ML17332A492

Amends 186 & 172 to Licenses DPR-58 & DPR-74,respectively, Re Administrative & Editorial Changes to TS

ML17332A492
Person / Time
Site:	Cook
Issue date:	12/30/1994
From:	Hannon J Office of Nuclear Reactor Regulation
To:
Shared Package
ML17332A493	List: ML17332A492 ML17334B522
References
NUDOCS 9501090346
Download: ML17332A492 (49)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 205RH)001 NDIANA MICHIGAN POWER COMPANY OCKET NO. 50-315 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 MENDMENT TO FACILITY OPERAT NG LICENSE Amendment No.

186 License No. DPR-58 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated November 15,

1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I;

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9501090346 941230 PDR ADOCK 050003l5 P

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-58 is hereby amended to read as follows:

Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised through Amendment No.

186, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 30, 1994 John N. Hannon, Director Project Directorate III-I Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

TTACHMENT TO LIC NS AMENDMENT NO. 186:

TO FACILITY OPERATING LICENSE NO. DPR-58 DOCKET NO. 50-3 5

Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

EMOVE 3/4 3-9 3/4 3-53a 3/4 3-55 3/4 3-56 3/4 7-41 3/4 9-8 3/4 11-12 B 3/4 3-6 B 3/4 9-3 5-2 5-9 6-1 6-4 NSERT 3/4 3-9 3/4'-53 a 3/4 3-55 3/4 3-56 3/4 7-41 3/4 9-8 3/4 11-12 B 3/4 3-6 B 3/4 9-3 5-2 5-9 6-1 6-4

LE 3 3-1 Continued SIGN TIO CO IT 0 ETP P-7 With 2 of 4 Power Range Neutron Flux Channels greater than or equal to 11% of RATED THERMAL POWER or 1 of 2 Turbine First Stage Pressure channels greater than or equal to 37 psig.

P-7 prevents or defeats the automatic block of reactor trip on:

Low flow in more than one primary coolant loop, reactor coolant pump under-voltage and under-frequency, turbine trip, pressurizer low

pressure, and pressur-izer 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 Flux channels greater than or equal to 31$ of RATED THERMAL POWER P-8 prevents or defeats the automatic block of reactor trip caused by a low coolant flow condition in a single loop.

P-10 With 3 of 4 Power range neutron flux channels less than 9% of RATED THERMAL POWER.

P-10 prevents or defeats the manual 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. 8S,

4QO, 386

r P ~

II 4

Ep

Unit and Common Area Pire Detectio stem etec o

ste cat o

Total Number Ul Cable Tunnels a)

Quad 1 Cable Tunnel b) Quad 2 Cable Tunnel c) Quad 3N d) Quad 3S e)

Quad 3M f) Quad 4 (x/y)*

Zhm (x/y)*

0/3 0/4 0/3 0/3 0/3 0/5

$amke,

(~gg)*

0/4 0/7 0/4 0/3 0/4 0/6 Ul Charcoal Filter Ventilation Units a) 1-HV-AES-1 b) 1-HV-AES-2 c) 1-HV-ACRF d) 1-HV-CIPX e) 1-HV-CPR f) 12-HV-AFX 0/1++4**

0/1++++4 0/]+*~

0/1***+%

0/1*****

P/1*~*C Ul Containment~~*

a)

RCP 1 b)

RCP 2 c)

RCP 3 d)

RCP 4 e) Cable Trays 1/0 1/0 1/0 1/0 58/p~~~

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

COOK NUCLEAR P1ANT - Li'.;:T 3/4 3-53a AMENDMENT NO. 480, ~,

186

-i POST-ACCIDENT MONITORING INSTRUMENTAT ON NSTRUME MINIMUM CHANNELS OPEIQBLg 1 ~

2 ~

3 ~

4o 5 ~

6.

7 ~

8 ~

9 ~

10.

11

'2

'3

'4.

15.

16

'7

'8

'ontainment Pressure Reactor Coolant Outlet Temperature-To~ (Wide Range)

Reactor Coolant Inlet Temperature>>

T~~ (Wide Range)

Reactor Coolant Pressure-Wide Range Pressurizer Water Level Steam Line Pressure Steam Generator Water Level-Narrow Range Refueling Water Storage Tank Water Level Boric Acid Tank Solution Level Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator Limit Switches**~

PORV Block Valve Position Indicator Limit Switches Safety Valve Position Indicator Acoustic Monitor Incore Thermocouples (Core Exit Thermocouples)

Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

Containment Sump Level Containment Water Level 2

2 2

2 2

2/steam generator 1/steam generator 2

1 1/steam generator*

1**

1/Valve 1/Valve 1/Valve 2/Core Quadrant One Train (3 Channels/Train) 1 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.

COOK NUCLEAR PLANT UNIT 1 3/4 3-55 Amendment No. 496, 4k&, 46%, 46$,

186

E 4 OST-CCIDENT MON TO ING I IO V

LANC E U REM M

M M

M M

M M

M M

M M

les)

M M(2)

M M

1. Containment Pressure

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

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

4.

ReaRc or Coolant Pressure-Wide Range

5. Pressurizer Water Level

6. Steam Line Pressure

7. Steam Generator Water Level-Narrow Range 8.

RWST Water Level

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

'17.Containment Sump Level 18.Containment Water Level R

R R

R R

R R

R R

R R

R R(l)

R(3)

R 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 NUCLEPK rLANT - UNIT 1 3/4 3-56 AMENDMENT NO. M, 444, 186

TABLE 3 7-6 W PRESSURE C RBO IOXIDE SYS EMS 7-ON CAP CITY CATION ION OD Diesel Generator 1AB Room Diesel Generator 1CD Room Diesel Generator Fuel Oil Pump Room 4 KV Switchgear Rooms Control Rod Drive, Transf.

Switchgear Rooms Engineered Safety Switchgear Room Switchgear Room Cable Vault Cross-zoned Heat Cross-zoned Heat Heat Manual Manual Manual Cross-zoned Ionization and Infrared Auxiliary Cable Vault Control Room Cable Vault (Backup)*

Penetration Cable Tunnel Quadrant 1

Penetration Cable Tunnel Quadrant 2

Penetration Cable Tunnel Quadrant 3N Penetration Cable Tunnel Quadrant 3M Penetration Cable Tunnel Quadrant 3S Penetration Cable Tunnel Quadrant 4

Ionization Manual Manual Manual Manual Manual Manual Manual

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

COOK NUCLEAR PLEX' d>IT 1 3/4 7-41 AMENDMENT NO. M, 430)

186,

4

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ELING OPERATIO S

TRAV SPENT FUEL STORAGE POOL 'SUILDI G*

I I G

CO TION FOR PE ON 3.9.7

'Loads in excess of 2,5DD 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.

~C~O 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.

URVEILLANCE RE UIREMENTS 4.9.7.1 Crane interlocks which prevent crane travel with loads in excess of 2,5OD 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 6

24,240 in.-lbs.

prior to moving each load over racks containing fuel.

• Sha.ed, system with Cook Nuclear Plant

- Unit 2.

COOK NUCLEAR PLANT - UNIT 1 3/4 9-8 AMENDMENT NO. 404 44k~

186

\\

IO CTIV E

UENTS GAS OUS RADWASTE TREATMENT TING CONDITION OPE ON 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 pro]ected gaseous effluent air doses due to gaseous effluent releases to unrestricted areas (gee Figure 5.1-3) uhen 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.

~CT~IO 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.

URVEILTANCE RE UIREHEN S

4.11.2.4 Doses due to gaseous releases to UNRESTRICTED AREAS shall be pro)ected 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 31-]2 AMENDMENT NO. m, ~

186

F" V$

TION ASES 3 4 EMOTE S

0 I

S UMENTATION 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 APPENDI REMOTE SHUTDOWN INSTR ENTA ON 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 IRE DETE ION INSTRUME T ON SYSTEMS ETECTO 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 facilityuntil 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 NSTRUMENT ION 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. W, 458, 186

E ELING OPERAT 0 S

AS S

3 4 9 10 and 3 4 9 ll WATER LEVEL -

EACTOR VESS L S 0 GE 00 The restri.ctions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap ectivity 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 ere required before any subsequent movement of fuel assemblies or control rods.

4 9 2

0 G

OOL VENT LA 0

The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly willbe 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 end HEPA filters, the air-aerosol uniform mixing test acceptance criteria were not rigorously mete 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 petsonnel in the spent fuel pool area and could close the doot when passage through the door was completed or i.n 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.ees 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. M ~

186

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COOK NUCLEAR PLANT UNIT AMENDMENT NO.

186

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ESIGN TURES gAPACI'~

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.

S CCLASSFCTO 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 applicable Surveillance Requirements.

5 8 METEROLOGICAL TOWER LOC T 0 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 4Q, 487., 440, 186

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INISTRA IVE CO 0

S 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 hi.s 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 AhD 0 FSITE 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.

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).

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 providing technical support to the plant to ensure nuclear safety.

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.

AC LITY ST FF 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. M, ~, ~,

186

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.4

D I STRATIVE CONT OLS 6

3 FACILITY ST FF 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 facilitystaff 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 REV EW 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 Safety 6 Assessment Operations

/

Technical Support Radiation Protection 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'he qualifications of, Section 4.2.3 of ANSI N18.1-1971.

I COOK NUCLEAR PLANT - UNIT 1 6-4 AMENDMENT NO. 40,

454, 186

rely,S REGS

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

172 License No.

DPR-74 l.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated November 15,

1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I'.

C.

D.

E.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No.

DPR-74 is hereby amended to read as follows:

Technical S ecifications The Technical Specifications contained in Appendices A and B,

as revised through Amendment No. i72, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of the date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

December 30, 1994 John N. Hannon, Director Project Directorate III-I Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

~ 0 I ~

TTACHMENT TO ICENSE AMENDMENT NO.

172 ACILITY OPERATING ICENSE NO. DPR-74 DOCKET NO. 50-316 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendmeni, number and contain vertical lines indicating the area of change.

~REMOV 3/4 3-4 3/4 3-8 3/4 3-46 3/4 3-47 3/4 3-52 3/4 9-7 3/4 9-13 3/4 11-12 B 3/4 3-3 B 3/4 9-3 5-2 5-5 6-1 6-4 INSERT 3/4 3-4 3/4 3-8 3/4 3-46 3/4 3-47 3/4 3-52 3/4 9-7 3/4 9-13 3/4 11-12 B 3/4 3-3 B 3/4 9-3 5-2 5-5 6-1 6-4

TABLE 3 3-1 Continued EACTOR TRIP SYSTEM INSTRUMENTATION NCTIONAL UNIT 16.

Undervoltage-Reactor Coolant Pumps TOTAL NO.

CHANNELS OF CHANNELS

~OTRIP 4-1/bus MINIMUM CHANNELS OPERABLE APPLICABLE MODES 17-.

Underfrequency-Reactor 4-1/bus Coolant Pumps 18.

Turbine Trip A.

Lov Fluid Oil Pressure B.

Turbine Stop Valve Closure 19.

Safety Injection Input from ESF 20.

'Reactor Coolant Pump Breaker Position Trip 11 2

SS~

6¹ Above P-7 21.

Reactor Trip Breakers 22.

Automatic Trip Logic 1/breaker 1/breaker per operating loop 1t 2r 3*

4*

5*

1 g 2

3*

4*

5*

-0 1,

33 14 1

14 COOK NUCLEAR PLANT - UNIT 2 3/4 3-4 AMENDMENT NO 86, +&V, %&V 172

TABLE 3 3-Co ti ued ESIGN TION EKED P-7 With 2 of 4 Power Range Neutron Flux Channels h 11% of RATED THERMAL POWER or 1 of 2 Pressure before the First Stage channels 2 51 psig.

P-7 prevents or defeats the automatic block of reactor trip on:

Low flow in more than one 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 Po~er Range Neutron Flux channels a

31% of RATED THERMAL POWER.

P-8 prevents or defeats the automatic block of reactor trip caused by a low coolant flow condition in a

single loop.

P-10 With 3 of 4 Power Range Neutron flux channels (

9% of RATED THERMAL POWER..

P-10 prevents or defeats the manual block of:

Power range low setpoint reactor

trip, Inter-mediate range reactor

trip, and intermediate range rod stops.

Provides input to P-7.

'. '0.'i NUCLEAR PLANT - UNIT 2 3/4 3-8 AMENDMENT NO. SR, ~,

172

TABLE 3 '-10 POST-ACCIDENT MONITORING INSTRUMENTAT ON INSTRUMENT MINIMUM CHANNELS OPERABLE 1.

2.

3 ~

4.

5.

6.

7.

8.

9 ~

10

'1.

12 ~

13 ~

14 ~

15

'6

'7

~

18

'ontainment Pressure Reactor Coolant Outlet Temperature - Toz (Wide Range)

Reactor Coolant Inlet Temperature - T~gy (Wide Range)

Reactor Coolant Pressure - Wide Range Pressurizer Water Level Steam Line Pressure Steam Generator Water Level Narrow Range Refueling Water Storage Tank Water Level Boric Acid Tank Solution Level Auxiliary Feedwater Flow Rate Reactor Coolant System subcooling Margin Monitor PORV Position Indicator - Limit Switches***

PORV Block Valve Position Indicator Limit Switches

.Safety Valve Position Indicator - Acoustic Monitor Incore Thermocouples (Core Exit Thermocouples)

Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

Containment Sump Level Containment Water Level 2

2 2

2 2

2/Steam Generator 1/Steam Generator 2

1 1/Steam Generator*

1**

1/Valve 1/Valve 1/Valve 2/Core Quadrant One Train (3 Channels/Train) 1 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.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-46 Amendment No. 8&, 9S, 44&, ~

172

0

\\

TABLE 4.3-10 OST ACCIDENT MONITOR NG INSTRUMENTATION SURVEILLANCE RE UIREMENTS CHANNEL CHECK CHANNEL CAIZBPATI(N 1 ~

2 ~

3 ~

4 ~

5.

6 ~

7 ~

8 ~

9 ~

10.

11.

12

'3

'4.

15.

16

'7

'8.

Containment Pressure Reactor Coolant Outlet Temperature - To> (Wide Range)

Reactor Coolant Inlet Temperature - T~~~ (Wide Range)

Reactor Coolant Pressure Wide Range Pressurizer Water Level Steam Line Pressure Steam Generator Water Level Narrow Range RWST Water Level Boric Acid Tank Solution Level Auxiliary Feedwater Flow Rate Reactor Coolant System Subcooling Margin Monitor PORV Position Indicator Limit Switches PORV Block Valve Position Indicator Limit Switches Safety Valve Position Indicator - Acoustic Monitor Incore Thermocouples (Core Exit Thermocouples)

Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

Containment Sump Level Containment Water Level M

M M

M M

M M

M M

M M

M M

M M

M(2)

M M

Rf Rg R

~

R R

R R

Rf R

R

. R R(1) 4 R(3) 1.

R (2)

(3)

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 remaining Reactor Vessel Indication train OPERABLE.

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

The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT UNIT 2 3/4 3 47 AMENDMENT NO. 9&/ 9Sf 177

0 C.

Unit 2 and Common Area ire Detect o

S ste s

etection stem Locatio Auxiliary Building a) Elevation 573 b) Elevation 587 c) Elevation 609 d) Elevation 633 e) Elevation 650 f) New Fuel STGE Area Total Number

~ie (x/y)*

/poke (x/y)*

23/0 C

55/0 C

41/0 C

41/0 C

34/0 C

4/0 C

U2 East Main Steam Valve Enclosure U2 Main Steam Line Area El. 612 (Around Containment)

U2 NESW Valve Area El. 612 28/0++

13/0+4 2/0 U2 4KV Switchgear (AB)

U2 4KV Switchgear (CD)

U2 Engr. Safety System Switchgear 6 XFMR. Rm.

U2 CRD, XFMR & Switchgear Rm.

Inverter

& AB Bttry. Rms.

U2 Pressurizer Heater XFMR. Rm.

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.

Ul&2 AFWP Vestibule 0/3 0/3 0/5 0/5 0/2 0/2 0/14 0/17 12/0 4/0 2/0 C

U2 Control Room U2 Switchgear Cable Vault U2 Control Rm.

Cable Vault U2 Aux. Cable Vault 42/0 0/10*~

0/13 Q/76~*

0/6 Ul&2 ESW Basement Area U2 ESW Pump

& MCC Rms.

4/0 C

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'arly warning and notification) instruments.

++

circuit contains both smoke and flame detectors

• ~

two circuits of five detectors each two circuits 'of 38 detectors each COOK NUCLEAR PLANT - UNIT

.:/4 3-52 AMENDMENT NO. 44, 444, 45k, 172

E ELING OPERATIONS E TRAVEL -

SP NT EL STORAGE OOL I ITING CONDITION FOR 0 RA ION 3.9.7 Loads in excess of 2,500 pounds shall be prohibited from travel over fuel assembli.es 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.

~CT ON:

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.

URVEILIANCE RE UIR E

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 wi.th Cook Nuclear Plant - Unit l.

COOK NUCLEAR PLANT - UNIT 2 3/4 9-7 AMENDMENT NO. W, 44,

)7P

T, Q4

EFUELING 0 E

TIONS URVE ILANCE E

EMENTS Cont ued 3.

Verifying that the HEPA fi,lter 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 c&n plus or minus 10%.

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 10% during system operation when tested in accordance with ANSI N510-1980.

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:

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 radi.oactive methyl iodide when the sample is tested in accordance with ANSI N510-1980 (ASTM D 3803-1979, 30 C, 95%,R.H.).

COOK NUCI'~~I PLANT - UNIT 2 3/4 9-13 AMENDMENT NO. 44k, 440,

$ 7/

I 'I Kc 1

h

IOACT VE EF ENTS GASEOUS RADWAS E IMITING CONDITION FOR OPERA ON 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

~aste 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.

~OIION:

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 PIANT - UNIT 2 3/4 11-12 AMENDMENT NO. 4k, ~,

172

S UMENTATTON

~S~S 4

3 3 6 OST-CCIDE STR 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

DELE 4 3 3

8 IRE ETECTIO NS ENT 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 facilityfire 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.

4 3 3 9 RADIO CTIVE LI UID EFFLUENT NS UMENT TION 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,'45,

440, 172

Cl ~

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rent I

II c$

'I

EFUELING OPERATIONS

~SES 3 4 9 9

CO INME GE EXHAUST ISO TIO S

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 thi.s system is required to restrict the release of radioactive material from the containment atmosphere to the environment.

4 9 0

3 4 9 ll W

E EVEL -

C VESSEL G

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' 4

9 12 STORAGE POOL VENTIIATION S

The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly willbe 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 consi.stent 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 willbe 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. 50, 434 172

rv ')

STATE 0 ICHIGAN

)oop G

]oco SCALE 1:24 000 G

7000 re CT

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Cook Nuclear Plant Exclusion Area

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I COOK NUCLEAR PLANT UNIT 2 NENDMENT NO. 1?2

l<

0$V

5.4.2 The total water and steam v'olume of the reactor coolant system is 12,612 plus or minus 100 cubic feet at a nominal Tavg of 704F.

5.5 OROLOG C L 0

C 0

The meteorological tower shall be located as shown on Figure 5.1-3.

5.6 C I ICALI SPE E

5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

A K ff 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 wt%

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 MMD/MTU, or fuel of other enrichments with equivalent reactivity.

COOK NUCLEAR PLANT - UNIT 2 5-5 AMENDMENT NO. $b, 404, 424, 172

C \\

~

1 s7

~i

$t

<I

. I S

TIV CO OLS 6

ESPONSIBILI 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 ORGANIZAT ON NSITE AND OFFS TE 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.

I.ines 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).

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 providing technical support to the plant to ensure nuclear safety.

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.

AGILITY STAF 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.

COOX NUCLEAR PLANT - UNIT 2 6-1 AMENDMENT NO. 48, ~, 4%4

C

~

'Vl

INISTRATIVE CO OLS 3

C LITY ST FF UALI CATIONS 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

~NCTI~O 6.5.1.1 The PHSRC 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 Safety

& Assessment Operations Technical Support Radiation Protection 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. %t, ~, 438, 172