ML17333A361

Amends 208 & 192 to Licenses DPR-58 & DPR-74,revising TS 3/4.3.3.7,TS 3/4.7.9,TS 3.4.7.10,TS 6.2,TS 6.8 & Associated Bases to Relocate Fire Protection Program Elements from TS & Incorporate NRC-approved Fire Protection Program

ML17333A361
Person / Time
Site:	Cook
Issue date:	03/11/1996
From:	John Hickman NRC (Affiliation Not Assigned)
To:
Shared Package
ML17333A362	List: ML17333A361 ML17333A364
References
NUDOCS 9603210366
Download: ML17333A361 (42)

Text

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 D

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G SE Amendment No.208 License No.

DPR-58 1.

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

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated June 20,

1995, and supplemented December 19,

1995, 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.

0~210~~+ 9 o003i5 PDR ADOCK PDR P

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2. 'ccordingly, the following sections under paragraph 2.C of Facility Operating License No.

DPR-58 are hereby amended to read as follows:

2.C.(2)

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

as revised through Amendment No.

208

, are hereby incorporated in the license.

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

2.C.(4)

Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection.Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the SERs dated December 12,

1977, July 31,

1979, January 30,

1981, February 7,

1983, November 22,

1983, December 23,

1983, March 16,

1984, August 27,

1985, June 30, 1986, January 28,

1987, Hay 26,

1987, June 16,

1988, June 17,

1988, June 7,

1989, February 1,

1990, February 9,

1990, March 26, 1990, April 26,

1990, March 31, 1993, April 8,

1993, December 14,

1994, January 24, 1995, April 19,

1995, June 8,

1995, and March 11, 1996, subject to the following provision; The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

3.

This license amendment is effective as of the date of issuance, with full implementation within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION

~

John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:

Page 3 of License No. DPR-58*

Changes to the Technical Specifications Date of Issuance:

March 11, 1996

• Page 3 is attached, for convenience, for the composite license to reflect these changes.

t t

ATTACHMENT TO LIC NSE AMENDMENT N0.208 TO FACI ITY OPERATING IC NSE 0.

DPR-58 DOCKET NO. 50-315 UNIT ICENSE REMOVE Page 3

ECHNIC L SP C

CATIONS INSERT Page 3

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.

REMOVE V

IX XIII 3/4 3-51 3/4 3-52 3/4 3-53 3/4 3-53a 3/4 7-33 3/4 7-34 3/4 7-35 3/4 7-36 3/4 7-37 3/4 7-38 3/4 7-39 3/4 7-40 3/4 7-41 3/4 7-42 3/4 7-43 3/4 7-44 3/4 7-45 3/4 7-46 B 3/4 3-6 B 3/4 7-6 B 3/4 7-7 B 3/4 7-8 B 3/4 7-9 6-2 6-13 INSERT V

IX XIII 3/4 3-51 3/4 3-52 3/4 3-53 B 3/4 3-6 B 3/4 7-6 6-2 6-13 CD This amended license shall, be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I:

Part 20, Section 30'4 of Part 30, Section 40'1 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) aximum Power Level The licensees are authorized to operate the Donald CD Cook Nuclear Plant, Unit No.

1, at steady state reactor core power levels not to exceed 3250 megawatts (thermal).

(2)

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

as revised through Amendment No.2O8

, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications'3)

Less than Four Loo 0 eration The licensees shall not operate the reactor at power levels

~

above P-7 (as defined in Table 3'-1 of Specification 3 ~ 3 ~ 1 ~ 1 of Appendix A to this license) with less than four reactor coolant loops in operation until (a) safety analyses for less than four loop operation have been submitted, and (b) approval for less than four loop operation at power levels above P-7 has been granted by the Commission by amendment of this license.

(4)

Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31,

1979, January 30,

1981, February 7,

1983, November 22,

1983, December 23,

1983, March 16,

1984, August 27,

1985, June 30,

1986, January 28,

1987, May 26,

1987, June 16,

1988, June 17,

1988, June 7,

1989, February 1,

1990, February 9,

1990, March 26,

1990, April 26,

1990, March 31, 1993, April 8,

1993, December 14,

1994, January 24, 1995, April 19,

1995, June 8,

1995, and March 11, 1996, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire ~

Amendment No.

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INDEX LMI'11NGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIIUMENTS

/4.2 WER DISTR UTI N ITS 3/4.2.1 AXIALFLUX DIFFERENCE..

3/4 2-1 3/4.2.2 HEAT FLUX HOT CHANNELFACI'OR 3/4.2.3 NUCLEAR ENTHALPYHOT CHANNELFACTOR 3/4 2-5 3/4 2-9 3/4.2.4 QUADRANTPOWER TILTRATIO 3/4.2.5 DNB PARAMETERS 3/4.2.6 AXIALPOWER LEVEL 3/4 2-11 3/4 2-13 3/4 2-15

/4.

E N

3/4.3.1

'/4.3.2 REACTOR TRIP SYSTEM INSTRUMENTATION ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATION..

3/4 3-1 3/4 3-15 3/4.3.3 MONITORINGINSTRUMENTATION Radiation Monitoring Insmuneatatioa Movable Incore Detectors................

Seismic Instrumentation..

Meteorological Instrumentation Remote Shutdown Instrumeatation Post-Accident lnsuumentation Explosive Gas Monitoring lnstrumeatatioa

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3/4 3-35 3/4 3-39 3/4 3M 3/4 3<3 3/4 3<6 3/4 3-54 3/4 3-57

/4.4 R

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SY 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Stanup and Power Operation 3/4 4-1 COOK NUCLEAR PLANT-UNIT1 AMENDMENTu, uO, ae,208

V

INDEX LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS SECTION PAGE

/4.8 ELECTRICALPOWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating Shutdown 3/4 8-1 3/4 8-9 3/4.8.2 ONSITE POWER DISTRIBUTIONSYSTEMS A.C. Distribution - Operating A.C. Distribution - Shutdown D.C. Distribution - Operating D.C. Distribution - Shutdown D.C. Distribution - Operating - Train N Battery System 3/4 8.3 ALTERNATIVEA.C. POWER SOURCES 3/4 8-10 3/4 8-11 3/4 8-12 3/4 8-15 3/4 8-16 3/4 8-19 3/4.8.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/4 9-1 3/4.9.2 INSTRUMENTATION 3/4.9.3 DECAY TIME 3/4.9.4 CONTAINMENTBUILDINGPENETRATIONS 3/4 9-2 3/4 9-3 3/4 9-4 3/4 9.5 COMMUNICATIONS 3/4 9-5 COOK NUCLEAR PLANT-UNITI AMENDMENT23, 4A, kRS, 473, I$4,208

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INDEX BASES QP

/4.7 SY nt'/4.7.8 HYDRAULICSNUBBERS B 3/4 7-Sa

/4.8

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B 3/4 8 1 4.

FU L P

Tl 3/4.9.1 BORON CONCEYI'RATION.....

B 3/4 9-1 3/4.9.2 INSTRUMENTATION.

B 3/4 9-1 3/4.9.3 DECAYTIME 3/4.9.4 CONTAINMENT BUILDINGPENETRATIONS 3/4.9.5 COMMUNICATIONS 3/4.9.6 MANIPULATORCRANE OPERABILITY 3/4.9.7 CRANE ZtAVEL-SPENT FUEL STORAGE BUILDING..

3/4.9.8 RESIDUAL'.HEATREMOVALAND COOLANT CIRCULATION....

B 3/4 9-1 B 3/4 9-1 B 3/4 9-1 B 3/4 9-2 B 3/4 9-2 B 3/4 9-2 3/4.9.9 CONTAINMENTPURGE AND EXHAUST ISOLATIONSYSTEM.......:....

B 3/4 9-2 3/4.9.10 and 3/4.9.11'ATER LEVEL-REACTOR VESSEL AND STORAGE POOL 3/4.9. 12 STORAGE BQQL,VENTILATIONSYSTEM.".

3/4.9.13 SPENT RJEL'ASK MOVEMENT 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM 3/4.9.15 STORAGE POOL BORON CONCENTRATION..

B 3/4 9-3 B 3/4 9-3 B 3/4 94 B 3/4 94 B 3/4 9<

3/4. 0 SPE IA 3/4.10.1 SHUTDOWN MARGIN 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS B 3/4 10-1 B 3/4 10-1 COOK NUCLEAR PLANT-UNIT1

3/4 LIMITING,CONDITIONS FOR OPERATION AND SURVEILLANCEREQUIIUM1PPIS 3/4.3 INSTRUMENT!TATION Page Intentioaally De1eted COOK NUCLEARPLANT-UNIT1 Page 3/4 3-51 AMENDMENTm, ~, 208

3/4 LIMITINGCONDITION/4 FOR OPERATION ANDSURVEILLANCEREQIJIREMENTS 3/4.3 INSTRUMENTATION Page Intentionally Deleted COOK NUCLEARPLAbiT-UNIT1 Page 3/4 3-v2

3/4 LIMITINGCONDPHONS FOR OPERATION AND SURVEILLANCEREQUIREMI2GS 3/4.3 INSTRUMENTATION Table Intentionally Deleted COOK NUCLEAR PLANT-UNITI Age 3/4 3-~3 AMENDMEVF%~ 448 208

3/4 BASES 3/42 INSTRUMENTATION 3/4.

0 WN Tl N The OPERABILITYof the remote shwdowa instrumentation ensures that sufficient capability is available to permit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the contml taom.

This capability is required in the event coaual room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

/4.

AP D

R The OPERABILITYof the Appendix R remote shutdown insuumentation easures that sufIIcient instrumentation is available to permit shutdown ofthe facilityto COLD SHUTDOWN conditioas at the lccal shutdown indicauon (LSD panel.

Ia the event of a fir, aormal 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, fir watches will bc established in those fire areas where loss of normal power to the LSI panels could occur ia 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.

Thc details of how thcsc fire watches are to be implemented are included in a plant procedure.

ATION 3/4.3..

-A lN The OPERABILITYof the post-accident instrumentauoa ensures that sufficien information is availablc on selected plant parameters to monitor and assess these variables during and following an accident.

COOK NUCLEAR PLANT-UNITI Page B 3/4 M AMENDMEVFV4, 488, 4%, 208

0 3/4 BASES

, 3/4.7 PLANT SYSTEMS

~

3/4 7

'Ihc visual inslectioa frequency is based upoa maintaining a constant level of snubber protection to systems.

Thc method for determining the next iatcrval for thc visual inspection of saubbers is provided based upon the anmbcr of unacceptablc snubbers found during thc previous iaspcction. the category size for each snubber type, aad the previous inslxctioa iaterval per NRC Generic Letter 9049.

A snubber is considered unaccqehlc ifit Ms to satisfy the acceptaace criteria of the visual inspection.

Any inspection whose results required a shotter inspection interval willoverride the previous schedule.

Whea the cause of the rejection of a snubber is clearly established and remedied for that snubber <<nd for aay other snubbers that may be generically suscepable, aad verified by inservice functional testing, that smtbbcr may be cxemptcd Qom being counted as inoperable.

Generically susceptible saubbets are those which arc of a specific make or model and have the same design feamres directly tehted to rejeetioa of thc snubber by visual inspection, or are'imilarly located or exposed to thc same environmental conditions such as temleraaae, radiation, aad vibration.

When a snubber is found inoperable, an engineering evaluation is performed, in addition to the detetminatioa of the snubber mode of failure, in order to deteraunc ifaay safctyMated component or system has beea advetsely affecte by the inoperability ofthc snubber.

The digineering evaluatioa shall determine whether or not thc smtMmr mode of failure has imparted a significant effec or degradatioa oa thc supported componeat or system..

To provide assurance of saubber functional reliability, a tepteseatative sample of the installed saubbers will be

~functionally tested during plant shutdowns at 24 moath intervals.

Observed failures of these sample saabbcrs shall require functional testing of additional units.

'Ihs service life of a saubber is cvaluatcd via manufacturer's input and infotmatioa through considcrttion of the snubber service conditions and associated iastallatioa and maintenance tecords (newly installed snubber, seal replaced, spriag replaced, in high radiation area, in high temperaatre area, etc...). Thc requirement to monitor thc snubber service life is included to ensure that thc snubbers periodically undergo a performance evaluation in view oftheir age and operating coaditions.

These records willprovide statistical bases forfuture consideratioa ofsaubber service life. Thc requirements for the maintenaace of recotds aad the snubber service life review are not intended to affect plant operation.

The mtmber of snubbers to be functionally tested during each surveillance is based on calculatioas performed to allow cxtcnsion of thc surveillance interval from 18 months to 24 months, and therefore, the number of saubbet3 functionally tested dcviates from thc number required by thc Westiaghousc Standard Tcchnical Specifications (NUREG4452, 'Revision 4).

A list of individual snubbcrs with detailed information ofsnubber location and size aad ofsystem affecte shall be available at the plant ia accordance with Scctioa 50.71(c) of 10 CFR Part 50.

Thc accessibility of each snubber shall be determined aad approved by thc Plant Nuclear Safety Review Committee. The determination shall bc based upoa thc existing radiatioa levels and thc expected time to perform a visual inspection in each snubber location as well as other factors associated with accessibility during plant operations (c.g temperature, atmosphere, location.

etc.), and recommendations of Regulatory Guides 8.8 and 8.10.

The additioa or deletion of any saubber shall be made in accordance with Section 50.59 of 10 CFR Part 50.

COOK NUCLEAR PLANT-UNIT1 Page 8 3/4 74 AMENDMENTm, ~, ~, 208

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0 6.0 ADMINISIRATlVECONTROLS b.

At least one licensed Operator shall be in the control room when fuel is ia the reactor.

In addition. while the unit is in Mode I, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room.

C.

An individual qualified in radiation protection ptocedtttes shall be on site when fuel is in the reactor.

d.

All CORE ALTERATIONS shall be directly supervised by a licensed Senior Operator trained or qualified in refueling and CORE ALTERATIONS(SOKA) who has no other concurrent responsibilities during this operation.

e.

The amount of overtime worked by plant staff members performing safety-related functions must be limited in accordance with NRC Policy Statement on working houts (Generic Letter 82-12).

The Shift Supervisor, Assistant Shift Supervisor, and Unit Supervisor shall hold a Senior Operator License.

The Operations Superintendent must hold or have held a Senior Operator License at Cook Nuclear Plant or a similar reactor and one mid-level Operations Production Supervisor shall hold a current Senior Operator License.

The unexected

absence, for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, of the on-site individual qualified in radiation protection procedures is petmitted Imvided immediate action is taken to fill the required position.

COOK NUCLEAR PLANT-UNITI Pnge 6-2 AMIM)MENTÃ.~, 484, 208

e

6.0 ADMINISTRATIVECONTROLS 6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Rev. 2, February 1978.

Deleted.

c.

Deleted.

d.

PROCESS CONTROL PROGRAM implementation.

e.

OFFSITE DOSE CALCULATIONMANUALimplementation.

f.

Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 1.21, Rev. 1, June 1974, and Regulatory Guide 4.1, Rev.

1, April 1975.

Component Cyclic or Transient Limits program, which provides controls to track the UFSAR, Section 4.1, cyclic and transient occurrences to ensure that components are maintained with the limits.

Fire Protection Program implementation.

6.8.2 Each procedure and administrative policy of Specification 6.8.1 above, and changes thereto, including temporary changes, shall be reviewed prior to implementation as set forth in Specification 6.5 above.

6.8.3 A plant program for post-accident sampling shall be established, implemented, and maintained with will ensure the capability to obtain and analyze reactor coolant samples, containment atmosphere noble gas samples, and unit vent gaseous effluent samples for iodines and particulates under accident conditions. The program willinclude the following:

a.

Training of personnel, C.

Procedures for sampling and analysis, Provisions for maintenance of sampling and analysis equipment.

COOK NUCLEAR PLANT-UNIT1 Page 6-13 AMENDMENT&, 4$4, 4SQ, 4$44 ~, 208

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2055&4001 N IANA CHIGAN OW 0

P OCK T 50-3 NA COO NUC A

U N

N T

0 RAT NG CENSE Amendment No. 192 License No.

OPR-74 1.

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

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated June 20,

1995, and supplemented Oecember 19,

1995, 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.

2.

Accordingly, the following section under paragraph 2.C. of Facility Operating License No.

DPR-74 are hereby amended to read as follows:

2.C.(2)

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

as revised through Amendment No. >9~

, are hereby incorporated in the license.

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

2.C.(3)(o)

Indiana Michigan Power Company shall implement and maintain, in effect, all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31, 1979, January 30,

1981, February 7,

1983, November 22,

1983, December 23,

1983, March 16,

1984, August 27,

1985, June 30,

1986, January 28,

1987, Hay 26,

1987, June 16,

1988, June 17,

1988, June 7,

1989, February 1,

1990, February 9,

1990, March 26, 1990, April 26,

1990, March 31, 1993, April 8,

1993, December 14,

1994, January 24,

1995, April 19,

1995, June 8,

1995, and March 11, 1996, subject to the following provision:

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

3:.

This license amendment is effective as of the date of issuance, with full implementation within 180 days.

FOR THE NUCLEAR REGULATORY COMMISSION John B. Hickman, Project Manager Project Directorate III-1 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Attachments:

Pages 8 and 9 of License No. DPR-74*

Changes to the Technical Specifications Date of Issuance:

March ll, 1996

• Pages 8 and 9 are attached, for convenience, for the composite license to reflect these changes.

REMOVE Page 8

9 TTACHMENT TO LICENSE' ENDMENT NO. 192 FACILITY OPERATING LICENSE NO.

DPR-74 DOCKET NO. 50-316 UNIT 2 LICENSE INSERT Page 8

9 TECHNICAL SPECIFICATIONS 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.

REMOVE V

IX XIII 3/4 3-50 3/4 3-51 3/4 3-52 3/4 3-52a 3/4 7-27 3/4 7-28 3/4 7-29 3/4 7-30 3/4 7-31 3/4 7-32 3/4 7-33 3/4 7-34 3/4 7-35 3/4 7-36 3/4 7-37 3/4 7-38 3/4 7-39 3/4 7-40 B 3/4 3-3 B 3/4 7-6 B 3/4 7-7 B 3/4 7-8 B 3/4 7-9 6-2 6-13 INSERT V

IX XIII 3/4 3-50 3/4 3-51 3/4 3-52 B 3/4 3-3 B 3/4 7-6 6-2 6-13

Indiana Michigan Power Company shall implement and

maintain, in effect, all provisions of the approved Fire Protection Program as described in the Updated Final Safety Analysis Report for the facility and as approved in the SERs dated December 12, 1977, July 31,

1979, January 30,

1981, February 7,

1983, November 22,

1983, December 23,

1983, March 16, 1984, August 27,

1985, June 30,

1986, January 28,

1987, May 26,

1987, June 16,

1988, June 17,

1988, June 7,

1989, February 1, 1990, February 9,

1990, March 26,

1990, April 26,

1990, March 31, 1993, April 8,

1993, December 14, 1994,, January 24, 1995, April 19,

1995, June 8,

1995, and March 11, 1996, subject to the following provision:

~

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Amd. 64, 6-18-84.

Deleted per Amd.

121, 5-17-90.

Amd. 6, 6-16-78.

Deleted per Amd. 68, 4-8-85.

S ent Fuel Pool Stora e

The licensee is authorized to store D.

C.

Cook, Unit 1

and Unit 2 fuel assemblies, new or irradiated, up to a total of 3613 fuel assemblies in the shared spent fuel pool at the Donald C.

Cook Nuclear Plant subject to the following conditions:

Fuel stored in the spent fuel pool shall not have a

nominal enrichment greater than 4.95X Uranium-235.

Amd. 13, 10-N-79'.

Deleted per Amd. 63, 4-27-84.

The provisions of Specification 3/4.9.7 are not applicable for loads being moved over the pool for the duration of the spent fuel pool reracking project.

Control of loads moving over the spent fuel pool during the spent fuel pool reracking project shall comply with the criteria of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants."

Administrative controls shall be in place to prevent any load not rigged in compliance with the criteria of NUREG-0612 from passing over the spent fuel pool with the crane interlocks, required by T/S 3/4.9.7, disengaged.

Amendment No. fg, fg, /PE, f/), fPP,192

(v)

Secondar Water Chemistr onitorin Pro ram The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation.

This program shall be described in the station chemistry manual and shall include:

l.

Identification of a sampling schedule, for the critical parameters and control points for these parameters; 2.

Identification of the procedures used to measure the values of the critical parameters; 3.

Identification of process sampling points; 4.

Procedure for the recording and management of data; 5.

Procedures defining corrective actions for off control point chemistry conditions; and 6.

A procedure identifying (a) the authority responsible for the interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.

Amendment No.

fP, fPP,192

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INDEX LIMITINGCONDITI:.NS FOR OPERATION'AND SURVEILLANCEREQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTIONLIMITS 3/4.2.1 AXIALFLUX DIFFERENCE 3/4.2.2 HEAT FLUX HOT CHANNELFACTOR 3/4.2.3 NUCLEAR ENTHALPYHOT CHANNELFACI'OR 3/4.2.4 QUADRANTPOWER TILTRATIO.........

3/4.2.5 DNB PARAMETERS MODE 1 MODES 2, 3, 4 AND 5 3/4.2.6 ALLOWABLEPOWER LEVEL 3/4.3 INSTRUMENTATION 3/4 2-1 3/4 2-5 3/4 2-9 3/4 2-13 3/4 2-15 3/4 2-19 3/4 2-19 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 3/4 3-14 3/4.3.3 MONITORINGINSTRUMENTATION Radiation Monitoring Instrumentation Movable Incore Detectors....

Seismic Instrumentation Meteorological Instrumentation...

Remote Shutdown Instrumentation Post-Accident Instrumentation Explosive Gas Monitoring Instrumentation

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3/4.4 REACTOR COOLANT SYSTEM 3/4 3-34 3/4 3-38 3/4 3-38a 3/4 3-39 3/4 3Q2 3/4 345 I

3/4 3-53 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION Startup and Power Operation 3/4 4-1 COOK NUCLEAR PLANT-UNIT2 AMENDMENT39, 4Q, %$, ASS, 192

b' INDEX LIMITINGCONDITIONS FOR OPI".RATION AND SURVEILLANCEREQUIREMENTS SECTION PAGE 3/4.8 ELECTRICALPOWER SYSTEMS 3/4.8.1 A.C. SOURCES Operating Shutdown..

3/4 8-1 3/4 8-9 3/4.8.2 ONSITE POWER DISTRIBUTIONSYSTEMS A.C. Distribution - Operating A.C. Distribution - Shutdown D.C. Distribution - Operating D.C. Distribution - Shutdown D.C. Distribution - Operating - Train N Battery System 3/4 8.3 ALTERNATIVEA.C. POWER SOURCES 3/4.8.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION 3/4 8-10 3/4 8-11 3/4 8-12 3/4 8-15 3/4 8-16 3/4 8-19 3/4 9-1 3/4 9-2 3/4.9.3 DECAY TIME 3/49-3 3/4.9.4 CONTAINMENTBUILDINGPENETRATIONS

~I 3/4 9.5 COMMUNICATIONS..

3/4 9-4 3/4 9-5 COOK NUCLEAR PLANT-UNIT2 AMENDMENT~, 4$6, 49, 192

47 3/4.7.8 SEALED SOURCE CONTAMINATION........

B3/474 o

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o B3/48 I 3/4.9.1 BORON CONCENTRATION 3/4.9.2 INSTRUMENTATION 3/4.9.3 DECAY TIME.

3/4.9.4 CONTAINMENTBUILDINGPENEIRATIONS 3/4.9.5 COMMUNICATIONS 3/4.9.6 3/4.9.7 3/4.9.S 3/4.9.9 MANIPULATORCI4QK OPERABILITY..

CRANE TRAVEL-SPENT FUEL STORAGE BUILDING...............

RESIDUALHEAT REMOVALAND COOLANT CIRCULATION..........

CONTAINMENTPURGE AND EXHAUSTISOLATION SYSTEM 3/4.9.12 3/4.9.13 STORAGE POOL VENTILATIONSYSTEM SPENT FUEL CASK MOVEMENT 3/4.9.14 SPENT FUEL CASK DROP PROTECTION SYSTEM 3/4.9.15 STORAGE POOL BORON CONCEN'HtATION 3/4.9.10 and 3/4 9.11 WATER LEVEL-REACTOR VESSEL AND STORAGE POOL.....

B3/4 9-1 B3/49-1 83/49-1 B3/49-1 B3/49-1 B3/49-2 B3/49-2 B3/49-2 B3/49-3 B3/49-3 B3/49-3 B3/494 B3/4%4 B3/4%4

/4 3/4.10.1 SHUTDOWN MARGIN.

B34 10.1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTIONLIMITS......

B34 1M COOK NUCLEAR PLANT-UNITZ AMENDMENT4P, 486, 4VS, 192

3/4 LIMI'HNGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREh62fIS 3/4B INSTRUMENTATION Page Intentionally 1Meted COOK NUCLF~ PLAÃ1'-UNIT2 Pagt 3/4 3-50 AMENDMENT44 44&, 192

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURV19LLANCE REQUIIKMEVXS 3/42 INSTRUMENTATION Page Intentionally IMeted COOK NUCLEAR PLANT-UNIT2 Page 3/4 3-51 AMENDMENT~~ 192

3/4 LMITPiGCONDGTONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/49 INSTRUMENTATION Table Intentionally IMeted COOK NUCLEARPLANT-Uk' Page 3/4 3-52 AMENDMENTA, 44$, 484 ~, 192

3/4 BASES 3/4,3 INSTRUMENTATION 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITYofthe post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables during and followingan accident.

3/4.3.3.7 Deleted.

3/4.3.3.9 EXPLOSIVE G S MONITORINGINSTRUMENTATION This instrumentation includes provisions for monitoring the concentrations of potentially explosive gas mixtures in the Waste Gas Holdup System.

The OPERABILITY and use of this instrumentation is consistent with the requirements ofGeneral Design Criteria specified in Section 11.3 ofthe Final Safety Analysis Report for the Donald C. Cook Nuclear Plant.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 3-3 AMENDMENT64, 44$, 449, kVS, 48$, 192

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3/4 BASES

'/4.7 PLANT SYSIXMS 4 7.7 included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions.

These records willpzovide statistical bases for futuze consideration of mubber service life.

The requirements forthc maintenance ofrecords and the snubber service lifereview aze not intended to affect plant operation.

Thc number of snubbers to be functionally tested during each surveillance is based on calculations performed to allow extension of the surveillance interval hem 18 months to 24 months, and thezefoze, the number ofsnubbezs functionally tcstcd deviatcs hem thc mtmber requited by thc Westinghouse Standazd Tcchzucal SpcciQcations (NUIKG4452, Revision 4).

A list of individual snubbezs with detailed infozznation of snubber location and size and of system affected shall be available at the plant in accozdance with Section 50.71(c) of 10 CFR Patt 50.

Thc accessibility of each saubber shall bc detezznined and approved by the Hant Nuclear Safety Review Committcc. The determination shall be based upon thc existing radiation levels and the expcctcd time to perfozm a visual inspecuon in each snubber location as well as other factors associated with accessibility during plant operations (e.g., temleratuze, atmosp!tete, location, etc.), and thc recommendations of Regulatory Guides 8.$ and 8.10.

The addition or deletion ofany snubber~

be made in accordance with Section 5099 of 10 CFR Pazt 50.

/4.7.

S E

The limitations on removable contatnituuion for sources requiring leak testing, including alpha cmittczs, is based on 10 CFR 70.39(c) limits for plutonium. This limitarion will ensure that leakage Gom bypzoduct, souzce and special nuclear material souzces willnot exceed allowable imake values.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 7A AMENDMENTa, m, ue, 444 F4, 4',

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e 6.0 ADMINISIIUTIVE CONTROLS b.

t At least onc licensed Operator ge11 be in the control mom whca fuel is m the reactor.

In addition, while the unit is in Mode 1, 2, 3, or 4, at last one licensed Senior Operator shall be in the conttol toom.

C, An individnal qualificd in t3diatfon protection procedures shall be oa site when fuel is in the teactor.

AllCORE ALTERATIONSshall bc directly supervised by a licensed Senior Operator trained or qualificd ia refueling and CORE ALTERATIONS(~) who has ao other concurrent responsibilities dating this operation.

Co Thc amount ofovertime worked by plant staff mcmbets pctfotming safety~ functions must bc limited ia accordance with NRC Policy Statcmcttt oa working hours (Generi Letter &2-12).

The Shift Supervisor, Assistant ShHt Supervisor, and Unit Supervisor shall hold a Senior Qerator License.

The Opetations Suleriatcadcat must hold or have held a Senior Olerator licenso at Cook Nuclear Plant or a similar teactor aad oac mid-level Operations production Supervisor shall hold a current Senior Operator License.

The unexpected

absence, for a period of time uot to excccd 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, of the on-site individual qualified in radiatioa protection procedures is permitted provided immediate action is taken to fill thc required position.

COOK NUCLEAR PLANT-UNIT2 AMZNDMENT~,~ ~,

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e 6.0 ADMINISTRATIVECONTROLS 6.8 PROCEDURES ND PROGRAMS 6.8.1 Written procedures shall be established, implemented and maintained covering the activities referenced below:

a.

The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Rev. 2, February 1978.

b.

Deleted.

c.

Deleted.

d.

PROCESS CONTROL PROGRAM implementation.

e.

OFFSITE DOSE CALCULATIONMANUAL'mplementation.

f.

Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 1.21, Rev. 1, June 1974, and Regulatory Guide 4.1, Rev. 1, April 1975.

g.

Component Cyclic or Transient Limits program, which provides controls to track the UFSAR, Section 4.1, cyclic and transient occurrences to ensure that components are maintained with the limits.

h.

Fire Protection Program implementation.

6.8.2 Each procedure and administrative policy of Specification 6.8.1 above, and changes thereto, including temporary changes, shall be reviewed prior to implementation as set forth in Specification 6.5 above.

6.8.3 A plant program for post-accident sampling shall be established, implemented, and maintained with will ensure the capability to obtain and analyze reactor coolant samples, containment atmosphere noble gas samples, and unit vent gaseous effluent samples for iodines and particulates under accident conditions. The program willinclude the following:

a.

Training of personnel, b.

Procedures for sampling and analysis, c.

Provisions for maintenance of sampling and analysis equipment.

COOK NUCLEAR PLANT-UNIT2 Page 6-13 AMENDMENTA, 43S, 47$, 478, 486,192

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