication) 17.. Containaent Swap Level 1**** 1B. Contatnaent Mater Level 2**** Stcaa Ccnerator Mater Level Channels can be used as a substitute for the corresponding auxiliary feed+ster flou ate channel lnstruacnt Ak subcoollng aargln readout can be used as a<substitute for the subcoollng aonltor instruacnt.

    *<+ hcoustlc aonitox'lng of PORV position (1 channel per throe valvos - headered discharge) can be used as s

substitute for the PORV Indicator - Llalt Sultches instruacnts.

    ~~** The x'cqulrcaents fox these tnstruacnts will becoae offectivo after the level tx'ansaltters are aodl fled or rcplnccd und becoac operational. The schedule for aodlf ication or rcplaceaent of the transalttcrs ls desex lbcd ln   xhc Bases.

I.1 2(.&feet-}ve-be fore =s tart-xjp-fxxltoW~mefuetf~xxtxx9~r

tf N ~ 4 aH E I qf

)I

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued

2. With two or more block valves inoperable, within 1 hour 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 hour 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.

E

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILIANCE RE UIREMENTS 4.4.11.1. Each of the three PORVs shall be demonstrated OPERABLE: 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.

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 3.4.11.a or 3.4.11.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 con]unction with the requirements of Specifications

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

COOK NUCLEAR PLANT - UNIT 1 3/4 4-36 AMENDMENT NO. )gg,)PP,144

YC C' A

CONTAINMENT SYSTPIS CONTAjNMENT AiR LOCKS LIMITING CONOlTION FGR OPERATiON 3.6.1.3 Each containment air lock shall, be OPERABLE with:

a. Both doors closed exc pt when the air lock is being used for normal transit entry and exit through the containment, then a least one air lock door shall be closed, and

b. An overall air lock I eakage rate of < 0.05 L at P, 12 psig.

APPLICABiL'i(: MODES I, 2, 3 and 4. ACTION: With an air lock inooerable, restore the afr lock to OPERABLE s.atus within 24 hours or be in at least HOT STANDBY within the next 6 hours and in COLD SAUTOGWN within the fol lowing 30 hours. SURVE'LLANCE QEGUIR~>ENTS 4.6.1.3 Each containment air lock shall be denonstrated OPERABLE: 9y visual inspection af. r each ooening to verify that the seal has not been damaged.

b. g j II wl bout a simulated pressure for=e on the door by pressurizing t.".e gap betwe n the seals to IZ psig and verifying a seal leakage o7 no great r -han 0.: L .

"~emp;cn co poendix "~" o= 10 C.=. O. C. COOK-UNIT I 3/4 6-4

u' CONTAINMENT SYS~S SVRVEILLAHCE REQUIRB1ENTS (Continued) C. At least once per 6 months, perform an air leakage test wi hout a simulated oressure force on -.he door oer 4.6.1.3.b.,

d. At leas leakage air a'1<~~

then perform an air leakage tes with a simulated oressure force on the door, by pressurizing the 0.0005 L . test at P between the seals to 12 psig and verifying a seal leakage of no greater than once oer 6 months by conducting an overall air lock (12 osiq) and by verifying that the overall lock 1 eakage kate is within i ts 1 imi t.

e. A leas. once,per 6 months by verifying that only one door in each air lock can be opened at a time.

0. C. COOK-VNIT 1 3/4 6-i

k 1" f,i q II" li 4p'

CONTAINMENT SYSTEMS SURVEILLANCE REQUIREMENTS Continued va its associated actuator, control or power circuit by o e-cycJ.i st, above, and verification of -'erformance isolation 4.6.3.1.2 Each isolation valve specified in Table 3.6-1 shall be demon-strated OPERABLE during the COLD SHUTDOWN or REHJELING MODE at least once per 18 months by:

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.

                    ~ ~  ~

I COOK NUCLEAR PLANT -, UNIT 1 3/4 6-15 AMENDMENT NO.)gj,],gg

O TABLE 3. 6-1 Conti nu<<~d n n 0 0 ISOl&TION TIHE I VALVE NUHBER FUNCTION 1N SFCONDS cr C. CONTAINHENT PURGE EXllAUST Continued **

12. VCR-205 Upper Comp. Purge Air Inlet

13. VCR-206 Upper Comp. Purge Air Outlet

14. VCR-207* Cont. Press Relief Fan Isolation HANUAL ISOIATION VALVES ICH-ill ICH-129 ICH-250 ICH-251 RIIR RIIR to RC Inlet to Boron Injection ~

Cold Legs Pumps Boron I nj ec t ion In+vs C7~/H NA NA n 1CH-260 ICH-265 ICH-305 Safety Injection In+m Safet Injection uc on from I~ w.v c NA NA NA Q ICH-306 8HR uction from Sump ICH-311 RllR to RC )lot l.egs

10. ICH-321 RllR to RC llot Legs ll. NPX 151 VI Dead Weiglit Tester

12. PA-343 Containment Service Air

13. SF-151 Refueling Water Supply

14. SF- I 53 Refueling Water Supply

15. SF- 159 Refueling Cavity Drain to Purification System NA

16. SF-160 Refueling Cavity Drain to Puri.fication System NA 1/. SI-1/1 Safety Injection Test Line NA

18. S I - 172 Accumulator Test Line NA uI

k(

   ~l WIE

PLANT SYSTEMS AUXILIARYFEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 At least three independent steam generator auxiliary feedwater pumps I and associated fl,ow paths shall be OPERABLE with:

                           <<gc -5mven

1. Two~feedwater pumps, each capable of being powered from separate f 2.

+~

emergency busses, and

                                +wc-bin~

Onep,feedwater pump capable of being powered from an OPERABLE steam supply system.

b. At least one auxiliary feedwater flowpath in support of Unit 2 shutdown functions shall be available.

APPLICABILITY: Specification 3.7.1.2.a - MODES 1, 2, 3. Specification 3.7.1.2.b - At all times when Unit 2 is in MODES 1, 2, or 3. ACTIONS: When Specification 3.7.1.2.a is applicable: With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in'OT SHUTDOWN within the following 6 hours.

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.

C. With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliary feedwater pump to OPERABLE status as soon as possible. When Specification 3.7.1.2.b is applicable: With no flow path to Unit 2 available, return at least one flow path to available status within 7 days, or provide equivalent shutdown capability in Unit 2 and return at least one flow path to available status within the next 60 days, or have Unit 2 in HOT STANDBY within the next 12 hours and HOT SHUTDOWN within the following 24 hours. The requirements of Specification 3.0.4 are not applica'ole. D. C. COOK - U.'i:: 1 3/4 7-5 Amendment ."o. 92,ic9.:.: 131

1 gj 'g

    'H "pl 4y
~ A 0

I l I I REFUEL".IG OPEI~T.QNS CRANE TRAVEL - SPENT FUEL S,ORAGE POOL BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Loads in excess of 2,SCO oounds shall be prohibi ed from travel ~'ver fuel assemolies in the stor ace oool. Loads carried over the soent fuel pool and the heights at wnich they may be carr..'ed over racks c"ntaining fuel shall be limited in such a way as to oreclude impact energies over 24,240 in.-lbs., if the loads are dropped from the crane. APPLICABiLITY: With fuel assemblies in the storage pool. ACTION: With the requirements of the above specification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0 3 ~ are not applicable. SURVEILLANCE RE UIREMENTS 4.9.7.1 Crane interlocks whicn prevent crane travel with loads in excess of 2,500 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation. 4.9.7.2 The potential impact energy due to dropping the crane's load shall be determined to be ( 24,240 in.'-lbs. prior to moving each load over racks containing fuel.

   "Shared system with     D. C. COOK - UNIT 2 D. C. COOK
     ~  ~
               - UNIT   1             3/4 9-8                  Amendment No. jgg,ll3

l~ II Afz I II

5.0 OESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall .be as shcwn fn Ffgure 5.1-1. LQM POPULATION ZONE 5.1.2 The lcw population zone shall be as shown in Figure 5. 1-2. Site Soundarv Fcr Gaseous and Liquid Ef'uents S.L.3 Qe s1aa bau~ndarnd,far gaseaus and IIqusd erfluena sha11 be shown fn F f gure S-.-i=3.;T-

5. 2 CONTAINMENT CONFIGURATION 5.2.1 The reacto~ containment building fs a steel lined, reinforced concrete building of cylindrical shape, with a deme roof and having the following design features:

a. Nominal fnside diameter

115 feet.

b. Nominal fnside height ~ 160 feet.>>

c. Minimum thickness of concrete walls 3'6".

d. Minimum thickness of concrete roof

2'6".

e. Minimum thickness of concrete floor pad ~ 10 feet.

f. Nominal thickness of steel liner, side and dome 3/8 fnches.

g. Nominal thickness of steel liner, bottom ~ 1/4 inch.

h. Net free volume ~ 1.24 x 10 cubic feet.'

crom graae c ev. a to inside of dome.

0. C. COOK-UNIT 1 5-1 Amendment No. 69

0 0

ADMINISTRATIVE CONTROLS

6. 3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable position, except for (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 response 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 requirements and recommendations of Section 5.5 of ANSI N181-1971 and ggpandkx "A" of 10 CFR Part 55.

6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION 6.5.1.1 The PNSRC shall function to advise the Plant Manager on all matters related to nuclear safety. D. C. COOK - UNIT 1 6-5 Amendment No. 49 ~ ~ 33

4'f ~ 1 tj! r 5 pk 0

ADMINISTRATIVE CONTROLS PZSPONSZBILZTZES 6.5. 1.6 The PNSRC shall be responsible or:

a. Revi.ew of 1) all proc es reau'ed by Soeci 'ation 6. 8 and cnanges thereto, 2) any other proposed procedures or changes thereto as determined by the P'ant Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear safety.

C ~ Review of all proposed charges to Append'x "A" Technical Specifications. Review of all proposed changes or modifications to plant systems or ecuipment that affect nuc'ear safetv.

e. Invest'gat'on of all violat'ors of the Technical Specifications including the preparation and forwarding of reports covering evaluat'on and recommendat'ons to prevent recurrence to the Chairman of the NSDRC.

Review of all REPORTABLE EVENTS. Rev'ew of fac'lity operations to detect potent'al safety hazards. Performance o special reviews, inves .'cations of analyses and reports thereon as recuested by the Chairman of the NSDRC. Review of the Plant Security Plan" and implement'ng procedures and shall ubmit recommended changes to the Chairman of 0he NSDRC. J ~ Review o the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.

k. Rev'ew of every unplanned crsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation and recomme..dat'ons to prevent recurrence to the NSDRC.

Review of changes t'he PROCESS CON ROL PROGB ~l, OFESZ=E DOSE CALCULATZCN:ANNUAL, and radwaste treatment s"stem. D. C. COOK UNIT 1 6-7 Amendment No. 87

Docks t No. 316 Page 5 of 11 (1) Deleted by Amendment 63. (m) Deleted by Amendment 19. (n) Deleted by Amendment 28. (o) Fire Protection Amendment Tha li.censee may proceed rt.th and is required to No. 12 complete the modifications identified in Table 1 of tha Fi.re Protection Safety Evaluation Report for the Donald C. Cook Nuclear Plant dated June 4, 1979. These modifi.cati.ons shill be completed in accordance rith the dates contained in Table 1 of that SER or Supplements thereto, Administrative controls for fire protection as described in the licensee's submittils dated January 31, 1977 and October 27, 1977 shall be implemented ind maintained. Amendment (p) Deleted by Amandmant /g l No'4, 121

0 A,j1 li a" <<J I 0 JQ ~ I P c U

DE."INITIONS SOLIDIFICATION

 ]..29  SOLTDIFTCATEON  shall   be the conversion of radioactive         liquid, resiny and sludge wastes    from  liquid         syscems into a form that meets  shipping and burial site requirements.

OFFSITE DOSE CALCULATION HAiVUAL (ODCA)

 ~.30 The OFFSITE DOSE CALCULATION <<VUAL shall contain the              methodology and parameters used in the calculation of offsite doses due              to radioactive gaseous and liquid effluents, in the calculation of gaseous              and  liquid effluent monitoring alarm/trip secpoints and the conduct of              environmental radiological monitoring program.

GASEOUS RADVASTE TREATMENT SYSTEM'.31 A GASEOUS RADVASTE TREAT~fENT SYSTE.'i is any system designed and installed to reduce radioactive gaseous effluents by. collecting primary coolanc system off-gases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment. VEVTELATION EAST TREATMENT SYSTci 1.32 A VENTELATEON EXHAUST TREATAEiVT SYSTEN is any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by pa'ssing ventilat on or vent exhaust gases through charcoal absorbers and/or HEPA filters fo" the purpose of removing iodines or particulategrom the gaseous exhaust scream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Feature (ESF) atmospher'c cleanup systems are not considered to be VENTILATION EXHAUST TREAT'fENT SYSTEM components. PURGE-PURGING 1.33 PURGE or PURGING is the controlled process of discharging air or gas from a confinement to maintain tempez'ature, pressuze, humidity, concentration or ocher operating condicion, in such a manner that replacement air or gas is required to purify the confinement. VEHTLVC 1.34 VENTiNG is the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidicv, concentration or ocher operating condition, in such a manner that replacement air or gas is not provided or required during VENTTVG. Vent, used in system names, does not imply a VENTING process. D. C. COOK - UNIT 2 1-7 AMENDMENT itO. 51

f}

\"

4

2.1.1 Tne coabination of ~4" PO4:-R, pzessuri"er pressu=e, and the highest operating loop coolant eaperatuze (Tavg ) shall not exceed the Limits sholem in Figure 2.1-1 for 4 loop operat'on. BAHTS: t"henevez the point defined by the coabfnation of the hiehest operating loop average tecrperature and Td"=KM. PO4=% has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY vithin L hour. 2.1.2 The Reactor CooLan" Systen pressu"e shall not exceed 2735 psig. MODES 1 and 2

whenevez the Reactor Coolant System pressuze has exceede'd 2735 in HOT Sd%)BY vith the Reactor Coolant System pressu-e vithin ps', be L&t within 1 its hou". MODES 3, 4 a..d 5 whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressu"e to vf,thin 'ts Limit vithin 5 adnutes. D. C. COOK - POT 2 2-1 NO. 82

1 11 0

TABI "- 2. 2-1 Continuad)

                 , ACTOR TRIP   SYSTEM INSTRU.      iATION TRIP SETPOINTS FUNCTIONAL UNIT            TRIP SETPOINT                      ALLOWABLE VALUES 13.Stcam Generator         Greater than or equal to           Creater than or equal to Water Level-Lov-Lov 21i of narrov range                   19.2% of narrov range instrument span - each             instrument span - each steam generator                    stean generator 14.Steam/Feedvater Flov    Less6than or equal to 1.47         Less6than or equal to 1.56     I Mismatch and Lov       x 10 lbs/hr of stean ilov          x  10   lbs/hr of  steam  flov Stean Cenerator        at RATED THERMAL POWER             ac RATED THERKhL    POWER Water Level            coincident vith stean              coincident vith stean generator vater level              generator vater level greater than or equal to           greacer than or equal to 254 of narrov range                24%   of narrov range instrument span      - each       instrument span - each-acean generator                    stean generator

15. Unde rvo1 tage Greater than or equal to Creater than or equal to t Reactor Coolant 2905 volts - each bus 2870 volts - each bus Pumps 16.Underfrequency- Creacer chan or equal to Greater than or equal co I Reactor Coolant 57.5 Hr, - each bus 57.4 Hx - each bus Pumps 17.Turbine Trip Pl< "~.~ /

h. Lov 2z4p Syagae Creater than or equal to Creacer than or equal to Pressure 58 psig 57 psig B. Turbine Stop Creacer than or equal to Creacer than or equal co Valve Closure 1% open lt open 1&.Safety Infection Not Applicable Noc hpplicable Inpuc from ESF 19.Reactor Coolant Pump Not Applicable Not hpplicable Breaker Position Trip COOK NUC~~ PLAÃE - UNIT 2 2-6 AMENDNBPZ NO. 82~ 134

4~ 4~

3 . 4. 2 PO~ DISTRISUTION LI.-ITS AXIAL FLUX DIFF>~rWCE LIMNI INC CONDITION FOR OPERA ION 3.2.1 che

 'n The target he COLR.

indicated A)(IAL FLLX band about a cargec DIFFER-""'tCE (AFO) f'ux d'fference. The target ~ shall. be maintained vithin is specifi~d APPLICASILITY: RODE 1 above 50% RATED THHviAL POt'ER* AC.:Pl:

               'ich   the indicated AXIAL F'G( DIFFER~ICE oucside of che target band about the cargec                 Lux dif erence and vich THH~AL POt'ER:

Above 90% or 0.9 x APL (vh'chever is Less) of RATH) THD<~L PO'ER, vichin 15 minutes: a) E'ther restore the indicated AFD to within the target band Limits, or b) Reduce THER."aL PO"ER to less chan 90% or 0.9 x APL (vh'chever is Less) of RATED TH~R~iAL POVER. Bet-een 504 and 904 or 0.9 x AP'whichever is Less) af RATED i ~~ R <<La PO~ C,R; a) ": 'ER C?E. KTION may conti.": e provided:

                                       .ne '.'ndi cared     AFO has .".ot been outside of the target      band  for mare "'.".an 1 hour penalcy deviation c.. u1ac've d '."" "Le zrevious 24 hours, and

2) The indicated AFO is -" =.". n che Lim'cs speci" ed in the COLS. Othe. ise, red ce THEB."iAL ?O'NER to less chan 50% of RATED THER."AL ?Ql'ER with' 30 minutes and reduce the Pove. Range Neutron FLux-High Tr'.'p Setpoints "o Less than or equaL co 553 of RATED THERMAL PC'ER vith'n the next 4 hours.

Su .:eilLance test'ng of the Paver Range 'ieucron Flux Channels mav be per"or ed "rsuant to Speccation 4.3.L.L.L provided the'ind'ca"ed AFO is ma'ncained the Lim't specif'ed 'n:he COL%. A cacal o= La hours operat'on may be acc -ulated vich:he AFO outside o= =he target band during ch's testing vithouc penalty dev'c'n. ~ See Specia; .est Except'on 3.'0.2 C OK .'."~C. D9. P~~iT - 'IT 2 3/4 2-L AvENO~ENT NO.44,107,'; "

0 y I

p n O a~l RTf VO MINIMUM TOTAL NO. C))ANHELS C))ANNELS APPLlCABLE

   ]'lJ}f~C"   Og)Qj t)(~                       X9 'HG3'   WORMS~       ~OQP.S

l. )fanual Reactor Trip 1, 2 and*

2. Poser Range, Neutron Flux 4 1, 2 and ~

3. Pouar Range, Neutron Flux 4 2 1, 2

            )llgh Poaitivo Rate

4. ,Power Range) Neutron Flux, 4 11 2

            ))igh )fegativa Rata

5. Intermediate Range, 1, 2 and,~

            }loutron Flux

6. Source Range, Neutron Flux A. Startup 2¹ and

B. S}nftdoun 3, 4 and 5

7. Ovortomporature AT Four Loop Operation 1) 2

a. Ovorpouar hT Four Loop Operation 1, 2 CA ld

TABLE 3.3-1 Continued (- REACTOR TRIP SYSTEM INSTRUMENTATION MINIMUM TOTAL NO- CHANNELS CHANNELS APPLICABLE n UNCTIONAL UN T OF CHANNELS TO TRIP OPERABLE MODES ACTION no O Iq

9. Pressurizer Pressure-Low 1, 2 I

10. Pressurizer Pressure--High 1, 2 Q

ll. Pressurizer Water Level High 2 2 1, 2

12. Loss of Flow Single Loop 3/loop 2/loop in 2/loop in (Above P-8) any opera- each opera-ting loop ting loop

13. Loss of Flow Two Loops 3/loop 2/loop in 2/loop in 1 (Above P-7 and below P-8) two opera- each opera-ting loops ting loop

14. Steam Generator Water 3/loop 2/loop in 2/loop in . 1, Level 2

Low-Low any opera- each opera-ting loop ting loop

15. Steam/Feedwater Flow 2/loop-level 1/loop-level 1/loop-level 1, 2 Mismatch and Low Steam and coincident and Generator Water Level 2/loop-flow with 2/loop-flow mismatch in 1/loop-flow mismatch or same loop mismatch in 2/loop-level same loop and 1/loop-flow mismatch 0

l 0

O 0 TAliLE 3~3-1 Concinu <I) n REACTOR TRIP SYSTEM INSTRUMENTATION MIHIMUM TOTAL NO. Cl lANNELS CltANNELS APPLICABLE FUNCTIOHAL UHIT OF CllAHHELS TO TRIP OP ERA BI.E MODES ACTION P4 M

16. Undervoltage-Reactor Coolant Pumps 4- 1/bus

11. Underfrequency-Reactor Coolant Pumps 4-1/bus lr>. Tuel>i>>u Trip A. Lou Fluid Oil Pressure B. Turbine Stop Valve Closure

19. Safety Injection Input from ESF 1, 2

20. Reactor Coolant Pump B reake r Position Trip Above P-/ I /1> reak<<r 1/breaker per operat-ing loop

21. Reactor Trip Breakers 1, 2, 1, 13.

3* 4* 5* 14

22. Automatic Trip Logic

0 TABLE 3.3-6 (Cont nued) TABLE VOTgTION CTION 20 - Vi.th the number of channels OPERABLE less than required by the Ainimum Channels Operable requirement,, comply with the ACTION requirements of Spec fication 3.4. 6. 1. ACTION 21 Pith the number of channels OPERABLE less than required by the Ainimum Channels Operable requirement, perform area surveys of the monitored area with portable monitoring instrwentation at least once per day. ACTION 22 Pith the number of channels OPERABLE less than required by the Ainimum Channels Operable requirements, comply with the ACTION requirements of Specf'cation 3.9.9. This ACTION is not requ'red during the performance of containment integrated leak rate test. ACTION 22A - Pith the number oi OPERABLE Channels less than required by the Ainimum Channels OPERABLE requirements: e'ther restore the inoperable Channel(s) to OP qABLE status within 7 days of the event, or

2. prepare and submit a Special Repor" to the Commission pursuan" to Specification 6.9.2 "ithin 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedu'e for rqstoring the system to OPERABLE status. g-[~5

3. Technical Specificat'on Sect'ons 3.0.3, 3.0.4 Not Applicable.

ACTION 22B - 'ith the number of OPERABLE Channels less than required by the Ainimum Channels OPERABLE requ'rements.

l. either restore the inoperable Channel(s) to OPERABLE status within 7 days of the even", or

2. prepare and submit a Special Repor" to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

3. In the event of an accident involving radiological releases initiate the preplanned alternate method of monitoring the appropriate parameter(s) within 72 hours.

Not Applicable. D. C. COOK - UNIT 2 3/4 3-36 Amendment No. 80,119

t IA gJ i~ l P 4' 'I E

TABLE 3.3-11 Unit 2 and Common Area Fire Detection S stems Total Number Detection S stem Location of Detectors Heat Smoke (x/y)%'lsae (>/7)* (x/y)* Auxiliary Building a),",Elevation 587 55/OC b)(Elewatfon 609 41/OC c},"Elevation 633 d) evat on e Elevation 650 41/OC 23/OC 34/OC

f) Nev Fuel STGE brea 4/OC U2 East Main Steam Valve Enclosure 28/0~ U2 Main Steam Line brea El. 612 (Around Containment) ]3/Pk+ U2 NESTS Valve Area El. 612 2/0 U2 4KV Svitchgear (AB) 0/3 0/2 U2 4KV Svitchgear (CD) 0/3 0/2 U2 Engr. Safety System Svi.tchgear 6 XFMR. Rm. 0/5 0/14 U2 CRD, XFMR 6 Svt.tchgear Rm. Inverter & AB Bttry. Rms. 0/5 0/17 U2 Pressurixer Heater XFMR. Rm. 12/0 U2 Diesel Fuel Oil XFMR. Rm. 0/1 U2 Diesel Generator Rm. 2AB 0/2 U2 Di,esel Generator Rm. 2CD 0/2 U2 Di.esel Generator Ramp Corr. 4/0 Ul&2 AFVP Vestibule 2/OC U2 Control Room 42/0 U2 Svitchgear Cable Vault 0/10~ 0/13 U2 Control Rm. Cable Vault P/76~ U2 Aux. Cable Vault 0/6 Ul&2 ESV Basement Area 4/OC U2 ESV Pump & MCC Rms. 9/0 C System protects area common to both Units 1 and 2

*(x/y) x is number of Functi.on h (early varning fire detecti.on        and notification only) instruments.

y is number of Function B (actuation of fire suppression systems and early varning and notifi.cation) instruments. circuit contains both smoke and flame detectors

*~        tvo circui.ts of five detectors each tvo circuits of 38 detectors each COOK NUCLEAR PLANT     - UNIT  2              3/4 3-52               AMENDMENT   N0.61, 115

T 3.3-10 POST-ACCIDENT RING INSTRUHENTATION n

INSTRUHENT HINIHUH CllANNELS OPERABLE n

1. Containment Pressure 2 Re a c to r Co o 1 a n t 0 u t 1 e t Te m p e ra t u re - T Il Id e R a n g e )

I l0 T (

g gH 3. Reactor Coolant Inlet Temperature - T (Wide Range) 2

4. Reactor Coolant Pressure - Wide Range

5. Pressurizer Mater Level

6. Steam Line Pressure 2/Steam Generator

7. Steam Generator Mater Level - Narrow Range 1/Steam Generator S. Refueling Mater Storage Tank Mater Level

9. Boric Acid Tank Solution Level

10. Auxiliary Feedwater Flow Rate 1/Steam Generator*

Ds I ll. Reactor Coolant System Subcooling Hargin Honitor

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

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

14. Safety Valve Position Indicator - Acoustic Honitor l/Valve

15. Incore Thermocouples (Core Exit Thermocouples) 2/Core Quadrant

16. Reactor Coolant Inventory Tracking System One Train (3 channels/Train)

(Reactor Vessel Level Indication)

17. Containment Sump Level 1****

18. Containment Mater Level 2**4*

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

            **          &~i'ubcooling margin readout can be used as a substitute for the                                                    monitor instrument.
            *** Acoustic monitoring of PORV position (1 channel per three valves - headeredsubcooling                              discharge) can be used as a i t-'..'             substitute for the PORV Indicator - Limit Switches instruments.

0*** The 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 transmf.tters is describe" in the Bases.

                  -Amendment-No-.-gg-;        95-{Effect)ve-before"start-up                     fN'tON7ttgt'~ur ~~scheduled

s ~i Ig t'y lg P'! I f 0 l i4 ~ l

EACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued

2. With two or more block valves inoperable, Within 1 hour 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 hour 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

 .4.11.1   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.

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 3.4.11.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 4.8.2.3.2.d.

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

COOK NUCLEAR PLANT - UNIT 2 3/4 4-33 AMENDMENT NO. g, gj, ]31

~ (k ~ I

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

a. At 1'east once per 12 hours by verifying that the following valves are Valve Number Valve Function Valve Position

a. IMO- 390 a. RWST to RHR a. Open

b. IMO-315 b. Low head SI b. Closed to Hot Leg

c. IMO-325 c. Low head SI c. Closed to Hot Leg

d. IMP-262* d. Mini flow line d. Open

e. IMO-263* e. Mini flow line e. Open f . IMO-261* f. SI Suction f. Open

g. ICM-305* g. Sump Line g. Closed

h. ICM-306* h. Sump Line h. Closed

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

c ~ By a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and

2. Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

These valves must change position during the switchover from injection to recirculation flow following LOCA.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-4, AMENDMENT N0.7$ ,131

lg 1d, e gi 5 ~

CONTAINMENT SYSTEMS CONTAINHENT AIR LOCKS LIMITING CONDITION FOR OPERATION 3.6.1.3 Each containmen air lock shall be OPERABLE with:

a. Both doors closed exce'pt when the air lock is being used for normal transit entry and exit through he containment, then at least one air lock d"or shall be closed, and

b. An overall air lock leakage rate of <<0.05 L at: P 12 sg ps i g.

APPLICABILITY: MQQES I, 2, 3 and 4. ACTIQM: With an air lock inoperable, maintain at least one door closed; restore the air lock to OPERABLE status within 24 hours or be in at least HGT STAIVGBY within the next 6 hours and in COLD SHUTOOWN within the ol1owinc 30 hours. SUR EILLANC REOrJIREN N 4.6.1.3 Each containment air lock shall be denonstrated OPERABLE:

a. *After each opening, except when the air lock is beina used for multiple entries, then at least once per 72 hours, by performing an air leakage test without a simulated pres-sure force on the door by pressurizing the volume between the door seals to 12 psig and verifying a seal leakage rate of no greater han 0.5 L .

b A a er.aoaIng an air leakage test withou a simulated pressure force on the door per

Specification 4.6.1.3.a; then by performing an air leakage with a simulated pressure force on the do~r by pressurizing the volume betwe n the door seals to 12 psig and verifying seal leakage rate of no areater than 0.0005 L, "Exemption to Appendix "J" of 10 CFR 50.

0. C, COOK UNIT - 2 '3/4 "-4

   . p~ fa~

I I Pj! g4 k 0

I CONTAINMENT T 5;YSTEHS SURVEILLANCE REQUIREMENTS (Continued)

c. At least once per 6 months Py conducting an overall air lock leakage test at P +12 psig) and by verifying that the overall air lock leakage kate is within its limit.

    -'d. At least once per     6 months by  verifying that only one door in each air lock can     be opened at  a  time.

0. C. COOK,- UNIT 2 3/4 6-5

1 k lc t ft I t'

TABLE 3.6- 1 (Continued) CONTAINHFNT ISOLATION VALVES CD CD PZ ISOLATION TIME VALVE NUtlBER FUNCTION Itl SECONDS D. HANUAL ISOLATION VALVES (1) (Continued)

3. 1CH-250 Boron Injection 4aLat. NA

4. 1CH-251 Boron Injection In+et NA

5. 1CH-260 Safety Injection I nlwt

6. 1Cti-265 Safety Injection I+1~

                                           ~c. 7

7. 1CH-305 IIIIR/Suction From Sump NA

8. 1CH-3C6 RIIR,'uction From Sump

9. 1CH-311jt RIIR to RC Ilo t Legs

10. ICH-321II RIIR to RC llot Legs NA E . OTIIER

1. CS-442-1 Seal Wtr. to RCP Nl

2. CS-442-2 Seal Wtr. to RCP 82 NA

3. CS-442-3 Seal Wtr. to RCP N3 NA

4. CS- 442- 4 Seal Wtr. to RCP N4

%tj qT SYSTEMS AE LIARY FEEDWATER SYSTEM 1" ~u>" LIMITING CONDITION FOR OPERATION 3.7.1.2

a. At least t ee independent steam generator auxiliary feedwater pumps and associate flow paths shall be OPERABLE with:

Two eedwater pumps'ach capable of being powered from separate emergency busses, and

2. Oneee/eedwater pump capable of being powered from an OPERABLE steam supply system.

b. At least one auxiliary feedwater flow path in support of Unit 1 shutdown functions shall be available.,

APPLICABILITY: Specification 3.7.1.2.a - MODES 1, 2, 3. Specification 3.7.1.2.b - At all times when Unit 1 is in MODES 1, 2, or 3. ACTIONS: S pecification 3.7.1.2.a is applicable:

a. With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pumps to OPERABLE status within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDO'".f within the -following 6 hours.

b. With two auxiliary feedwater pumps inoperable, be in at least HOT STANDBY within 6 hours and in HOT SHUTDOWN within the following 6 hours.

With three auxiliary feedwater pumps inoperable, immediately initiate corrective action to restore at least one auxiliarf feedwater pump to OPER%LE status as soon as possible. When Specification 3.7."1.2.b is applicable: With no flow path to Unit 1 available, return at least one flow path 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 hours and HOT SHUTDOWN within the following 24 hours. The requirements of Specif cation 3.0.4 are not applicable. COOK - UNIT 2 3/4 7-5 Amendment No. 82 s > ~6

ELECTRICAL POWER SYSTEMS 3/4.8.3 Alternative A.C.

                ~
     ~                ~ ~ Power Sources LIMITING CONDITION   FOR OPERATION 3.8.3.1 The steady state bus voltage for the manual alternate reserve source" shall be greater than or equal to 9OX of the nominal bus voltage.

APPLICABILITY: Whenever the manual alternate reserve source (69 kV) is connected to more than two buses. ACTION: With bus voltage less than 90K nominal, adjust load on the remaining buses to maintain steady state bus voltage greater than or equal to 90K limit. SURVEILLANCE RE UIREMENTS s 4.8.3.1 No additional surveillance requirements other than those requiraed by Specifications 4.8.1.1.1 and 4.8.1.2,

Shared with D. C.. Cook Unit g.

D. C. COOK - UNIT 2 3/4 8-20 AMENDMENT NO, ~2

g 0 REFUELING OPERATIONS CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING* LIMITING CONOITION FOR QPERATiON 3.9.7 Loads in excess of 2,500 pounds shall be prohibited from travel over fuel assemblies in the storage pool. Loads carried over the spen fuel pool and the heights at which they may be carried over racks con air i.. fuel sha'll be limited in such a way as to preclude impact energies over 24,240 in.-lbs., if the loads are dropped from the crane. APPLICABILITY: Mith fuel assemblies in the storage pool. ACTION: With the requirements of the abo~pecification not satisfied, place the crane load in a safe condition. The provisions of Specification 3.0.3 are not applicable. SURVEILLANCE R EQUI REMEN TS 4.9.7.t i I k loads in excess, of 2,500 pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at leas: once per 7 days thereafter during crane operation. 4.9.7.2 The potential impact energy due to dropping the crane's load shall be determined to be < 24,240 in.-lbs. prior to moving each load over racks contain~ng fuel. ~Shared system with 0. C. COOK - UNIT i 0.~ C.

    ~  COOK  - UNIT2                    3/4 9- 7                 Amendment Ho. 8786

41 4 tg 1 v

L 3~9 ~9 sSLc Qc ca ~ p- ge @cd'ollusc solac 4 5'75cM sha~~ OPDQL=-. Dur "g Care ALceracioo5 or aave=cac of f.wadiaccd fuel vi-him cue coute' azc. srzaM: Mi 5 cbc C"acai;=e" c Pur~e ed bhau5= 'salacious sysceP '=-opcrabgc, c ose cac j di 'c c Bur c a d Mau5c pe c a o.os p 0<i~ .o$ dircc~ accc55 f ~ Che COG a'=-C"- a=aspee e CQ Co.e du 5'EC a~SChere ~ 'stowe prov.'5 ="5 of Spec'"cacio" 3.0.3 'a=a coc a-ylicablc. 4.9.9 ~se Ca="a'--"" ~"-g a"H ""chouse malat'ca sys- stroll >c 5-=aced OP~EL:- - =; " 0 '-ours pr='ot co =he sar= o" and a- Least dace pcr 7 days duri"g C"~"- 0:~~'0!iS oy 'er =g '-, ="ac coc 5'"-e"c ?urge and ause 5OLac , occu 5 oQ ~ ual 5 a oc a d ozL a h gh a~c cu c5 sigma'-~ ca~ o" chc coaca'~e" c radiaciau "oaicar'"g ~c=~ccLcac'ocL channels.

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6. 0 ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except for (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 response and analysis of the plant for transient and accidents.
6.4 TRAINING 6.4.1 A retraining and replacemenc training program for the facility staff shall be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and AppeacL4e-"A" of 10 CFR Part 55. 6.5 REVIEW AND AUDIT
~~6. 5. 1 PLANT NUCLEAR SAFETY REVIEW COLMITTEE PNSRC 6.5.1.1 The PNSRC shall function to advise the Plant Manager on all matters related to nuclear safety.~~
COMPOSITION 6.5.1.2 The PNSRC shall be composed of the: Chairman: Plant Manager or Designee Member: Assistant Plant Manager - Maintenance Member: Assistant Plant Manager - Operations Member: Operations Superintendent Member: Technical Superintendent - Engineering Member: Technical Superintendent - Physical Sciences Member: Maintenance Superintendent Member: Plant Radiation Protection Supervisor Member: Qc Superintendent D. C. COOK - UNIT 2 6-5 Amendment No. 73,118
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