ML17331B073

Proposed Tech Specs,Reflecting Administrative & Editorial Changes

ML17331B073
Person / Time
Site:	Cook
Issue date:	11/15/1993
From:	INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:
Shared Package
ML17331B072	List: ML17331B071 ML17331B073
References
NUDOCS 9312010010
Download: ML17331B073 (71)

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TABLE 3.3-1 Continued DESIGNATION CONDITION AND SETPOINT FUNCTION P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats Flux Channels greater than or the automatic block of equal to 11% of RATED THERMAL reactor trip on: Low POWER or 1 of 2 Turbine First flow in more than one Stage Pressure channels gzeater primary coolant loop, than or equal to 37 psig. reactor coolant pump under-voltage and undez<<frequency, tuzbine trip, pressurizer low pressure, and pressur-izer high level ~w flow in a particular loop can be evidenced by either a detected low flow or by the opening of the reactoz coolant pump breaker.

P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats Flux channels greatez than or the automatic block of equal to 31% of RATED THERMAL reactor trip caused by a lo~kSW coolant flow condition in a single loop,~

P-10 With 3 of 4 Power range neutron P-10 prevents or flux channels less than 9% of defeats the manual RATED THERMAL POWER. block of: Power range low setpoint reactor tri p, Intermediate range reactor trip, and intermediate range rod stops'rovides input to P-7.

COOK NUCLEAR PLANT - UNIT 1 3/4 3-9 AMENDMENT NO. SSI 120 93120i00i0 PDR ADOCK 93iii5 050003i5 P PDR

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TABLE 3.3<<10 (Continued)

Un t and Commo ea e etectio Systems Total Number Detector S stem Locat o o Detectors peat ~arne Smoke (x/y)* (x/y)* (x/y)*

Ul Cable Tunnels a) Quad 1 Cable Tunnel 0/3 0/4 b) Quad 2 Cable Tunnel 0/4 0/7 c) Quad 3N 0/3 0/4 d) Quad 3S 0/3 0/3 e) Quad 3M 0/3 0/4 f) Quad 4 0/5 0/6 Ul Charcoal Filter Ventilation Units a) HV-AES-1 0/1~

b) -AES-2 0/1~

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HV-ACRF HV-CIPX HV-CPR 12-HV-AFX 0/1~

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a) RCP 1 1/0 b) RCP 2 1/0 c) RCP 3 1/0 d) RCP 4 1/0 e) Cable Trays 58/p~~

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.

Originally installed to automatically deluge charcoal filters.

However, manual actions are nov necessary.

The fire detection Instruments located within the Containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate tests.

Thermistors are located within all cable trays which contain combustible cables, in both upper and lover containment throughout quadrants 1-4.

OOK NUCLEAR PLANT - UNIT 1 3/4 3-53a A?KNDMENT NO. kgb', 172

OS - CC E 0 0 G S U E T 0 MINIMUM

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l. Containment Pressure 2

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

3. Reactor Coolant Inlet Temperature- Tco~ (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 Lovel- Narrow Range 1/Steam Generator

8. Refueling Water Storage Tank Water Level 2

9. Boric Acid Tank Solution Level 1

10. Auxiliary Foedwater Flow Rate 1/Steam Generator*

11. Reactor Coolant System Subcooling Margin Monitor 1%*

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. Incoro Thermocouples (Core Exit Thermocouples) 2/Core Quadrant

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

(Reactor Vessel Level Indication)

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17. Containment Sump Level 1&, Containment Mater Level ~i44a 2

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 bo used as a substitute for the subcooling monitor instrument.

Acoustic monitoring of PORV position (1 channel per three valves - headered discharge) can be used ss a substitute for the PORV Indicator - Limit Switches instruments.

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-the=4ev~mns mdi~t~cous R=107Ww COOK NUCLEAR PLANT - UNIT 1 3/4 3-55 Amendment No. 406, 4&t, 168

1 s s S 1' TABLE 4.3-7 POST<<ACCIDENT MONITORING INSTRUMENTATION SURVEILIANCE RE UZK2KNTS CHANNEL CHANNEL INSTRUMENT CNBCN CALTBBATTON

1. Containment'ressure

2. Reactor Coolant Outlet Temperature-HOT (Vide Range)

3. Reactor Coolant Inlet Temperature-TCO (Wide Range) M R

4. Realc or Coolant Pressure-Wide Range M R

.5. Pressurizer Water Level M R

6. Steam Line Pressure M R

7. Steam Generator Water Level-Narrow Range R

8. RWST Water Level M R

9. Boric Acid Tank Solution Level M R 10.Auxiliary Feedwater Flow Rate M R 11.Reactor Coolant System Subcooling M R Margin Monitor 12.PORV Position Indicator - Limit Switches M R 13.PORV Block Valve Position Indicator- ~

M R Limit Switches 14.Safety Valve Position Zndicator-Acoustic Monitor 15.Incore Thermocouples (Core Exit Thermocoup les)M R(l) 16.Reactor Coolant Inventory Tracking System M(2) R(3)

(Reactor Vessel Level Tndication)

17. Containment Sump Level~ M R 18.Containment Water Level M R (1) Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

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

s (3) Coaplerion of ohannel elaibrarion for sensors eo be perforned below P-13 in MODE 3.

-~hese Dmeruments=wk3.-1Mecome=eke COOK NUCLEAR PLANT - UNZT 1 3/4 3-56 AMENDMENT NO. $ $ ,144

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~~ON OD Diesel Generator lAB Room Cross-zoned Heat Diesel Generator 1CD Room Cross-zoned Heat Diesel Generator Fuel Oil Pump Room Heat 4 KV Switchgear Rooms Manual Control Rod Drive, Transf. Switchgear Rooms Manual Engineered Safety Switchgear Room Switchgear Room Cable Vault Cross-zoned Ionization and Infrared Auxiliary Cable Vault Ionization Control Room Cable Vault (Backup)* Manual Penetration Cable Tunnel Quadrant 1 Manual Penetration Cable Tunnel Quadrant 2 Manual Penetration Cable Tunnel Quadrant 3N Manual Penetration Cable Tunnel Quadrant 3M Manual Penetration Cable Tunnel Quadrant 3S Manual Penetration Cable Tunnel Quadrant 4 Manual 2

• Control Room Cable Vault CO System i's only required to be operable when the Cable Vault Halon System is-operable.

COOK NUCLEAR PLANT - UNIT 1 3/4 7-41 AMENDMENT NO.

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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: 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 app'licable.

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 ( 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. 797,113

RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The gaseous radvaste treatment system and the ventilation exhaust treatment system shall be used to reduce the radar.oactive materials in gaseous vaste priox to their discharge vhen the proJected gaseous effluent aix doses due to gaseous effluent releases to unrestricted ax'eas (See Figure 5.13) vhen averaged over 31 days, vould exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous vaste prior to their discharge vhen the proJected doses due to gaseous effluent releases to unrestricted areas (See Figure $ .1-3) vhen averaged over 31 days vould exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. Vi.th gaseous vaste being discharged vithout treatment and in excess of the above limits, prepaxe and submit to the Commission vithin 30 days, puxsuant to Specification 6.9.2, a Special Report which includes the folloving information:

Identification of the inoperable equipment or subsystems and the reason for inoperability.

2. Action(s) taken to restoxe the inoperable equipment to operable status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UZREHENTS 4.11.2.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be proJected at least once per 31 days in accoxdance vith the ODCM, whenever the gaseous waste treatment system or ventilation exhaust treatment system is not operational.

COOK NUCLEAR PLANT - UNIT 1 3/4 11-12 AMENDMENT NO.

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DESIGN FEATURES CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3613 fuel assemblies.  !

5.7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified a's Category I Items in the FSAR shall be designed and maintained to the ori.ginal design provisions contained in the FSAR with allowance for normal degradati.on pursuant to the applicant Surveillance Requirements.

5.8 METEOROLOGICAL TOV7ER LOCATION 3

5.8.1 The meteorological tower shall be located as shown on Figure 5.1P.

5.9 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.9.1 The components identified in Table 5.9-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.9-1.

COOK NUCLEAR PLANT - UNIT 1 5-9 AMENDMENT No. 8, 127 I 169

ADMINISTRATIVE CONTROLS

6. 1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in vriting the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor (or during his absence from the control zoom complex, a designated individual) shall be responsible for the control room command function. h management directive to this effect signed by the Vice President - Nuclear Operations shall be zeissued to all station personnel on an annual basis.

6.2 ORGANIZATION ONSZTE hND OFFSITE ORGANIZATIONS 6.2.1 Onsite and offsite organizations shall be established for unit operation and cozporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear pover plant.

Lines of authority, zesponsibQiry, and communication shall be established and defined for the highest management. levels thzough intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts vill be documented in the and updated in accordance vith 10 CZR 50.71(e). vP'~4 ~

b. The Plant Manager shall be responsible for overall unit safe operation and. shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

.c. The Vice President - Nuclear Operations shall have corporate responsibi.lity for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff

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in operating, maintaining, and providing technical support to the plant to ensuze nuclear safety.

d. The individuals vho train the operating staff and those who carry out health physics and quality. assurance functions may report to the appropriate onsite manager; hovever, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

FACILITY STAFF 6.2.2 The Facility organization shall be sub]ect to the follovtng:

a. Each .on duty shift shall be composed of at least the minimum shift crew composition shovn in Table 6.2-1.

COOK NUCLEAR PLANT - UNIT 1 6-1 AMENDMENT NO.7', gqr,154

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6. 3 FACILXTT STAFF UALIFICATIONS 6.3.L Each member of the faciliey staff shall meet or exceed tha minimum qualifica tions of AHS X N18.1 1 971 for comparable positions, except for (1) che Plant Radiation Protection Manager, vho shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor, vho shall have a bachelor's degree oz equivalent in a scientific or enginaexing discipline vieh specific tzaining in plane design, and response and analysis of the plant for tzansients and accidents and, (3) the Operations Superintendent, vho must hold or have held a Seni.or Operator License as specifi.ed in Section 6.2.2.h.

6. 4 TRAZHING 6.4.1 A retraining and replacement training pxogram for ehe facility staff shall be maintained under the direction of the Training Manager and shalL meet or I exceed tha requirements and recommendations of Section 5.5 of ANSI H18.1-1971 and 10 CFR Pare 55.

6.5 REVIEW AHD AUDIT

6. 5. 1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PHSRC FUNCTION 6.5.L.L The PHSRC shalI function to advise the Plant Manager on all matters related co nucleaz safety.

COMPOSITION 6.5.1.2 The PNSRC shall be composed of hssistanc Plant Managers,'epartment Superintendents, or supervisory personnel reporting directly to the Plant Manager, Assistant PLant Managers or Department Superincandents from ehe functional areas listed belov:

Licensing hctivt.ties Technical Support Safety 6 Assessmanc Radiation Protection Operaeions Maintenance The Chairman, his alternate and 'other members and chair alternates. of che PHSRC shall be designated by the Plane Managez. Zn addition to the Chairman, the PNSRC membership shall consist of one individual from each of the areas designated above.

k PNSRC members and altexnaces shall meet or exceed the minimum qualificati.ons of AHSX NL8.1-1971 Section 4.4 for compazabla posieions. The nuclear pover plant operations. individual shall meet ehe qualifications of Section 4.2.2 of AHSL N18.1-1971 excepe for ehe zequirement to hold a cuzzanc Senior Operaeor License.

Tha operaeions individuaL muse hold or have held a Senior Operacor License at Cook Nuclear Plane or ~ similar reactor. The maintenance individual shall meet the qualifications of Section 4.2.3 of ANSX N18.1-1971.

COOK NUCLEAR PLANT - UNIT 1 6-4 AMENDMEHT NO. P'g, f$ $ /7,1

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INSTRUYENTATION EASES 3 4. 3. 3. 5 REMOTE'HUTDOQN INSTRUMENTATION OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control roon. This capability is required in the event contral room habitability is lost and i.s consistent with General Design Cri.teria 19 of 10 CFR 50.

3 4.3.3.5. 1 APPENDIX R REMOTE SHUTDOMH INSTRUMENTATION The OPERABILITY of the Appendix R remote shutdown instrumentation ensures that sufficient instrumentation is available to permit shutdown of the facility to COLD SHUTDORf condi.tions at the local shutdown indication (LSI) panel. In the event of a fire, normal power to the LSI panels may be last.

As a result, capability to repair the LSI panels fram Unit 2 has been provided. If the alternate power supply is not available, fire watches will be established in those fire areas where loss af normal power to the LSI panels could occur in the event of fire. This will consist of either establishing continuous fire watches or verifying, OPERABILITY of fire detectors per Specification 4.3.3.7 and establishing hourly fire watches.

The details of how these fire watches are to be implemented are included in, a plant procedure.

3 4.3.3.7 FIRE DETECTION INSTRUMENTATION SYSTEMS DETECTORS OPERABILITY of the fire detcctian systcms/detectors ensures that adequate detection capability i.s available for the prompt detection of fires. This capability is required in order to detect and locate fires in their carly stages. Prompt detection of the fires .will reduce the potential for damage to safety related systems or components in the areas of the specified systems and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection systems i.s inoperable, the ACTION statements provided maintain the facility's fire protection program and allows far continued operation of the facili.ty until the inoperable system(s)/detectar(s) are restored to OPERABILITY. However, i.t is not our intent to rely upon the compensatory action for an extended period of time and action will be taken to restore the minimum number of detectors to OPERABLE status ~ithin a reasonable period.

3 4.3.3.8 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentati.on ensures that sufficient information is available on selected plant parameters to monitor and assess these variables duri.ng and following an accident. 4 C4&%44~ch edl13rO~

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

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~ l EFUELINC OPERATIONS BASES 3 4.9.10 AND 3 4.9.11 VATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to zemove 99i of the assumed 10% iodine gap activity released from the rupture of an irzadiated fuel assembly. The minimum vacez depth is consistent with the assumptions of che accident analysis. Mater level above the vessel flange in MODE 6 vill vary as the zeaccor vessel head and the system internals are removed. The 23 feet of vater are required before any'ubsequent movement of fuel assemblies oz'ontrol rods."

3 4.9.12 STORAGE POOL VENTILATION SYSTEM The limitations on the storage pool ventilation system ensure chat all zadioaccive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine zemoval capacity are consistent with the assumptions of the accident analyses.

The 1980 version of ANSI N510 is used as a tasting guide. This standard, however, is intended to be rigorously applied only to syscems which, unlike the storage pool ventilacion system, are designed to ANSI N509 standards. For the specific case of the aireaerosol mixing uniformity test requiz'ed by ANSI N510 as a prerequisite to in-place leak testing of charcoal and HEPA filters, the air-aerosol uniform mixing cast acceptance cr'aria were noc rigozously met. For this reason, a statistical correction factor vill be applied to applicable surveillance case results vhare required.

In order co maintain the minimum negative pressure required by Technical Specifications (1/8 inch M.G.) during movemenc of fuel vichin the storage poo'1 or during crane opezacion vith loads over the pool, the cz'ane bay roll-up door and the drumming room roll-up door, located on the 609-foot elevation of the auxiliary'uilding, must be closed. Hovever, they may be opened during these operations under administrative control. If cha crane bay door needs to be opened during fuel movement, an example of an administzative control might be to station an individual at the door who would beos in cosssunioation with personnel in'he spent fuel pool area and oould the door when passage was completed or in the event of an emergency, For the drumming room door, an example of an administrative concz'ol might be to require tha door co be z'eclosed after normal ingress and egress of personnel or material, or to station an individual at the door if the door needs co remain open for an extended period of time.

Should the doors become blocked or stuck open vhile under administzative contzol, Technical Specification requirements vill noc be considered to be violated provided the Action Statement requirements of Specification 3.9.12 .

az'a expeditiously followed, i.e., movement. of fuel within the storage pool. or crane operation vith loads over cha pool is expeditiously suspended.

D. C. COOK - UNIT 1 B 3/4 9-3 Amendment No. 7g p 124

O TABLE 3.3- 1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION MIN IMUM I TOTAL NO. CHANNELS CiiANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION H

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

17. Unde rfrequency-Reactor Coolant Pumps 4- I/bus 3 1 iti. Turbin<<Trip A. Low Fluid Oil Pressure B. Turbine Stop Valve Closure

19. Safety Injection Input from ESF ), 2

20. Reactor Coolant Pump Breaker Position Trip Above P-7 1/br cake r 1/breaker per operat-ing loop

21. Reactor Trip Breakers 1, 2, 1, 13, 3*, 4*, 5* 14

22. Automatic Trip Logic 2 i+

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o ued With 2 of 4 Power Range Neutron P-7 prevents or defeats Flux Channels > 11$ of RATED the automatic block of THERMAL POWER or 1 of 2 Pressure reactor trip on: Low Before the First Stage channels flow in more than one

> 51 psig. primary coolant loop, ~

reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressurizer high level. Low flow in a particular loop can be evidenced by either a detected low flow or by the opening of the reactor coolant pump breaker.

P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats Flux channels > 31% of RATED the automatic block of THERMAL. POWER. reactor trip caused by bar- a low coolant flow condition in a single loop, P-10 'Pith 3 of 4 Power Range Neutron.  ? -10 prevents or defeats Flux channels < 9S of RATED the manual block of:

THER.fAL POWER. Power range low setpoint reactor trip, Intermediate range reactor trip, and intermediate range rod stops ~

Provides input to P-7.

D. C. COOK - UNIT 2

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1. Containment Pressure 2'

2. Reactor Coolant Outlet Temperature - Tso< (Mide Range)

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

4. Reactor Coolant Pressure - Wide Range 2

5. Pressurizer Water Level 2

6. Steam Line Pressure 2/Stea'm Generator

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

8. Refueling Water Storage Tank Mater Level 2

9. Boric Acid Tank Solution Level 1

10. Auxiliary Feedwater Flow Rate 1/Steam Generator*

11. Reactor Coolant System Subcooling Margin Moni.tor 1**

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

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

14. Safety Valve Position Indicator - Acoustic Monito&~+- 1/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 1S. Containment Water Level Steam Generator Mater 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 substi.tute 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 Indicator - Limit Switches instruments.

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CHANNEL CHANNEL INSTRUMENT CHECK CAI.lBRATIOH

1. Containment Pressure M R

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

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

4. Reactor Coolant Pressure - Wide Range COLD LD R

5. Pressurizer Water Level R

6. Steam Line Pressure R

'j. Steam Generator Water Level - Narrow Range M R

8. RWST Water Level M

9. Icteric Acid Tank Solution Level M

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

12. PORV Position Indicator - Limit Switches M R

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

14. Saf'ety Valve Position Indicator - Acoustic Monitor M R

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

16. Reactor Coolant Inventory Tracking System M(2) R(3)

(Reactor Vessel Level Indication)

17. Containment Sump Level~ R

18. Containment Water Level~ R Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

With one train of Reactor Vessel Level Indication inoperable, Subcooling Margin Indication and Core Exit Thermocouples may be used to perform a CHANNEL CHECK to verify the remainI.pg Reactor Vessel Indi. cation train OPERABLE.

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

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COOK NUCLEAR PLANT - UHIT 2 3/4 3.-47 AMENDMENT NO. 9g, gg, f3)

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TABLE 3.3-,11 Unit 2 and Common Area Pire Detection Systems Total Number Detec'tion S stem Location of Detectors Heat Elame Smoke (x/y)+ '(x/7)+ (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) Nev Fuel STGE Area 4/OC U2 East, Main Steam Valve Enclosure 28/0+a U2 Hain Steam Line Area El. 612 (Around Containment) 13/P~

U2 NESTS Valve Area El. 612 2/0 U2 4KV Svitchgeax (AB) 0/3 0/2 U2 4KF Svitchgear (CD) 0/3 0/2 U2 Engr. Safety System Svitchgear & XFMR. Rm. 0/5 0/14 U2 CRD, XFHR & Svitchgear Rm.

Invertex 6 AB Bttxy. Rms. 0/5 0/17 pQ4e ~g Qv U2 Pressurizer Heater . Rm. 12/0 U2 Di.esel Fuel Oil . Rm. 0/1 U2 Diesel Generator Rm. 2AB 0/2 U2 Diesel Generator Rm. 2CD 0/2 U2 Diesel Generator Ramp Corr. 4/0 U162 AFVP Vestibule 2/OC U2 Control Room 42/0 U2 Svitchgeax Cable Vault 0/10%%+ 0/13 U2 Contxol Rm. Cable Vault 0/76~

U2 Aux. Cable Vault 0/6 Ul&2 ESV Basement Area 4/OC U2 ESV Pump & HCC Rms. 9/0 C System protects area common to both Units 1 and 2

• (x/y) x is number of Function A (early vaxning fixe detection and notification only) instruments.

y is number of Function B (actuati.on of fixe suppression systems and early varning and notification) instruments.

circuit contains both smoke and.flame detectors tvo circuits of five detectors each .

tvo circui.ts of 38 detectors each .

COOK NUCLEAR PLANT - UNIT 2 3/4 3-52 AH NDHENT Nn. $ X. ZZ$

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1 p 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 containina fuel shall be limited. in such a way as to precl'ude 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 requi rements 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 RE UIREMENTS 4.9.7.1 . Crane interlocks 'hich prevent crane travel with 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

'oads thereafter during crane operation.

4.9.7.2 The potential impact energy due to dropping the crane's load shall be determined to be c 24,240 in.-lbs. prior to moving each load over racks containing fuel.

• Shared system with D. C. COOK - UNIT D.~ C.~ COOK - UNIT2 3/4 9-7 Amendment'o. 5? 86 j

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REHJELING OPERATIONS SURVEILLANCE RE UIREHENTS Continued ~

3. Verifying thac the HEPA filter banks remove greater than or equal to 99'f the DOP when they az'e tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation syscem ac a flow z'ate of 30,000 cfm plus or minus 10%.

4 . Verifying within 31 days after removal chat a laboratory analysis of a carbon sample from either ac lease one test canister or ac lease cwo carbon samples removed from one oz the charcoal adsorbers demonscraces a removal efficiency of greacer chan or equal to 90% for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1980 (AS'QI D 3803-1979, 30 C95% R.H.). The carbon samples noc obcained from tesc canisters shall be prepared by either:

(a) Emptying one entire bed from a removed adsorbex tray, mixing the adsorbent thoroughly, and obcaining samples ac lease, cwo inches in cLiamecer and with a length equal co the thickness of che bed, or.

(b) Emptying a longitudinal sample from a adsorber tray, mixing the adsorbent thoroughly, and obtaining samples ac least two inches in diamecer and wich a lengch equal to che thickness of che bed.

f Subsequent to reins ailing the adsorber =ay used for obtaining the carbon sample, the system shall be demonscrated OPERA8LE by also verifying thac "he charcoal adsorbers remove greacer than or equal to 99% of a halogenaced hydrocarbon xef-igerant test gas when they are tested in-place in accordance with ANSI H510-1980 while operacing che ventilation system ac a flow race of ~086 Goo cfm plus or minus 10%.

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5. Verifying a system flow race of 30,000 cfm plus or minus 10% during system operation when tested in accordance with ANSI H510-1980.

o. Afcez'vexy 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal, adsorber operation by either:

l. Verifying within 31 days after removal that a laboracozy analysis oz a carbon sample obtained from a test canister demonscraces a removal efficiency of greater chan or equal to 90% for xadioact:ive mechyl iodide when the sample is tested in accordance with ANSI H510-1980 (ASTH D 3803-1979, 30 C, 95% R.H.)

COOK HUCLF~ PLANT - UNIT 2 3/4 9-13 Amendmenc Ho. QE, 1 40

\ ~ I RADIOACTIVE EFFLUENTS GASEOUS RADVASTE TREATKENT LYING CONDITION FOR OPERATION 3.11.2.4 The gaseous radvaste treatment system and the ventilation exhaust treatment system shall be used to reduce the radioactive materials in gaseous vaste prior to their discharge vhen the projected gaseous effluent air doses due to gaseous effluent releases to unrestricted azeas (See Figure 5.1g3) vhen averaged over 31 days, vould exceed 0.2 mrad for gamma radiation and 0.4 mrad foz beta radiation. The ventilation exhaust treatment system shall be used to zeduce radioactive materials in gaseous vaste prior to their discharge vhen the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) vhen averaged over 31 days vould exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTXON:

a ~ Pith gaseous waste being dischazged vithout treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Repozt vhich includes the folloving information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperabi.lizy.

2. Action(s) taken to restore the inoperable equipment to operable status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UZRE.'KITS 4.11.2.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance vith the ODQf, whenever the gaseous vaste treatment system or ventilation exhaust treatment system is not operational.

COOR NUCLEAR PLANT - UNIT 2 3/4 11-12

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VOLUME 5.4.2 The total water and steam volume of the reactor coolant 0 system is 12,612 plus or minus 100 cubic feet at a nominal Tavg of 70 F.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-X.

5.6 FUEL STORAGE CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

ff equivalent A Keff to less than 0.95 when flooded with unborated water,

b. A nominal 8.97-inch center-to-center distance between fuel assemblies, placed in the storage racks.

C. The fuel assemblies will be classified as acceptable for Region 1, Region 2, or Region 3 storage based upon their assembly average burnup versus initial nominal Cells acceptable for Region 1, Region 2, and

'nrichment.

Region 3 assembly storage are indicated in Figures 5.6-1 and 5.6-2. Assemblies that are acceptable for storage in Region 1, Region 2, and Region 3 must meet the design criteria that define the regions as follows:

1. Region 1 is designed to accommodate new fuel with a maximum nominal enrichment of 4.95 wt% U-235, or spent fuel regardless of the discharge fuel burnup.

2. Region 2 is designed to accommodate fuel of 4.95%

initi,al nominal enrichment burned to at least 50,000 MWD/MTU, or fuel of other enrichments with equivalent reactivity.

3. Region 3 is designed to accommodate fuel of 4.95%

initial nominal enrichment burned to at least 38,000 MWD/MTU, or fuel of other enrichments with equivalent reactivity.

COOK NUCLEAR PLANT - UNIT 2 5-5 AMENDMENT NO. SS, ZW, Z2Z ZR7, 152

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ADMINISTRATIVE CONTROLS

6. 1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible fox overall facility operation and shall delegate in vziting the succession to this responsIbIlity during his absence.

6.1.2 The Shift Supexvisor (or during his absence fx'om the control room complex, a designated indivt.dual) shall be responsible for the control zoom

' management directive to this effect signed by the Vice command function.

President - Nuclear Operations shall be reissued to all station pezsannel on an annual basis.

6.2 ORGANIZATION ONSZTE AND OFFSITE ORGANIZATIONS 6.2.1. Onsite and offsite organizations shall be established fox'nit operatIon and corporate management, respectively. The onsIte and, offsite organizatI.ons shall include the posI.tions for activities affecting the safety of the nuclear paver plant.

a. Lines of authority, responsibility, and communication shall be established and defined foz the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appzopriate, in the form of organizational charts. These organizational charts vill be documented in the%8 gn updated tJ FiPt-in accordance vith 10 CFR 50.71(e).

b. The Plant Manager shall be responsible for overall unI,t safe operation and shall have control ovex those onsite actIvities necessary'or safe operation and maintenance of the plant.

C ~ The Vice PresI.dent - Nuclear OperatIons shall have corporate responsibility for overall plane, nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in opexating, maintaining, and providIng technical support to the plant to ensuxe nuclear safety.

d. The individuals vho train the operating staff and those vho carry out health physics and quality assurance functions may report to the appropriate onsite manager; hovevex, they shall have suffIcient organizatIonal fxeedom to ensuze their independence from operating pressures.

FACILITY STAFF 6.2.2 The FacI.lity organization shall be sub]ect to the follovtng:

a. Each on duty shift shall be composed of at least the minimum shift crev composition shovn in Table 6.2-.1.

COOK NUCLEAR PLANT - UNIT 2 6-1 AMENDMENT NO.P jt,gl7 s138

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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.}.-1971 for comparable positions, except for (1) the Plant Radiation Protection Manager, vho shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (2) the Shift Technical Advisor, vho. shall have a bachelor's degree or equivalent in a scientific oz engineering discipline vith specific training in plant design, and response and analysis of the plant for transients and accidents and, (3) the Operations Superintendent, vho must hold or have held a Senior Operator License as specified in Section 6.2.2.h.

6.4 TRAXNZNG 6.4.1 A retraining and replacement training program for the facility staff shall be maintained under tha direction of the Training Managez and shall meet oz l exceed the requirements and recommendations of Section 5.5 of ANSI N18.1.-1971 and 6.5 REVIEW AND AUDIT 6.5.1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE PNSRC FUNCTION 6.5,1.1 Tha PNSRC shall function to advise the Plant

~ ~ ~ Manager on all matters related to nuclear'afety.

COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Depaztment Superintendents, or supervisory personnel reporting directly to the Plant Manager, Assistant Plant Managers or Department Superintendents from the functional areas listed belov:

Licensing Activities Technical'uppozt Safety 6 Assessment Radiation Protection Operations Maintenance The Chairman, his alternate and other members and their alternates of the PNSRC shall ba designated by the Plant Manager. In addition to the Chairman; the PNSRC membership shall consist of one individual from each of tha areas designated above.

PNSRC members and alternates shall meet or exceed the minimum qualifications of ANSX N18.1-1971 Section 4.4 for comparable positions. The nuclear pover plant operations individual shall meet the qualif ications of Section 4. 2. 2 of ANSI N18.1-1971 except for. the requirement to hold a current Senior Operator License.

Tha operations individual must hold or have held a Senior Operator License at Cook Nuclear Plant or a similar reactor. The maintenance individual shall meet the qualif'cations of Section 4.2.3 of ANSI N18.1-1971.

COOK NUCLEAR PLANT - UNIT 2 6-4 AMENDMENT NO. )P,f77,138

INSTRUMEHTATION BASES 3 4 . 3 . 3 . 6 POST ACCIDE iT INSTRUMENTATION~

The OPKQILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

3 4.3.3.7 available o rement of the neutron flux spatial di '4 on within the reactor core. bility is reqlxiz o 1) monitor the core flux patterns that are represe he peak core power density and 2) limit the core r axial power profile su the total power peakin " . is maintained within acceptable m'ts.

3 4.3.3.8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures warning capability is available for the prompt detection of fires.

that'dequate This capability is required in order to detect and locate fires in their early stages. Prompt de'tection of fires will z'educe the potential for damage to safety-related equipment and is an integral element in the ovez'all facility fire protection program.

In the event that a portion of the f'e detection instrumentation is inoperable the establishment of frequent fire patrols in the affected

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areas is required to provide detection capability until the inoperable instrumentation is zesrored to OPERABILITY. Use of containment temperature monitoring is allowed once per hour inoperable.

if containment fize detecrion is 3 4.3.3.9 RADIOACTIVE L'I UID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provide to monitor and control, as applicable, the release of radioactive material in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approval methods in the'ODCM to ensure that the alarm/trip will occur prioz'o exceeding the limits of 10 CFR Part 20. The OPER.-'VILITY and use of th's instrumentation 'is consistent with the z'equiremen"s of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

D. C. COOK - UNIT 2 8 3/4 3-3 Anendment No. 61,115,119

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'EFUELING'OPERATIONS BASES 3 4.9.9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will automatically isolated upon detection of high radiation be levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

3 4.9.10 AND 3 4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99& of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.'ater level above the vessel flange in MODE 6 will vary as the reactor vessel head and the system internals are removed. The 23 feet of water are required before any subsequent movement of fuel assemblies or control rods.

3 4.9.12 STORAGE POOL VENTILATION SYSTEM The limitations o'n the storage pool ventilation system ensure that'll radioactive material released from an irradiated = el assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptio..s of the accident analyses.

The 1980 version of ANSI N510 is used as a testing guide. This standard, however, is intended to be rigorously applied only to systems which, unlike the storage pool ventilation system, are designed to ANSI N509 standards. For the specific case of the air-aerosol mixing uniformity test required by ANSI N510 as a prerequisite to in-place leak testing of charcoal and HEPA filters, the air-aerosol uniform. mixing test acceptance criteria were not rigorously met. For this reason, a statistical correction factor will be applied to applicable surveillance test results where required.

In order to maintain the minimum negative pressure required by Technical Specifications (1/8 inch W.G.) during movement of fuel within the storage pool or during crane operation with loads over the pool, the crane bay roll-up door and the drumming room roll-up door, located on the 609-foot elevation of the auxiliary building, must be closed. However, they may be opened during these operations under administrative control. if th'e crane bay door needs to be opened during fuel movement, an example of an administrative control might be to station an individual at the door who would )e j.n communication with personnel in the spent fuel pool area and could~ma- C. a)S the door when passage, was comple'ted or in the event of an emergency. For the drumming room door, an example of an administrative control might be to require the door to be reclosed after normal ingress and egress of personnel or materi.aJ,. or to station an individual at the door extended period of time.

if the door needs to remain open for an D. C. COOK - UNIT 2 B 3/4 9-3 Amendment No. $ g, 111

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ATTACHMENT 3 TO AEP:NRC:1137B PROPOSED REVISED TECHNICAL SPECIFICATION PAGES

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TABLE 3 3-1 Continued DESIGNATION CONDITION AND SETPOINT FUNCTION P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats Flux Channels greater than or the automatic block of equal to 11% of RATED reactor trip on: Low or 1 of 2 Turbine First flow in more than one THERMAL'OWER Stage Pressure channels greater primary coolant loop, than or equal to 37 psig. reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressur-izer high level. Low flow in a par'ticular loop can be evidenced by either a detected low flow or by the opening of the reactor coolant pump breaker.

P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats Flux channels greater than or the automatic block of equal to 31% of RATED THERMAL reactor trip caused by POWER a low coolant flow condition in a single loop.

With 3 of 4 Power range neutron P-10 prevents or flux channels less than 9% of defeats the manual RATED THERMAL POWER. block of: Power range low setpoint reactor trip, Intermediate range reactor trip, and intermediate range rod stops.

Provides input to P-7.

COOK NUCLEAR PLANT - UNIT 1 3/4 3-9 AMENDMENT NO. 85, 4QO

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TABLE 3 3-10 (Continued)

Unit 1 and Common Area Fire Detection S stems Total Number Detector S stem Location of Detectors Heat ~lame Smoke (x/y)* (x/y)* (~)*

Ul Cable Tunnels a) Quad 1 Cable Tunnel 0/3 0/4 b) Quad 2 Cable Tunnel 0/4 0/7 c) Quad 3N 0/3 0/4 d) Quad 3S 0/3 0/3 e) Quad 3M 0/3 0/4 f) Quad 4 0/5 0/6 Ul Charcoal Filter Ventilation Units a) 1-HV-AES-1 0/] *****

b) 1-HV-AES-2 P/1*****

c) 1-HV-ACRF P/1*****

d) 1-HV-CIPX 0/].**4k%

e) 1-HV-CPR 0/1 **+4*

f) 12-HV-AFX 0/] **~C Ul Containment******

a) RCP 1 1/0 b) RCP 2 1/0 c) RCP 3 1/0 d) RCP 4 1/0 e) Cable Trays 1/0 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 notificati'on) instruments.

Originally installed to automatically deluge charcoal filters.

However, manual actions are now necessary.

The fire detection instruments located within the Containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate tests, 'I Thermistors located within cable trays which'ontain combustible cables, in both upper and lower containment throughout quadrants 1-4.

COOK NUCLEAR PLANT - UNIT 1 3/4 3-53a AMENDMENT NO. 430,

TABLE 3.3-11 POST-ACCIDENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTRUMENT OPERABLE 1~ Containment Pressure 2 2~ Reactor Coolant Outlet Temperature- TDT (Wide Range) 2 3 ~ Reactor Coolant Inlet Temperature- T~i~ (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 1**

12. PORV Position Indicator Limit Switches*** 1/Valve PORV Block Valve Position Indicator Limit Switches 1/Valve Safety Valve Position Indicator Acoustic Monitor 13

'4.

1/Valve

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

16. Reactor Coolant Inventory Tracking System One Train (Reactor Vessel Level Indication) (3 Channels/Train) 17 Containment Sump Level 1

'8.

Containment Water Level 2

• 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.

COOK NUCLEAR PLANT UNIT 1 3/4 3-55 Amendment No. ~, 4k&, 46k, 468

TABLE 4 3-POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION

1. Containment Pressure

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

3. Reactor Coolant Inlet Temperature-(Wide Range ) M R TCO

4. ReaRc or Coolant Pressure-Wide Range M R

5. Pressurizer Water Level M R

6. Steam Line Pressure M R 7, Steam Generator Water Level-Narrow Range M R 8 ~ RWST Water Level M R

9. Boric Acid Tank Solution Level M R 10.Auxiliary Feedwater Flow Rate M R 11.Reactor Coolant System Subcooling M R Margin Monitor 12.PORV Position Indicator - Limit Switches M R 13.PORV Block Valve Position Indicator- M R Limit Switches 14.Safety Valve Position Indicator-Acoustic Monitor 15.Incore Thermocouples (Core Exit Thermocouples) M R(1) 16.Reactor Coolant Inventory Tracking System M(2) R(3)

(Reactor Vessel Level Indication) 17.Containment Sump Level M R 18.Containment Water Level M R (1) Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

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

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

COOK NUCLEAR PLANT - UNIT 1 3/4 3-56 AMENDMENT NO. 54) 444

TABLE 3.7-6 LOW PRESSURE CARBON DIOXIDE SYSTEMS LOCATION ACTUATION PERIOD Diesel Generator 1AB Room Cross-zoned Heat Diesel Generator 1CD Room Cross-zoned Heat Diesel Generator Fuel Oil Pump Room Heat 4 KV Switchgear Rooms Manual Control Rod Drive, Transf. Switchgear Rooms Manual Engineered Safety Switchgear Room Manual Switchgear Room Cable Vault Cross-zoned Ionization and Infrared Auxiliary Cable Vault Ionization Control Room Cable Vault (Backup)* Manual Penetration Cable Tunnel Quadrant 1 Manual Penetration Cable Tunnel Quadrant 2 Manual Penetration Cable Tunnel Quadrant 3N Manual Penetration Cable Tunnel Quadrant 3M Manual Penetration Cable Tunnel Quadrant 3S Manual Penetration Cable Tunnel Quadrant 4 Manual

• Control Room Cable Vault CO> System is only required to be operable when the Cable Vault Halon System is inoperable.

COOK NUCLEAR PLANT - UNIT 1 3/4 7;41 AMENDMENT NO. M, 430,

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.

~GTION:

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 s 24,240 in.-lbs. prior to moving each load over racks containing fuel.

• Shared system with Cook Nuclear Plant - Unit 2.

COOK NUCLEAR PLANT - UNIT 1 3/4 9-8 AMENDMENT NO.

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RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The gaseous radwaste treatment system and the ventilation exhaust treatment system shall be used to reduce the radioactive materials in gaseous waste prior to their dischaige when the projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas (gee Figure 5. 1-3) when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad for beta radiation. The ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION:

a. With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability.

2. Action(s) taken to restore the inoperable equipment to operable status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.11.2.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the ODCM, whenever the gaseous waste treatment system or ventilation exhaust treatment system is not operational.

COOK NUCLEAR PLANT - UNIT 1 3/4 11-12 AMENDMENT NO. 60, 454

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DESIGN FEATURES CAPACITY 5.6.4 The fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3613 fuel assemblies.

5 7 SEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I Items in the FSAR shall be designed and maintained to the original design provisions contained in the FSAR with allowance for normal degradation pursuant to the applicant Surveillance Requirements.

5 8 METEROLOGICAL TOWER LOCATION 5.8.1 The meterological tower shall be located as shown on Figure 5.1-3.

5 9 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.9.1 The components identified in Table 5.9-1 are designed and shall be maintained within the cyclic or transient limits of Table 5.9-1.

COOK NUCLEAR PLANT - UNIT 1 5-9 AMENDMENT NO. 40>

L I ADMINISTRATIVE CONTROLS 6 1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor (or during his absence from the control room complex, a designated individual) shall be responsible for the control room command function. A management directive to this effect signed by the Vice President - Nuclear Operations shall be reissued to all station personnel on an annual basis.

6 2 ORGANIZATION ONSITE AND OFFSITE ORGANIZATIONS 6.2.1 Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined for the highest management level through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.71(e).

b. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

co Vice President - Nuclear Operations shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in d.'he operating, maintaining, and providing plant to ensure nuclear safety.

technical support to the The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

FACILITY STAFF 6.2.2 The Facility organization shall be subject to the following:

a. Each on duty, shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

COOK NUCLEAR PLANT - UNIT 1 6-1 AMENDMENT NO.

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 Plant Radiation Protection Manager, who shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (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 and, (3) the Operations Superintendent must hold or have held a Senior Operator License as specified in Section 6.2.2.h.

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 Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 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.

COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Department Superintendents, or supervisory personnel reporting directly to the Plant Manager, Assistant Plant Managers or Department Superintendents from the functional areas listed below:

Licensing Activities Technical Support Safety & Assessment Radiation Protection Operations Maintenance The Chairman, his alternate and other members and their alternates of the PNSRC shall be designated by the Plant Manager. In addition to the Chairman, the PNSRC membership shall consist of one individual from each of the areas designated above.

PNSRC members and alternates shall meet or exceed the minimum qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions. The nuclear power plant operations individual shall me'et the qualifications of Section 4.2.2 of ANSI N18.1-1971 except for the requirement to hold a current Senior Operator License.

The operations individual must hold or have held a Senior Operator License at Cook Nuclear Plant or a similar reactor. The maintenance individual shall meet the qualifications of Section 4:2.3 of ANSI N18.1-1971.

COOK NUCLEAR PLANT - UNIT 1 6-4 AMENDMENT NO. 49,

INSTRUMENTATION BASES 3 4 3 3 5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

3 4 3 3 5 1 APPENDIX R REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the Appendix R remote shutdown instrumentation ensures that sufficient instrumentation is available to permit shutdown of the facility to COLD SHUTDOWN conditions at the local shutdown indication (LSI) panel. In the event of a fire, normal power to the LSI panels may be lost. As a result, capability to repair the LSI panels from Unit 2 has been provided. If the alternate power supply is not available, fire watches will be established in those fire areas where loss of normal power to the LSI panels could occur in the event of fire. This will consist of either establishing continuous fire watches or verifying OPERABILITY of fire detectors per Specification 4.3.3.7 and establishing hourly fire watches. The details of how these fire watches are to be implemented are included in a plant procedure.

3 4 3 3 FIRE DETECTION INSTRUMENTATION SYSTEMS DETECTORS OPERABILITY of the fire detection systems/detectors ensures that adequate detection capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages.

Prompt detection of the fires will reduce the potential for damage to safety related systems or components in the areas of the specified systems and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection systems is inoperable, the ACTION statements provided maintain the facility's fire protection program and allows for continued operation of the facility until the inoperable system(s)/detector(s) are restored to OPERABILITY. However, it is not our intent to rely upon the compensatory action for an extended period of time and action will be taken to restore the minimum number of detectors to OPERABLE status within a reasonable period.

3 4 3 3 8 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

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

REFUELING OPERATIONS BASES 3 4 9 10 and 3 4 9 ll WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10$ iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis, Water level above the vessel flange in MODE 6 will vary as the reactor vessel head and the system internals are removed. The 23 feet of water are required before any subsequent movement of fuel assemblies or control rods.

3 4 9 12 STORAGE POOL VENTILATION SYSTEM The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

The 1980 version of ANSI N510 is used as a testing guide. This standard, however, is intended to be rigorously applied only to systems which, unlike the storage pool ventilation system, are designed to ANSI N509 standards. For the specific case of the air-aerosol mixing uniformity test required by ANSI N510 as a prerequisite to in-place leak testing of charcoal and HEPA filters, the air-aerosol uniform mixing test acceptance criteria were not rigorously met.

For this reason, a statistical correction factor will be applied to applicable surveillance test results where required.

In order to maintain the minimum negative pressure required by Technical Specifications (1/8 inch W.G.) during movement of fuel within the storage pool or during crane operation with loads over the pool, the crane bay roll-up door and the drumming room roll-up door, located on the 609-foot elevation of the auxiliary building, must be closed. However, they may be opened during these operations under administrative control. If the crane bay door needs to be opened during fuel movement, an example of an administrative control might be to station an individual at the door who would be in communication with personnel in the spent fuel pool area and could close the door when passage through the door was completed or in the event of an emergency. For the drumming room door, an example of an administrative control might be to require the door to be reclosed after normal ingress and egress of personnel or material, or to station an individual at the door if the door needs to remain open for an extended period of time.

Should the doors become blocked or stuck open while under administrative control, Technical Specification requirements will not be considered to be violated provided the Action Statement requirements of Specification 3.9.12 are expeditiously followed, i.e., movement of fuel within the storage pool or crane operation with loads over the pool is expeditiously suspended.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 9-3 AMENDMENT NO.

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

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

17. Underfrecpxency-Reactor 4-1/bus Coolant Pumps

18. Turbine Trip A. Low Fluid Oil Pressure B. Turbine Stop Valve Closure

19. Safety Injection Input 1I 2 from ESF

20. 'Reactor Coolant Pump Breaker Position Trip Above P-7 1/breaker 1/breaker per operating loop

21. Reactor Trip Breakers 1I 2/ 1 33 3* 4* 5* 14

22. Automatic Trip Logic 1I 2I 1 3* 4* 5* 14 COOK NUCLEAR PLANT UNIT 2 3/4 3-4 AMENDMENT NO 86,

~ TABLE 3 3-1 Continued DESIGNATION CONDITION AND SETPOINT FUNCTION P-7 With 2 of 4 Power Range Neutron P-7 prevents or defeats Flux Channels ~ 11$ of RATED the automatic block of THERMAL POWER or 1 of 2 Pressure reactor trip on: Low before the First Stage flow in more than one channels ~ 51 psig. primary coolant loop, reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low pressure, and pressurizer high level.

Low flow in a particular loop can be evidenced by either a detected low flow or by the opening of the reactor coolant pump breaker.

P-8 With 2 of 4 Power Range Neutron P-8 prevents or defeats Flux channels E 31% of RATED the automatic block of THERMAL POWER. reactor trip caused by a low coolant flow condition in a single loop.

P-10 With 3 of 4 Power Range Neutron P-10 prevents or defeats flux channels < 9% of RATED the manual block of:

THERMAL POWER.. Power range low setpoint reactor trip, Inter-mediate range reactor trip, and intermediate range rod stops.

Provides input to P-7.

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

TABLE 3.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM CHANNELS OPERABLE 1~ Containment Pressure 2 2 Reactor Coolant Outlet Temperature - THpT (Wide Range) 2 Reactor Coolant Inlet Temperature - Tp~i~ (Wide Range)

~

3 ~ 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 (Reactor Vessel Level Indication) '(3 Channels/Train)

17. Containment Sump Level 1

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.

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 Indicator Limit Switches instruments.

1 COOK NUCLEAR PLANT UNIT 2 3/4 3-46 Amendment. No. 9&, 9S, ~, ~

TABLE 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHANNEL CHECK CAIZHfQTION 1~ Containment Pressure M R 2 ~ Reactor Coolant Outlet Temperature T~~ (Wide Range) M R 3 ~ Reactor Coolant Inlet Temperature T~gp (Wide Range) M R 4 ~ Reactor Coolant Pressure Wide Range M R

5. Pressurizer Water Level M R

6. Steam Line Pressure M R 7~ Steam Generator Water Level Narrow Range M R 8~ RWST Water Level M R 9 ~ Boric Acid Tank Solution Level M R

10. Auxiliary Feedwater Flow Rate M R 11 Reactor Coolant System Subcooling Margin Monitor M R

12. PORV Position Indicator Limit Switches M R

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

14. Safety Valve Position Indicator - Acoustic Monitor M R

15. Incore Thermocouples (Core Exit Thermocouples) M R(1) 16 Reactor Coolant Inventory Tracking System M(2) R(3)

'7 (Reactor Vessel Level Indication)

~ Containment Sump Level M R 18 ~ Containment Water Level M R Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

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

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

COOK NUCLEAR PLANT UNIT 2 3/4 3-47 AMENDMENT NO. 9&, 9S,

TABLE 3 3-11 Unit 2 and Common Area Fire Detection S stems Total Number Detection S stem Location of Detectors Heat ~lame Smoke (x/y)* (x/y)* (x/y)*

Auxiliary Building a) Elevation 573 23/0 C b) Elevation 587 55/0 C c) Elevation 609 41/0 C d) Elevation 633 41/0 C e) Elevation 650 34/0 C f) New Fuel STGE Area 4/0 C U2 East Main Steam Valve Enclosure 28/0**

U2 Main Steam Line Area El. 612 (Around Containment) 13/0**

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 6 AB Bttry. Rms. 0/5 0/17 U2 Pressurizer Heater XFMR. Rm. 12/0 U2 Diesel Fuel Oil Transfer Pump Rm. 0/1 U2 Diesel Generator Rm. 2AB 0/2 U2 Diesel Generator Rm. 2CD 0/2 U2 Diesel Generator Ramp Corr. 4/0 Ul&2 AFWP Vestibule 2/0 C U2 Control Room 42/0 U2 Switchgear Cable Vault 0/10*~ 0/13 U2 Control Rm. Cable Vault 0/76*%%*

U2 Aux. Cable Vault 0/6 Ul&2 ESW Basement Area 4/0 C U2 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 Funct:ion B '(actuation of fire suppression systems warning and notification) instruments.

and'arly circuit contains both smoke and flame detectors

• • • two circuit:s of five detectors each

~** two circuits of 38 detectors each COOK NUCLEAR PLANT - UNIT 2 3/4 3-52 AMENDMENT NO. 44,

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'hey 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.

S VEILLANCE 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 s 24,240 in.-lbs. prior to moving each load over racks containing fuel.

• Shared system with Cook Nuclear Plant - Unit 1.

COOK NUCLEAR PLANT - UNIT 2 3/4 9-7 AMENDMENT NO. M, 96

~ ~ ~

REFUELING OPERATIONS SURVEILLANCE RE UIREMENTS Continued

3. Verifying that the HEPA filter banks remove greater than or equal to 99%

of the DOP when they are tested in-place in accordance with ANSI N510-1980 while operating the exhaust ventilation system at a flow rate of 30,000 cfm plus or minus 108.

4. Verifying within 31 days after removal that a laboratory analysis of a carbon sample from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers demonstrates a removal efficiency of greater than or equal to 90$ for radioactive methyl iodide when the sample is tested in accordance with ANSI N510-1980 (ASTM D 3803-1979, 30 C, 95% R.H.). The carbon samples not obtained from test canisters shall be prepared by either:

(a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or (b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

Subsequent to reinstalling the adsorber tray used for obtaining the carbon sample, the system shall be demonstrated OPERABLE by also verifying that the charcoal adsorbers remove greater than or equal to 99% of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with ANSI N510-1980 while operating the ventilation system at a flow rate of 30,000 cfm plus or minus 10%.

5. Verifying a system flow rate of 30,000 cfm plus or minus 108 during system operation when tested in accordance with ANSI N510-1980.

C. After every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of charcoal adsorber operation by either:

1. Verifying within 31 days after removal that a laboratory analysis of a carbon sample obtained from a test canister demonstrates a removal efficiency of greater than or equal to 90$ for radioactive methyl iodide when the sample is'ested in accordance with ANSI N510-1980 (ASTM D 3803-1979, 30 C, 95',R.H.).

COOK NUCLEAR PLANT - UNIT 2 3/4 9-13 AMENDMENT NO. ~, 440

RADIOACTIVE EFFLUENTS GASEOUS RADWASTE TREATMENT LIMITING CONDITION FOR OPERATION 3.11.2.4 The gaseous radwaste treatment system and the ventilation exhaust treatment system shall be used to reduce the radioactive materials in gaseous waste prior to their discharge when the projected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days, would exceed 0.2 mrad for gamma radiation and 0.4 mrad.

for beta radiation. The ventilation exhaust treatment system shall be used to reduce radioactive materials in gaseous waste prior to their discharge when the projected doses due to gaseous effluent releases to unrestricted areas (See Figure 5.1-3) when averaged over 31 days would exceed 0.3 mrem to any organ.

APPLICABILITY: At all times.

ACTION'.

With gaseous waste being discharged without treatment and in excess of the above limits, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report which includes the following information:

1. Identification of the inoperable equipment or subsystems and the reason for inoperability.

2 ~ Action(s) taken to restore the inoperable equipment to operable status.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.11 '.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the ODCM, whenever the gaseous waste treatment system or ventilation exhaust treatment system is not operational.

COOK NUCLEAR PLANT - UNIT 2 3/4 11-12 AMENDMENT NO. M, 438

C ~

A C

~ 4 S v VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 12,612 plus or minus 100 cubic feet at a nominal Tavg of 70 F.

5.5 METEOROLOGICAL TOWER LOCATION meteorological tower shall be located as Figure 5.1-3.

Bl The shown on CRITICALITY - SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. A K,qq equivalent to less than 0.95 when flooded with unborated water,

b. A nominal 8.97-inch center-to-center distance between fuel assemblies, placed in the storage racks.

c ~ The fuel assemblies will be classified as acceptable for Region 1, Region 2, or Region 3 storage based upon their assembly burnup versus initial nominal enrichment. Cells acceptable for Region 1, Region 2, and Region 3 assembly storage are indicated in Figures 5.6-1 and 5.6-2. Assemblies that are acceptable for storage in Region 1, Region 2, and Region 3 must meet the design criteria that define the regions as follows:

1. Region 1 is designed to accommodate new fuel with a maximum nominal enrichment of 4.95 wtS U-235, or spent fuel regardless of the discharge fuel burnup.

2. Region 2 is designed to accommodate fuel of 4.95% initial nominal enrichment burned to at least 50,000 MWD/MTU, or fuel of other enrichments with equivalent reactivity,

3. Region 3 is designed to accommodate fuel of 4.95% initial nominal enrichment burned to at least 38,000 MWD/MTU, or fuel of other enrichments with equivalent reactivity.

COOK NUCLEAR PLANT - UNIT 2 5-5 AMENDMENT NO. 55, 404,

DMINISTRATIVE CONTROLS 6 1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

6.1.2 The Shift Supervisor (or during his absence from the control room complex, a designated individual) shall be responsible for the control room command function. A management directive to this effect signed by the Vice President - Nuclear Operations shall be reissued to all station personnel on an annual basis.

6 2 ORGANIZATION ONSITE AND OFFSITE ORGANIZATIONS 6.2.1 Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the nuclear power plant.

a. Lines of authority, responsibility, and communication shall be established and defined for the highest management level through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organizational charts. These organizational charts will be documented in the UFSAR and updated in accordance with 10 CFR 50.71(e).

b. The Plant Manager shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

C. The Vice President - Nuclear Operations shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and pioviding technical support to the plant to ensure nuclear safety.

d. The individuals who train the dperating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager;'owever, they 'shall have sufficient organizational freedom to ensure their independence from operating pressures.

FACILITY STAFF 6.2.2 The Facility organization shall be subject to the following:

a. Each on duty shift shall be composed of at least the minimum shift crew composition shown in -Table 6.2-1.

COOK NUCLEAR PLANT - UNIT 2 6-1 AMENDMENT NO. 58,

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 Plant Radiation Protection Manager, who shall meet or exceed qualifications of Regulatory Guide 1.8, September 1975, (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 and, (3) the Operations Superintendent, who must hold or have held a Senior Operator License as specified in Section 6.2.2.h.

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 Manager and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 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.

COMPOSITION 6.5.1.2 The PNSRC shall be composed of Assistant Plant Managers, Department Superintendents, or supervisory personnel reporting directly to the Plant Manager, Assistant Plant Managers or Department Superintendents from the functional areas listed below:

Licensing Activities Technical Support Safety & Assessment Radiation Protection Operations Maintenance The Chairman, his alternate and other members and their alternates of, the PNSRC shall be designated by the Plant Manager. In addition to the Chairman, the PNSRC membership shall consist of one individual from each of the areas designated above, PNSRC members and alternates shall meet or exceed the minimum qualifications of ANSI N18.1-1971 Section 4.4 for comparable positions. The nuclear power plant operations individual shall meet the qualifications of Section 4.2.2 of ANSI N18.1-1971 except for the requirement to hold a current Senior Operator License.

The operations individual must hold or have held a Senior Operator License at Cook Nuclear Plant or a similar reactor. The maintenance individual shall meet the qualifications of Section 4.2.3 of ANSI N18.1-1971..

COOK NUCLEAR PLANT - UNIT 2 6-4 AMENDMENT NO.

INSTRUMENTATION BASES 3 4 3 3 6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and following an accident.

3 4 3 3 DELETED 3 4 3 3 8 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for, the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. Use of containment temperature monitoring is allowed once per hour if containment fire detection is inoperable.

3 4 3 3 9 RADIOACTIVE LI UID EFFLUENT INSTRUMENTATION The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the release of radioactive material in liquid effluents during actual or potential releases. The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approval methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11.3 of the Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 3-3 AMENDMENT NO. 44,

REFUELING OPERATIONS BASES 3 4 9 9 CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment vent and purge penetrations will be automatically isolated upon detection of high radiation levels within the containment. The OPERABILITY of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

3 4 9 10 AND 3 4 9 11 WATER LEVEL - REACTOR VESSEL AND STORAGE POO The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis. Water level above the vessel flange in MODE 6 will.,vary as the reactor vessel head and the system internals are removed. The 23 feet of water are required before any subsequent movement of fuel assemblies or control rods.

3 4 9 12 STORAGE POOL VENTILATION SYSTEM The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.

The 1980 version of ANSI N510 is used as a testing guide. This standard, however, is intended to be rigorously applied only to systems which, unlike the storage pool ventilation system, are designed to ANSI N509 standards, For the specific case of the air-aerosol mixing uniformity test required by ANSI N510 as a prerequisite to in-place leak testing of charcoal and HEPA filters, the air-aerosol uniform mixing test acceptance criteria .were not rigorously met.

For this reason, a statistical correction factor will be applied to applicable surveillance test results where required.

In order to maintain the minimum negative pressure required by Technical Specifications (1/8 inch W.G.) during movement of fuel within the storage pool or during crane operation with loads over the pool, the crane bay roll-up door and the drumming room roll-up door, located on the 609-foot elevation of the auxiliary building, must be closed. However, they may be opened during these operations under administrative control. If the crane bay door needs to be opened during fuel movement, an example of an administrative control might be to station an individual at the door who would be in communication with personnel in the spent fuel pool area and could close the door when passage through the door was completed or in the event of an emergency. For the drumming room door, an example of an administrative control might be to require the door to be reclosed after normal ingress and egress of personnel or material, or to station an individual at the door if the door needs to remain open for an extended period of time.

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

V I ~ ~

~ b