Text

Amends 87 & 73 to Licenses DPR-58 & DPR-74,respectively, Revising Tech Specs to Update Offsite Organization Chart & Organization & Responsibilities of Plant Nuclear Safety Review Committee
	ML17321A774
Person / Time
Site:	Cook
Issue date:	08/05/1985
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
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References
	NUDOCS 8508140450
	Download: ML17321A774 (72)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-315 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 87 License No.

DPR-58 1.

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

A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated December 17, 1984, supplemented

.by letter dated June 4, 1985, 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 comon 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 Comission'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-58 is hereby amended to read as follows:

8508i 40450 850805'DR

@DOCK 050003i5; P

PDR (2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

87, are hereby incorporated in the license.

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

3.

The change to the Technical Specifications is to become effective within 45 days of issuance of this amendment.

4.

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

F R THE NUC R

REGULATORY COMMISSION v

n g

C Operating Reactors B

nch ¹1 Division of Licensi

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 5, 1985

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET NO. 50-316 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No.

DPR-74 The Nuclear Regulatory COIImission (the Comission) has found that:

A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated December 17, 1984, supplemented by letter dated June 4, 1985, 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 ComIission; 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 Comission'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:

0

~r

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

73, are hereby incorporated in the license.

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

3.

The change to the Technical Specifications is to become effective within 45 days of issuance of this amendment.

4.

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

FOR THE NUCLEAR ULATORY COMMISSION en a,

Operating Reactors B

nch ¹I Division of Licens

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 5, 1985

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO.

87 FACILITY OPERATING LICENSE NO.

DPR-58 AMENDMENT NO.

73 FACILITY OPERATING LICENSE NO.

DPR-74 DOCKET NOS. 50-315 AND 50-316 Revise Appendix A as follows:

Unit 1

Unit 2 Remove'a es 1-2 3/4 0-1 3/4 3-12 3/4 3-14 3/4 3-26a 3/4 6-18 3/4 6-21 3/4 8-9 3/4 8-14 6-1 6-2 6-6 6-7 6-9 6-10 6-12

" 6-14 6-15 6-18 thru 27 1-2 3/4 1-1 3/4 3-11 3/4 3-13 3/4 3-25a 3/4 3-35 3/4 3-37 3/4 8-9 6-1 6-2 6-5 6-6 6-7 6-9 6-10 6-12 6-14 6-15 6-18 thru 28 Insert Pa es 1-2 3/4 0-1 3/4 3-12 3/4 3-14 3/4 3-26a 3/4 6-18 3/4 6-21 3/4 8-9 3/4 8-14 6-1 6-2 6-6 6-7 6-9 6-10 6-12 6-14 6-15 6-18 thru 24 1-2 3/4 1-1 3/4 3-11 3/4 3-13 3/4 3-25a 3/4 3-35 3/4 3-37 3/4 8-9 6-1 6-2 6-5 6-6 6-7 6-9 6-10 6-12 6-14 6-15 6-18 thru 25

DEFINITIONS REPORTABLE EVENT 1.7 REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.

CONTAINMENT INTEGRITY 1.8 CONTAINMENT INTEGRITY shall exist when:

1.8.1 All penetrations required to be closed during accident conditions are either:

a.

Capable of being closed by an OPERABLE containment auto-matic isolation valve system, or b.

Closed by. manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.

1.8.2 All equipment hatches are closed and sealed.

1.8.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, and 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2.

- CHANNEL CALIBRATION 1.9 A CHANNEL CALIBRATION'shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel 'steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation; This determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

D. C.

COOK - UNZT 1 1-2 Amendment No.

87

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3/4

~ITING COHOmONS FOR OPERATION ANO SURVEILLANCE RE UIREHEHTS 3/4. 0 APPLICABILITY ITIHG CONOITION FOR OPERATION 3.0.1 Lfaftfng Conditions for Operation and ACTION requfraetnts shall be applfcabIe duHng the OPERATIONAL NDES or other conditions specified for each speci f5catf on.

3.0.2 Adherence to the requfrements of the Lfaftfng Condftfon for Operatfon and/or assocfated ACTION within the specified tfaa interval shall constitute complfance with the specification.

In the event the Lfaftfng Condition for

'peration fs restored prior to expfratfon of the specified tfae interval,

'coapletfon of the ACTION statement fs not requfred.

3.0.3 Nen a Lf&tfngCondition for Operation fs not set, except as provided fn the associated ACTION requirements, within one hour action shall be initiated

. to place the unit in a %DE fn which the Speci ffcatfon does not apply by placing ft, as applicable, fn At least HOT STANDBY wfthfn the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUTDOWN within the followfng 5 Fears, and 3.

At least COLD SHiJTDO& within the sybsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Shore corrective aeasures are coepleted that permft operation under the ACTION requirements, the ACTION my be taken in accordance with the specified tfse lfafts as measured from the tfme of failure to meet the Lfaftfng Condftfon for Operation.

Exceptions to these requfrements are stated fn the indfvfdual Specffications.

3.0.4 Entty into an OPERATIONAL %DE or other speciffed applicability condi-tion shall not be made unless the condftions of the Lfaitfng Condftfon for Operagfon are met without reliance on provisions contained fn the ACTION stateaents unless otherwise excepted.

This provfsion shall not prevent passage through OPERATIOHAL HGDES as required to comply with ACTION statements.

3.0.5 Shen a systen, subsystem, train, component or device fs determined to be inoperable solely because its iaergency power source fs inoperable, or solely because fts normal 'power source is inoperable, ft say be considered OPERABLE for the purpose of satfsfying the requfrements of fts applicable Lfeftfng Condi ion for Operation, provided:

(1) fts corresponding normal or emergency power source is OPERABI i; and (2) all 'of fts redundant.systaa(s),.

subsystem(s),

train(s}, ceaponent(s) and device(s) ari OPERABLE, or likewise satisfy the requi.ements of this specification.

Unless both conditions (1) and (2) are satis ied, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actfon shall. be initiated to place Ne unit in a %0E in which the applicable Limiting Conditfon for Operation does not apply by placing it as applicable in:

At least HOT STANDBY ~ithin de next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUT"CWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 3.

At least COLD SHUTDC'.<H within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

~ This Specification fs not applfcab'le in BODES S or 6.

0.

C.

COOK - UNIT L 3/4 0-1

TABLE 4.3-1 REACTOR TRIP SYSTEM

$ NSTRUHENTATIOH SURVEILLANCE RE VIRENENTS FUNCTIONAL UNIT 1.

Manual Reactor Trip 2.

Power Range, Neutron Flux 3.

Power Range, Neutron Flux, High Positive Rate 4.

Power

Range, Neutron Flux, High Negative Rate 5.

Intermediate

Range, Neutron Flux CHANNEL CHECK N.A.

H.A.

N.A.

'HANNEL CAL I BR AT ION N.A.

D(2)

H(3) and q(6)

R(e)

R(S)

R(e)

CHANNEL FUNCTIONAL TEST S/U(1)

S/U(1)

NODES IH MHICH SURVFILLANCE

~

RE IREO H.A.

1, 2

1~

2 1, 2 and

+

O 6.

Soutce Range.

Neutron Flux 7.

Overtemperature hT 8,

Overpower hT 9.

Pressurizer Pressure Low 10.

Pressurizer Pressure-High I

11.

Presshrizer Mater Level--High I

12.

Loss of Flow -.Single Loop R(e)

N and S/U(1) 2(7), 3(7),

4 and 5

1, 2 1,

2 1, 2 1,

2 1,

2

0

TABLE 4.3-1 Continued NOTATION (1)-

(2)

(3)

(4)-

(5) lith the reactor trip system breakers closed and the control rod drive system capable of rod withdrawa1.

C If not performed in previous.?

days.

Heat ba1ance only, above 15K of RATED THERMAL POWER.

A Compare incore'o excore axial imbalance above 15% of RATED THERMAL POWER.

Recalibrate if absolute difference

> 3 percent.

Manual ESF functional input check every 18 months.

Each train tested every other month.

(6)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7)

Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

D. C.

COOK-UNIT 1

C 3/4 3-14 Amendment No. "

I TABLE 3.3-4 Continued

.ENGItlEEREO SAFETY FE'ATURE ACTUATION SYSTEM IHSTRlNENTATION TRIP SETPOItlTS FUllCTIONAL Ua(I T 6.

!10TOR DRIVEN AUXILIARYFEEOMATER PUNPS a.

Steam Generator Hater Level -- Low-Low b.

4 kv Ous Loss of Voltage TRIP SETPOINT 17K of narrow range instrument span each steam generator 3196 volts with a 2-second delay ALLOHAOLE VALUES 16K of narrow range instrument span each steam generator 3196, +10, -36 volts with a 2+.2 second delay c.

Safety Injection d.

Loss of thin Feedwater Pumps 7.

TURBINE DRIVEtl AUXILIARYFEEONTER a.

Steam Generator Hater Level -- Low-Low I

Beactod Coolant peep Ous Und~rvol lpge G.

LOSS OF POHER a,

4 kv Ous Loess of Voltage I

b 4(kv Bus Degraded Voltage llot hppl icable llot Applicable PUMPS

> 17K of narrow range instrument span each steam generator

>> 2750 Voltseach bus 3196 volts with a 2-second delay 3596 volts with a 2.0 min. time delay tlot Applicable tht Appl icable 16% or narrow range instrument span each steam generator

> 2725 Voltseach bus 3196 ~ +10, -36 volts with a 2<.2 second delay 3596, +36, -10 volts with a 2.0 minute t 6 second time delay

0Cff ClS N

TESTABLE DURING ISOLATION TIME Or.

r.

O 0

r.

W 0

b' C

O Cl M

0 63+ NCR-107 64 ~ NCR-108 65.

NCR-109 66.

NCR-110 67; NCR-252 68'CR-40 69.

QCM-250 70.

QCM-350 71 'CR-300 72.

QCR-301 73.

QCR-919 7'CR-920 75 RCR-100 76 ~

RCR 101

=77 ~ VCR-10 78.

VCR 11.

79 'CR-20 Boi VCR-21 81 ~ ZCR-100 82.

XCR-101

83. ZCR-102

84. XCR-103 X..iiii.'i-'~

1 ~

0CM-451 2.

GCM-452 dcM-453 cbM-454 5.

CCM-458 6.

CCM-459 7 ~

BCR-31

, 8 ~

- ECR-32 9.

ECR-33 10.

ECR-35 11 'CR-36 PRZ Liquid Sample PRZ Liquid Sample PRZ Steam Saaple PRZ Steam Sample Primary Mater to Pressure Relief Tank Containmnt Service hir RCP Seal Water Discharge RCP Seal Mater Discharge Letdown to Letdown Hx.

Letdown to Letdown Hx.

Dexinetaliaed Water Supply for Refueling Demineralixed Mater Supply for Refueling PRZ Relief Tank to Gas Anal.

PRZ Relief Tank to Gas Anal.

Glycol Supply to Fan Cooler Glycol Supply to Fan Cooler Glycol Supply from Fan Cooler Glyool Supply from Fan Cooler Control hit to Containment Control Air to Containmnt Isolation Control hir to Containment Isolation Control hir to Containment CCW from RCP Oil Coolers CCW from RCP Oil Coolers CCW from RCP Thermal Bat riet CCM from RCP Thermal Barrier CCM to RCP Oil Coolers 4 Thermal Barrier CCM to RCP Oil Coolers 4 Thermal Barrier Containment Aitbot ne Radiation Monitot Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor Cavity Cavity Yes Yes Yes Yes Yes Yes No No No No Yes Yes Yes Yes Yes Yes Yes Yes No No No No No No No No No No No No No No No 10 10 10 10 10 10 15 15 10 10 10 10 10 10 10 10 10 10 10 10 10 10 6o 6o 30 30 6o 6o 10 10 10 10 10

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1

~ - ~

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TESTABLE DURING XIJQ~XHAXIN ISOLATIOH TIME

~~KSQiDR 12.

VCR-205

13. VCR-206 VCR-207t 1 ~

ICM-111'.

ICM-129 3.

ICM-250 4 ~

ICM-251 5 ~

ICM-260 6.

ZCM-265.

7.

ICM-305 8 ~

ICM-306 9,

ICM-311 10, ICM-321 11.

NPX 151 VI

12. Ph-343 13 ~ SF-151 14, SF-153 15'F-159 16 ~ SF-160 17 SI-171 18 ~ SI-172 Upper Comp. Purge Air Inlet Upper Comp. Purge Air Outlet Cont. Press Relief Fan Isolation TI9%3968H, RHR to RC Cold Legs RHR Inlet to Pumps Boron In)eotion Inlet Boron In)cation Inlet Safety In)eotion Inlet Safety Ingeotion Inlet RHR Suotion from Sump RHR Suotion from Sump RHR to RC Hot Legs RHR to RC Hot Legs Dead Height Tester Containment Servioe Air Refueling Mater Supply Refueling Mater Supply Refueling Cavity Drain to Purifioation Systea Refueling Carity Drain to Purifioation System Safety In)eotion Test Line Aooumulator Test Line Tes Tes Tes Tes No Tes Tes Yes Yes Tes Tes Tes Tes Tes No Tes Yes Yes Tes Tes Tes HA NA NA NA Nh HA Hh NA NA NA HA NA Nh HA NA HA NA NA O

tt 0

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 2.

The pilot cell specific gravity, corrected to 77 F and full

'lectrolyte level, is P 1.200.

3.

-The pilot cell voltage is E2.10 volts, and 4.

The overall battery voltage is

~250 volts.

b.

At least once per 92 days by verifying that:

1.

The voltage of each connected cell is g 2.10 volts under float charge and has not decreased more than 0.05 volts from the value observed during the original acceptance

test, and 2.

The specific gravity, corrected to 77 F and full electrolyte 0

level, of each connected cell is ?1.200 and has not decreased more than 0.03 from the value observed during the previous

test, and 3.

The electrolyte level of each connected cell is between the minimum and maximum level indication marks.

Co At least once per 18 months by verifying that~

1.

The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration.

2.

The cell-to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

3.

The battery is capable of supplying the following emergency loads for the specified times with the battery charger disconnected.

The battery terminal voltage shall be maintained 210 volts throughout the entire test.

D. C.

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

87

ELECTRICAL POWER SYSTEHS SURVEILLANCE RE UIREMENTS (Continued) 2.

The pilot cell specific gravity, corrected to 77 F and 0

full electrolyte level, is ? 1.200.

3.

The pilot cell voltage is g 2.10 volts, and 4.

The overall battery voltage is

>250 volts.

b.

At least once per 92 days by verifying that:

1.

The voltage of each connected cell is ~2.10 volts under float charge and has not decreased more than 0.05 volts from the value observed during the original acceptance

test, and 2.

The specific gravity, corrected to 77 F and full electrolyte 0

level, of each connected cell is g 1.200 and has not decreased more than 0.03 from the value observed during the previous

test, and.

3.

The electrolyte level of each connected cell is between the minimum and maximum level indication. marks.

c.

At least once per 18 months by verifying that:

1.

The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration.

2.

The cell>>to-cell and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

3.

The battery charger will supply at least 10 amperes at

+ 250 volts for at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

d.

At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status the emergency loads for 'the specified times of Table 4.8-2 with the battery charger disconnected.

The battery terminal voltage shall be maintained

$ 210 volts...

throughout the entire test.

e.

At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test.

This performance discharge test shall be performed subsequent to the'atisfactory completion of the required battery service test.

D. C.

COOK - UNIT 1 3/4 8-14 Amendment No.

87

6.O hDMMISTRhTIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Hanager shall be responsible for overall facility operation and shall delegate in vriting thc succession to this responsi-bility during his absence.

6.2 ORGhNIZATION OFFSITE 6,2,1 The offsite organization for facility management and tcchnical support shall bc as shovn on Figure 6.2-1.

FhCILITY SThFF 6.2.2 The Facility organisation shall be as shovn on Figure 6.2-2 and:

'a.

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

b.

ht least onc licensed Operator shall be in the control room vhen fuel is in the reactor.

c.

ht least tvo licensed Operators shall be present in thc control room during reactor start-up, scheduLed reactor shutdovn and during recovery from reactor trips.

d.

hn individual qualified in radiation protection procedures shall be on site vhen fuel is in.the reactor.

e ~

hLL CORE hLTERhTIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling vho has no other concurrent responsibilities during this operation.

h site Fire Brigade of at least 5 members. shaLL be maintained onsite at all times.

The fire brigade shall not include 3

members of the minimum shift crev necessary for safe shutdovn of the'nit or any personnel required for other essential functions during a fire emergency.

The amount of overtime vorked by pLant staff members performing safety-related functions must be limited in accordance vith NRC Policy Statement on vorking hours (Generic Letter No, 82-12).

D. C.

COOK - 'UNIT 1 6-1 hmcndmcnt No. 77

C ~

'I k

c.a

CHAIRMANOF THE 8OARD AND CHIEF EXECUTIVE OFFICER AEPSC INDIANA8I MICHIGANELECTRIC CO.

AND OTHER AEP SU8SIDIARIES VICE CHAIRMAN ENGINEERING AND CONSTRUCTION AEPSC AND VICE PRESIDENT INDIANA4 MICHIGAN ELECTRIC COMPANY

~ 0 ~ ~ 0 ~ ~ ~ ~ ~ 0 ~I~tat ~ 0 ~ 1 ~ ~ ~ ~ 0 ~ 0 ~ 0 ~ ~ 0%

~

EXECUTIVE VICE PRESIDENT AND CHIEF ENGINEER VICE PRESIDENT VICE PRESIDENT OPERATIONS AND INDIANAa MICHIGAN

~ NUCLEAR ELECTRIC COMPANY MANAGEROF NUCLEAR OPERATIONS MANAGERS ENGINEERING DIVISIONS AEPSC

+0 ~ 0 ~ 0 ~ 0 ~ 0 ~ 0 ~ 0 F 0' 0 ~ OQ MANAGER QUALITY

'SSURANCE AEPSC PLANT MANAGER DONALDC. COOK NUCLEAR PLANT

~ 0 ~ 0 ~ ~ ~ ~ ~ 0 ~ 0 ~ ~ 0 ~ 0 ~ ~I~ 0 ~I~ ~ ~ 0 ~ 0 ~ ~I~ 0 ~ 0 ~ 0 ~

AEPSC QA SUPERVISOR (ONSITE)

- PLANTQC SUPERINTENDENT

.~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ 0 ~ ~ ~ 1 ~ ~ ~ ~ ~ ~

~ ~ 0 ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ 00 ~ 0 ~ 0 ~ ~ 0 ~ 0

~ ~ ~

ADMINISTRATIVE8I FUNCTIONAL SUPERVISION TECHNICAL DIRECTION

~i~ ~i~ ~ iooooo TECHNICAL LIAISON

~ ~ ~ ~ ~ ~ ~

FUNCTIONALDIRECTION D.C.COOK - Ui'IT 1 6-2

.ZGURE 6.2-1 ORGANIZATIONALRELATIONSHIPS WITHIN THE AMERICANELECTRIC POWER SYSTEM PERTAINING TO QA 4 QC AND SUPPORT OF THE DONALDC. COOK NUCLEAR PLANT Amendment Uo.

87

ADMINISTRATIVECONTROLS COMPOSITION 6.5.1.2 The PNSRC shall be composed of the:

Chairman:

Member:

. Member:

Plant Manager or Designee Assistant Plant Manager - Maintenance Assistant Plant Manager - Opeza't'ions Operations Superintendent Technical Superintendent Engineering Technical Superintendent - Physical Sciences Maintenance Superintendent Plant Radiation. Protection Supervisor QC Superintendent 6.5.1.3 All alternate members shall be appointed in writing by the PNSRC Chaizman to serve on a temporary basisg

however, no more than two alternates shall participate as voting members in PNSRC activities at any one time.

MEETING PRE UENCY 6.5.1.4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chaizman or his designated alternate.

QUORUM 6.5.1.5 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and sufficient members, including alternates, to constitute a majqrity.

~I~I D.C.

COOK - UNIT 1 6-6 Amendment No. 87

ADMINZSTRATZVE CONTROLS RESPONSIBILITIES 6.5.1.6 The PNSRC shall be responsible for:

a.

Review of 1) all proceudres required by Specification 6.8 and changes

thereto,

2) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b.

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

c ~

Review of all proposed changes to Appendix "A" Technical Specifications.

d.

Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety.

e.

Investigation of all violations of the Technical Specifications including the preparation and'orwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairmaq of the NSDRC.

Review of all REPORTABLE EVENTS.

g.

Review of facility operations to detect potential safety hazards.

Performance of special reviews, investigations of analyses and reports thereon as requested by the Chairman of the NSDRC.

Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.

Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC.

k.

Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the NSDRC.

Review of changes to the PROCESS CONTROL PROGRAM, OFFSZTE DOSE CALCULATION MANUAL, and radwaste treatment system.

D. C.

COOK - UNIT 1 6-7 Amendment No.

87

I'

ADMINISTRATIVECONTROLS COMPOSITION 6.5.2.2 The NSDRC shall be composed of the following Regular Members:

1.

Vice Chairman, Engineering and Construction Executive Assistant to the President, 1&MECo 3.

Executive Vice President and Chief Engineer 4.

Senior Vice President, Electrical Engineering and Deputy Chief Engineer 5.

Assistant Vice President, Mecpanical Engineering 6.

Vice President, Engineering Administration 7.

Vice President, Nuclear Operations (NSDRC Chairman) 8.

Assistant Vice President, Environmental Engineering 9.

Plant Manager, Donald C. Cook Nuclear Plant 10.

Design Division Manager ll.

Manager, Quality Assurance 12.

Consulting Engineer, Nuclear Operations Division 13.

Nuclear Safety and Licensing Section Manager, Nuclear Operations Division (NSDRC Secretary) 14.

Vice President, Fossil Plant Operations ALTERNATE MEMBERS 6.5.2.3 Designated Alternate Members shall be appcinted by the Vice Chairman, Engineering and Construction or such other person as he shall designate.

In addition, Temporary Alternate Members may be appointed by the NSDRC Chairman to serve on an interim basis, as required.

Temporary Alternate members are empowered to act on the behalf of the Regular. or Designated Alternate members

- for whom they substitute.

CONSULTANTS 6.R.2.4 Consultants shall be utilized as determined by the NSDRC Chairman to provide expert advice to the NSDRC.

MEETING FRE UENCY 6.5.2.5 The NSDRC shall meet at least once per six months.

D C.

COOK - UNIT 1 6-9 Amendment No. 87

~I 4

L IT+

it t ii ~

I

ADMINISTRATIVECONTROLS QUORUM 6.5.2.6 A quorum, the minimum number of regular members and alternates required to hold a NSDRC meeting, shall be eight (8) members, of whom no more than two (2) shall be Designated or Temporary Alternates.

The Chairman or Acting Chairman shall be present for all NSDRC meetings.

If the number of members present* is greater than a quorum, then the majority participating and voting at the meeting shall not have line responsibility for operation of the facility.

6.5.2.7 The NSDRC is responsible for assuring that independent** reviews of the following are performedc a.

The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed unde" the provision of 10 CFR 50.59 to verify that such actions did not constitute. an unreviewed safety question.

b.

Proposed'hanges to procedures, equipmerit or systems which involve an unreviewed safety question as defined in 10 CFR 50.59.

c.

Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59.

dO Proposed changes to Technical Specifications or this Operating License.

l Violations of codes, regulations, ordexs, Technical Specifications, license requirements, or of internal procedures ox instructions having nuclear safety significance.

f.

Significant operating abnormalities or deviations foxm normal and expected performance of plant equipment that affect nuclear safety.

g All REPORTABLE EVENTS h.

All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures,

systems, and components.

i.

Reports and meeting minutes of the PNSRC.

Regular NSDRC members are expected to attend the meeting whenever

possible, and alternates may attend as voting members only on an irregular basis. If both a regular member and his alternate attend a meeting, only the regular member may participate as a voting member, and the alternate is considered a guest.

~* Independent reviews may be performed by groups which report directly to the NSDRC and which must have NSDRC membership participation.

D. C.

COOK - UNIT 1 6 10 Amendment No.

87

ADMINISTRATIVECONTROLS m.

The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months.

n.

The performance of activities required by the Quality Assurance Program to meet the criteria of Regulatory Guide 1.21, Rev. 1, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975 at least once per 12 months.

AUTHORITY 6.5.2.9 The NSDRC shall report to and advise the Vice Chairman, Engineering and Construction,

AEPSC, on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of NSDRC activities shall be prepared, approved and distributed as indicated below:

a.

Minutes of each NSDRC meeting shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC< within 14 days following each meeting.

b.

Reports of reviews encompassed by Section 6.5.2.7 above, shall be

prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following completion of the review.

C ~

Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice Chairman, Engineering and Construction,

AEPSC, and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6. 6 REPORTABLE EVENT ACTION 6.6.1 Each REPORTABLE EVENT requiring notification to the Commission shall be reviewed by the PNSRC and submitted to the NSDRC and the Vice President, Nuclear Operations.

D. C.

COOX - UNIT 1 6-12 Amendment No.

ADMINISTRATIVECO ROLS 6.8.3 Temporary changes to procedur'es of 6.8.1 above may be made provided:

a.

The intent of the original procedure is not altered.

b.

The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected.

c.

The change is documented, reviewed by the PNSRC and approved by the Plant Manager within 14 days of implementation.

6.9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the follow'ing reports shall be submitted to the Regional Administrator unless otherwise noted.

STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic perfor-

mance of the plant.

6.9.1.2 The startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications.

Any corrective actions that were required to obtain satisfactory operation shall also be described.

Any additional specific details required in li'cense conditions based on other commit-ments shall be included in this report.

r V

D. C.

COOK - UNIT 1 6-14 Amendment No. 87

0

+I

l

ADHINISTRhTIVE CONTROLS povez operatfon),

supplementary reports shall bc submftted at least every three months until all three events have been complcted.-

hNNUhL REPORTS 6.9.1.4 haaual reports covcrfag the activities of thc uait is described belov for the previous calendar year shall be submitted prior to March 1 of each year.

The faftfal report shall bc submitted prior to March 1 of the year folloving fnitfal criticality.

6.9.1.5 ao Reports required oa an annual basis shall include:

N h tabulatfoa on an aanual basis of the number of station, utility and other pcrsoanel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man ran exposure according to vork aad gob functions, e.g.,

reactor operations aah surveiU,ance, faservfce inspection, routine maintenance, special mafntenaacc (describe maintenance),

vaste processing, aad refueling.

The dose assignmcnt to various duty functions may be estimates based on.

pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20X of the individual total dose need not be accouated for.

In the aggregate, at least 80X of the tbtal vhole body dose received from external sources shall be. assigacd to specific ma)or vork fuactions.

b.

The complete results of steam generator tube iaservicc inspections performed during the report period (reference Specification 4.4.5.5.b).

c.

Documentatfon of all challeages to thc pressurizer povcr operated relief valves (PORVs) or safety valves.

1 h single submittal may be made for a multiple unit station.

The submittal should combine those scctioas that are comaon to all units at the statfon.

2 This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

D. C.

COOK -'NIT 1

6-15 hmendment No. 77

ADMINISTRATIVECONTROLS The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and otheg nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nucleaz Power Operation.

Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1.

The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period:

a.

Volume (cubic meters)i b.

Total curie quantity (specify whether determined by measurement or estimate),

c.

Principal radionuclides (specify whether determined by measurement or estimate),

d.

Type of.waste (e.g., spent resin, compacted dry waste, evaporator bottoms),

e.

Type of container (e.g.,

LSA, Type A, Type B, Large Quantity),

and f.

Solidifcation agent (e.g.,

cement).

0 The radioactive effluent release report shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluent on a quarterly basis.

The radioactive effluent release reports shall include any change to the PROCESS CONTROL PROGRAM (PCP) and the OFFSITE DOSE CALCULATION MANUAL (ODCM) made during the reporting period.

MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience

. shall Qe submitted on a monthly basis to the'Director> Office Of Management and Program Analysis, U.S. Nuclear Regulatory Commission,,Washington, D.C.

20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report.

D. C.

COOK - UNIT 1 6-18 Amendment No.

87

ADMINISTRATIVECONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator-within the time period specified for each report.

These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference Specifications:

a.

Znservice Inspection Program. Review, Specification 4.4.10.

b.

ECCS Actuation, Specifications 3.5.2 and 3.5.3.

c.

Inoperable Seismic Monitoring Instrumentation, Specification 3.3.3.3.

d.

Inoperable Meteorological Monitoring Instrumentation, Specification 3.3.3.4.

e.

Seismic event analysis, Specification 4.3.3.3.2.

f.

Sealed Source leakage in excess of limits, Specification 4.7.7.1.3.

g.

Fire Detection Instrumentation, Specification 3.3.3.7.

h.

Fire Suppression

Systems, Specifications 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4.

D. C.

COOK - UNIT 1 6-19 Amendment No.

87

'lk ly 5',

K

ADMINISTRATIVECONTROLS 6.10 RECORD RETENTION F 10.1 The following records shall be retained, for at least five years:

a ~

b.

C ~

d.

e.

Records and logs of unit operation covering time interval at each power level.

Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

All REPORTABLE EVENTS submitted to the Commission.

Records of surveillance activities, inspections and calibrations required by these Technical Specifications.

Records of changes made to the piocedures required by Specification 6.8.1.

Records of sealed source leak tests and results.

Records of annual physical inventory of all sealed source material on record.>>

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a.

b.

C ~

d.

e.f.

go 3 ~

k.l.

m.

n.

Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report.

Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

Records of radiation exposure for all individuals entering radiation control areas.

Records of gaseous and liquid radioact'ive material released to the environs.

Records of radioactive shipments.

Records of transient or operational cycles for those facility components identified in Table 5.9-1.

Records of training and qualification for current members of the Plant staff.

Records of in-service inspections performed pursuant to these Technical Specifications.

Records of Quality Assurance activities required by the QA Manual.

Records of reviews performed for changes made to procedures or equipment. or review of tests and, experiments 'pursuant to.

10 CFR 50.59.

Records of meetings of the PNSRC and the NSDRC.

Records for Environmental Qualification which are covered under the provisions of paragraph 6.13.

Records of reactor tests and experiments.

Records of the service lives of hydraulic snubbers listed on Table 3.7-4 including the date at which service life commences and associated installation and maintenance records.

D. C.

COOK - UNIT 1 6-20 Amendment No.

87

ADMINISTRATIVECONTROLS 6.'l RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be

approved, maintained and adhered to for all operations involving personnel radiation exposure.

6'2 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20:

a ~

A High Radiation Area in which the intensity of radiation is greater than 100 mrem/hr but less than 1000 mrem/hr shall be barricaded and conspicuously posted as a High Radiation Area and entrance thereto shall be controlled by issuance of a Radiat'ion Work Permit and any individual or group of individuals permitted to enter such areas shall be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area.

b.

A High Radiation Area in which the intensity of radiation is greater than 1000 mrem/hr shall be subject to the provisions of 6.12.1.a

above, and in addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Operating Engineer on duty.

6.13 ENVIRONMENTAL UALIFICATION 6.13.1 By no later than June 30, 1982, all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of Division of Operating Reactors "Guidelines for Evaluating Environment Qualification of Class lE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position on Environmental Qualification of Safety-Belated Electrical Equipment",

December 1979.

Copies of these documents are attached to Order for Modification of License No. DPR-74, dated October 24, 1980.

6.13.2

.By no later than December 1, 1980, complete and auditable records must be available and,maintained at-a central location which, describe the env'ronmental qualification method used for all safety-related electrical equipment in sufficient detail to document 'the degree of compliance with the DOR Guidelines or NUREG-0588.

Thereafter, such records should be updated and maintained current as equipment is

replaced, further,.tested, or otherwise further qualified.

D. C.

COOK - UNIT 1 6-21 Amendment No.

ADMINISTRATIVECONTROLS 6.14 PROCESS CONTROL PROGRAM (PCP) 6.14.1 The PCP shall be approved by the Commission prior to implementation.

6.14.2 Licensee initiated changes to the PCP:

1.

Shall be'submitted to the Commission in the semi-annual Radioactive Effluent Release Report for the p riod in which the change(s) was made.

This submittal shall contain:

a.

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; b.

A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and c.

Documentation of the fact that the change has been reviewed,and found acceptable by the PNSRC.

2.

Shall become effective upon review and acceptance by the PNSRC.

6 15 OFFSITE DOSE CALCULATION MANUAL (ODCM)

6. 15. 1

., 6.15.2 The ODCM shall be approved by the Commission prior to implementation.

C Licensee initiated changes to the ODCMi Shall be submitte'd to the Commission in the Semi<<Annual Radioactive Effluent Release Report in the next report after the report period the change(s) was made effective.

This submittal shall contain:

a ~

b.

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information.

Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s);

A determination that the chang'e wil'1 not reduce the accuracy or reliability of dose calculations or setpoint determinations; and C ~

Documentation of the fact that the change has.reviewed and found acceptable by the PNSRC.

D.

CD COOK - UNIT 1 6-22 Amendment No.

IW

ADMINISTRATIVECONTROLS 2.

Shall become effective upon review and acceptance by the P¹RC.

6.15.3 Commission initiated changes:

1.

Shall be determined by the PNSRC to be applicable to the facility after consideration of facility design.

2.

The licensee shall prov'de the Commission with written noti-fication, of their determination of applicability including any necessary revisions to reflect facility design.

1 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Li id, Gaseous and Solidi 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

Shall be reported to the Commission in the Annual Operating Report for the period in which the evaluation was reviewed by the (PNSRC).

The discussions of each change shall contain:

a.

A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59<

b.

Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information; Co d.

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; An evaluation of the change which shows thepredicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto'.

An evaluation of the change which shows the expected maximum exposure to individuals in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto' A, comparison of the predicted releases of radioactive materials, in liquid and gaseous'ffluents and in solid waste, to the actual releases for the period prior to.when the changes are to be made; D. C.

COOK - UNIT 1 6-23 Amendment No.

87

ADMINISTRATIVECONTROLS g.

An estimate of the exposure to plant operating personnel as a result of the change; and h.

Documentation of the fact that the change was reviewed and found acceptable by the PNSRC.

2.

Shall become effective upon review and acceptance by the PNSRC.

6.16.2 Commission initiated changes:

1.

The applicability of the change to the facility shall be determined by the (PNSRC) after consideration of the facility design.

2.

The licensee shall provide the Commission with written notification of its determination of applicability including any necessary revisions to reflect facility design.

D. C.

COOK - UNIT 1 6-24 Amendment No.

87

gf r

t>'

II

DEFINITIONS REPORTABLE EVENT 1.7 A REPORTABLE EVENT shall be any of those conditions specified in 10 CFR 50.73.

CONTAINMENT INTEGRITY 1.8 CONTAZNMENT INTEGRITY shall exist when:

1.8.1 All penetrations required to be closed during accident conditions are either:

a.

Capable of being closed by an OPERABLE containment auto-matic isolation valve system, or b.

Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3.1.

1.8.2 All equipment hatches are closed and sealed.

1.8.3 Each air lock is OPERABLE pursuant to Specification 3.6.1.3, 1.8.4 The containment leakage rates are within the limits of Specification 3.6.1.2, and 1.8.5 The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is'PERABLE.

CHANNEL CALIBRATION 1.9' CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST.

The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channe'1 steps such that the entire channel is calibrated.

CHANNEL CHECK 1.10

'A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by.observation.

This determination shall

include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

D. C.

COOK - UNIT 2 1-2 Amendment No. 73

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4. 1. 1 BORATZON CONTROL SHUTDOWN MARGIN - T w 200 F I

av LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be ? 1.6kAk/k.

APPLICABILITY:

MODES 1, 2,* 3, and 4.

ACTION:

With the SHUTDOWN MARGIN C,1.6%hk/k, immediately initiate and continue boration at W10 gpm of 20,000 ppm boric acid solution or equivalent until the required SHUTDOWN MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be h1.6%4k/k:

a.

b.

Co Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable.

Zf the inoperable contiol rod is im-movable or untrippable, the above required SHUTDOWN MARGIN shall be increased by an amount at least equal to the with-drawn worth of the immovable or untrippable control rod(s).

When in MODES 1 or 2, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying

¹ that control bank withdrawal is within the limits of Specification 3.1.3.6.

When in MODE 2

, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to achieving reactor

¹¹ criticality by verifying that the predicted critical control rod position is within the limits of Specifichtion 3.1.3.6.

~

See Speci'al Test Exception 3.10.1

.¹ With Keff +1.0

¹¹ With Keff C le 0 D.

C.

COOK - UNIT 2 3/4 1-1 Amendment No.

TAALE 4 3-1 1

REhCTOR TAIP SYSTEH INS7AUHENTAT10M SUAVE7l.t.ANCE AE lABIENTS

~

fUHCTIGBhf. UNIT CffhNNEL CIIEN CffhNNEL l

CALI QADI PN CffllNNEL FUNCT IOHAt.

tESr BODES IH NIICfl SURVEIllnNCE AE UIAEO l.

Hanual Aeagtor Trip 2.

Poorer Itange, Neutron.flux Power Adnoe, Neutron Flux,

-. fffgfi Pnsf tive Ante Poorer flange, Nr.<itron F)uxor flI gfe flega tfve Aa te H.A.

. H.A.

H.h.

S(U(1)

O(2). H(3) and C)(6 N.A.

~

1,2 1,2

'1 ~ 2 5.

Intermr.diate

Range, Neutron Flux 6.

Source

Range, Neutron Flux 7.

Overtemperature hT 0.

Overpower aT 9.

- Pressurizer Pressure-l.os 10.

Pressurizer Pressure-Iligh ll; Presser'(zer Mater level-Jlfoh l2.

loss of Floe - Sfngle loop

~

S

'S S

S R(6)

R(6)

R

'iuf->)

~ and Sfu(i}

0, Zan'd*

2(7),3(7},4 and 5

1,2

'lo 2 lg 2 I

P

TABLE 4.3-1 Continued NOTATION (2)

(3)

With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal.

If not performed

$ n pre'vfbus 7 days.

Heat balance only, above 15% of RATED THERMAL POWER.

Adjust channel 'if absolute difference

~ 2 percent.

Compare incore to excore axial offset 'above 15K of RATED THERMAL POWER.

Recalibrate if absolute difference

~ 3 percent.

(4)

Manual ESF functional input check every 18 months.

(5) -

Each train tested every other month.

(6)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7)

Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

D.

C.

COOK - UNIT 2 3/4 3-13 Amendment No.

TABLE 3. 3-4 Continued ENGINEERED SAFEIY FEATURE ACTUATION SYSTEH IHSTRUMEHTATION TRIP SETPOINTS CD8 FUNCTIONAL UNIT 6.

MOTOR DRIVEN AUXILIARYFEEDMATER PUMPS a.

Steam Generator Mater Level --.Low-Low b.

4 kv Bus Loss of Voltage c.

Safety Injection d.

Loss of Hain Feedhater Pumps 7.

TURBINE DRIVEN AUXILIARYFEEOMATFR PUMPS a.

Steam Generator Mater Level Low Low b.

Reactor Coolant Pump Bus Undervoltage

8. 'OSS OF POMER TRIP SETPOINT.

R 21K of narrow range instrument..span each steam generator 3196 volts with a 2 second delay

. Not Applicable Hot Applicable 2 21K of narrow range instrument span each

. steam generator 2 2750. Vnlts--each bus ALLOMABLE VALUES R 2'f narrow range instrument span each steam generator 3196, +10, -36 volts with a 2 i 0.2 second delay Hot Applicable Not Applicable

? 20K of narrow range instrument span each steam generator 2 2725 Volts--each bus fD CL B

fO tt O

a.

b.

4 kv Gus Loss of Voltage 4 kv Bus Degraded Volt'age 3196 volts with a 2 second delay 3596 volts with a 2.0 minute time delay.

3196, +16', -36 volts with a 2.i.0.2 second delay

+36'8 ~ol tR Wth a 2.0 minute t 6 second time delay C

P

, TAbLE 3.3-6 RA91ATION HONITORING INSTRUHENTATION OPERATION HODE/INSTRUHENT HINIHUH CltANNELS OPERABLK HIGN ALARH/TRIP SETPOINT HEASURKHENT RANGE ACTIOH l.

HODKS 1,2,3, 6 4 a ~

b.

AREA HONITOR

i. Upper Containment PROCESS

~SNITORS

i. Particulate ii. Noble Gas

-1 4

A2 x normal channel read)ngIO to 10

<<R/hr 19 2 x normal channel read)ngl.SxlO to 1.5 uCi 20

-4 2 x normal channel zead)ng 10 to 10 uCi/cc 20 2.

HODE 6 a.

TRAIN A i, Containment Area Radiation Channel-VRS-2 101 ii. Particulate Channel-ERS-, 2301 iii. Nobl'e Gss Channel-ERS" 2305 b.

TRAIN b i, Contain<<ent Area Radiation Channel-VRS-2201 ii. Particulate Channel-ERS"2401 iii. Nobl'e Gas Channel ERS-2405

~ny 2/3 Channels any 2/3 Channels Sa<<e sa 2.a Ss<<e as 2.a Sa<<p ss 2.a Same ss

2. ~

Sa<<e as 2.a S}<<}e as 2. ~

-1 4

+~ x normal channel read)ng10 to 10

<<R/hr

~2 x normal channel reading 1.5xlo to 1.5 uci

-7

<2 x normal channel reading 10 to 10 uCi/cc 22 22 I

ea., Spent Fuel Storage

~15 <<R/hr 10 to 10

<<R/hr lg

-1 4

~ Mith fuel istorage pool or building.

TABLE 4.3-3 RADIATION MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENIS OPERATING MODE/I!ESTRUMENT I ~ MODES I ~ 2H 3 ~ 6 4 a. AREA HO.'EDITOR

i. Upper Containment',

PROCESS .'SNITORS

i. Particulate ii. Noble Gas CHANNEL CHECK S

S CHANNEL CALIBRATION R R CHANNEL FUNCTIONAL TEST M M HODES FOR WHICH SURVEILLANCE IS RE IRED I', 3 4 4 I,2,3@4 l, 2] 3 S 4 2, HODE 6 I P a. TRAIN h

i. Containment Area Radiation Channel S

ii. Particulate Channel S iii. Noble Gas Channel S R R-R M M M 6 6 6 b, TRAIN B i; Containment brea Raliation Channel ii. Particulate Channel iii. Noble Gas Channel 3o

a.

SPENT FLEL STORAGE S S S R R R M M M 6 6 6

Mfth fuel in the storage pool or building.

I

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 2. The pilot cell specific gravity, corrected to 77 F and full o electrolyte level, is P 1.200, 3. The pilot cell voltage is + 2.10 volts, and 4. The overall battery voltage is 5 250 volts. b. At least once per 92 days by verifying that: 1. The voltage of each connected cell is > 2.10 volts under float charge and has not decreased more than 0.05 volts from the value observed during the original acceptance

test, 2.

The specific gravity, corrected to 77 F and full electrolyte level, of each connected cell is 2 1.200 and has not decreased more than 0.03 from the value observed during the previous

test, and 3.

The electrolyte level of each connected cell is between the minimum and maximum level indication marks. c. At least once per 18 months by verifying that: 1. The cells, cell plates and battery racks show no visual indication of physical damage or abnormal deterioration. 2. The cell-to-cell and tezminal connections are clean, tight, and coated with anti-corrosion material. 3. The battery charger will supply at least 140 amperes at +250 volts for at least 4 hours. d. e. At least once per 18 months, during shutdown, by verifying that the battery capacity is adequate to supply and maintain in OPERABLE status the emergency loads for the specified times of Table 4.8-1A with the battery charger disconnected. The battery terminal voltage shall be maintained +210 volts throughout the entire test. At least once per 60 months, during shutdown, by verifying that the battery capacity is at least 80% of the manufacturer's rating when subjected to a performance discharge test. This performance discharge test shall be performed subsequent to the satisfactory completion of the required battery service test. D. C. COOK - UNIT 2 3/4 8-9 AMENDMENT NO.

6.0 hININISTRATIVE CONTROLS 6.1 RESPONSIBILITY'.1.1 The Plant Manager shall be responsible for overall facility operatioa and shall delegate in vriting thc succession to this respon-sibility during his absence.

6. 2 ORGANIZATION OFFSITE 6.2.1 The offsite orgaaixatfon for facility management aad technical support shall be as shovn ia Ffgurc 6.2-1.

1 FhCILITY SThFF I 6.2.2 The Facility organixatfon shall be as shovn on Figure 6.2-2 and: a. Each oa duty shift shall be composed of at least the minimum shift crev composition shovn fn Table 6.2-1. b. ht least oac licensed Operator shall be in the control room vhea fuel fs fa thc reactor. c. ht least tvo licensed Operators shall be present fn tha control room during reactor start-up, scheduled reactor shutdova and during recovery from reactor trips. d. ha individual qualified fn radiation protection procedures shall be oa site whea fuel fs fa the reactor. e. hLL CORE hLTERhTIONS shall be directly supervised by either a licensed Seafor Reactor Operator or Senior Reactor Operator Limited to Fuel Handlfag vho has no other coacurrcat responsibilities during this operatfoa. h sfte Fire Brigade of at least 5 members shall be maintained onsite at all tfmes. The Fire Brigade shall not include 3 members of the afaimum shift crev necessary for safe shut-dova of the uaft or any personael required for other essential fuactfons durfag a fire emergency. I The amount of overtime vorkcd by plant staff members pcriormiag safety-related functions must be limited fa accordance vfth the NRC Policy Statement oa vorkiag hours (Generic Letter No. 82 12) ~ D. C. COOR - UNIT 2 6-1 hmcndmeat No.

CHAIRMANOF THE BOARD AND CHIEF EXECUTIVE OFFICER AEPSC INDIANA5 MICHIGANELECTRIC CO. AND OTHER AEP SUBSIDIARIES VICE CHAIRMAN ENGINEERING AND CONSTRUCTION AEPSC AND VICE PRESIDENT INDIANA& MICHIGAN ELECTRIC COMPANY VICE PRESIDENT VICE PRESIDENT OPERATIONS AND INDIANA4 MICHIGAN NUCLEAR ELECTRIC COMPANY MANAGEROF NUCLEAR OPERATIONS ~ o ~ ~ o ~ ~ ~ ~ ~ ~ o ~ ~ ~ o ~ ~ ~ ~ ~ o ~ o ~ ~ o ~ o ~ o ~ ~ oo I I EXECUTIVE VICE PRESIDENT AND CHIEF ENGINEER MANAGERS ENGINEERING DIVISIONS AEPSC 4 ~ ~ o ~ o ~ ~ ~ ~ ~ ~ ~ o ~ ~ o ~ ~ o ~ oo MANAGER OUALITY ASSURANCE. AEPSC PLANT MANAGER DONALDC. COOK NUCLEAR PLANT ~ ~ ~ oo ~ ~ ~ o ~ o ~ ~ o ~ ~ o ~ ~ ~ ~ ~ ~ o ~ o ~ ~ ~ o ~ ~ ~ o ~ o ~ ~ ~ ~ AEPSC QA SUPERVISOR (ONSITE) 'I ~ ~ ~ ~ ~ ~ ~ ~ o ~ o ~ o ~ ~ ~ ~ ~ o ~ o ~ ~ o ~ ~ o ~ o ~ ~ o ~ o ~ ~ ~ o ~ ~ o ~ o ~ o ~ ~ ~ o ~ o ~ ~ ADMINISTRATIVE4 FUNCTIONAL SUPERVISION ~ ~. TECHNICAL DIRECTION ~ o o o o o o o o o ~ o' TECHNICAL LIAISON ~ ~ ~ ~ ~ FUNCTIONALDIRECTION Q.C.COOK -UhIT 2 6-2 VIGUPX 6.2-1 ORGANIZATIONALRELATIONSHIPS WITHIN THE AMERICANELECTRIC POWER SYSTEM PERTAINING TO QA 5 QC AND SUPPORT OF THE DONALDC. COOK NUCLEAR PLANT Amendment ho. 73

6 ' ADMINISTRATIVECONTROLS 6.3 FACILITY STAFF UALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum quali-fications of ANSI N18.1-1971 for comparable positions, except for (1) the Radiation Protection Supervisor who shall meet or exceed the qualifications of Regulatory Guide 1.8, September

1975, and (2) the Shift Technical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the facility staff shall he maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the Plant Manager and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976. 6.5 REVIEW AND AUDIT 6 ~ 5 1 PLANT NUCLEAR SAFETY REVIEW COMMITTEE (PNSRC) 6.5.1.1 The PNSRC shall function to advise the Plant Manager on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The PNSRC shall he composed of the: Chairman: Member: Member: Member: Member: Members Member: Members Member: Plant Manager or Designee Assistant Plant Manager - Maintenance Assistant Plant Manager - Operations Operations Superintendent Technical Superintendent - Engineering Technical Superintendent - Physical Sciences Maintenance Superint'endent Plant Radiation Protection Supervisor QC Superintendent D.C. COOK - UNIT 2 6-5 Amendment No. 73

ADMINISTRATIVECONTROLS ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the PNSRC Chairman to serve on a temporary basis;

however, no more than two alternates shall participate as voting members in PNSRC activities at any one time.

MEETING FRE UENCY 6.5. 1. 4 The PNSRC shall meet at least once per calendar month and as convened by the PNSRC Chairman or his designated alternate. ~UORUN 6.5.1.5 A quorum of the PNSRC shall consist of the Chairman or his designated alternate and sufficient members, including alternates, to constitute a majority. RESPONSIBILITIES 6.5.1.6 The PNSRC shall be responsible for: a. Review of 1) all procedures required by Specification 6. 8 and changes

thereto,

2) any other proposed procedures or changes thereto as determined by the Plant Manager to affect nuclear safety.

b. Review of all proposed tests and experiments that affect nuclear " safety. Co Review of all proposed changes to Appendix "A" Technical Specifications. d. Review of all proposed changes or modifications to plant systems or equipment that affect nuclear safety. D. C. COOK UNIT 2 6-6 Amendment No. 73

ADMINISTRATIVECONTROLS e. Investigation of all violations of the Technical Specifications including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairman of the NSDRC. f. Review of all REPORTABLE EVENTS. g. Review of facility operations to detect potential nuclear safety hazards. h. Performance of special reviews, investigations of analyses and reports thereon as requested by the Chairman of the NSDRC. i. Review of the Plant Security Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC. j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the Chairman of the NSDRC. k. Review of every unplanned onsite release of radioactive material to the environs including the preparation and forwarding of reports covering evaluation and recommen-dations to prevent recurrence to the NSDRC. l. Review of changes to the PROCESS CONTROL PROGRAM, OFFSITE DOSE CALCULATION MANUAL, and radwaste treatment system. AUTHORITY 6.5.1.7 The PNSRC shall: a. Recommend to the Plant Manager written approval or disapproval of items considered under 6.5.1.6(a) through (d) above. b. Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question. C ~ Provide written notification within 24 hours to the NSDRC of disagreement between the PNSRC and the Plant Manager;

however, the Plant Manager.shall have. respons&ility'or resolution of such disagreements pursuant to 6.1.1 above.

'.,C. COOK UNIT 2 6-7 Amendment No. 73

I

ADMINISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The NSDRC shall be composed of the following Regular Members: 1. Vice Chairman, Engineering and Construction 2. Executive Assistant to the President, ZaMECo 3. Executive Vice President and Chief Engineer 4. Senior Vice President, Electrical Engineering and Deputy Chief Engineer 5. Assistant Vice President, Mechanical Engineering 6. Vice President, Engineering Administration 7. Vice President, Nuclear Operations (NSDRC Chairman) 8. Assistant Vice President, Environmental Engineering 9. Plant Manager, Donald C. Cook Nuclear Plant 10. Design Division Manager ll. Manager, Quality Assurance 12. Consulting Engineer, Nuclear Operations Division 13. Nuclear Safety and Licensing Section Manager, Nuclear Operations Division (NSDRC Secretary) 14. Vice President, Fossil Plant Operations ALTERNATE MEMBERS 6.5.2.3 Designated Alternate Members shall be appointed by the Vice Chairman, Engineering and Construction or such other pezson as he shall designate. In addition, Temporary Alternate Members may be appointed by the NSDRC Chairman to serve on an interim basis, as required. Temporary Alternate members are empowered to act on the behalf of the Regular or Designated Alternate members for whom they substitute. go CONSULTANTS 6.5.g.4 Consultants shall be utilised as detezmined by the NSDRC Chairman to proCide expert advice to the NSDRC. 6.5.2.5 The NSDRC shall meet a< least once per six months. . D. C. COOK - UNIT 2 6-9 Amendment No. 73

C. tg fV S 81 1 ,~II

ADMINISTRATIVECONTROLS QUORUM 6.5.2.6 A quorum, the minimum number of regular members and alternates required to hold a NSDRC meeting, shall be eight (8) members, of whom no more than two (2) shall be Designated or Temporary Alternates. The Chairman or Acting Chairman shall be present for all NSDRC meetings. Zf the number of members present>> is greater than a quorum, then the majority participating and voting at the meeting shall not have line responsibility for operation of the facility. REVIEW 6.5.2.7 The NEDRC is resgonsihle for assuring that independents* reviews of the following are performed: a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of 10 CFR 50.59 to verify that such actions did not constitute an unreviewed safety question. b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59. c. proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59. d. Proposed changes to Technical Specifications or this Operating License. e. Violations of codes, regulations, orders, Technical Specifications, license requirements, or of internal procedures,. or instructions having nuclear safety significance. f. Significant operating abnormalities or deviations form normal and expected performance of plant equipment that affect nuclear safety. g. All REPORTABLE EVENTS h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety related structures,

systems, and components.

i. Reports and meeting minutes of the PNSRC. >> Regular NSDRC members are expected to attend the meeting whenever

possible, and alternates may attend as voting members only on an irregular basis.

Zf both a regular member and his alternate attend a meeting, only the regular member may participate as a voting member, and the alternate is considered a guest.

- Independent reviews may be performed by groups which report directly to the NSDRC and which must have NSDRC membership participation.

D. C COOK - UNIT 2 6-10 Amendment No. 73

ADMINISTRATIVECONTROLS m. The PROCESS CONTROL PROGRAM and implementing procedures for solidification of radioactive wastes at least once per 24 months. n. The performance of activities required by the Quality Assurance Program to meet-the criteria of Regulatory Guide 1.21, Rev. 1, June 1974 and Regulatory Guide 4.1, Rev. 1, April 1975 at least once per 12 months. AUTHORITy 6.5.2.9 The NSDRC shall report to and advise the Vice Chairman, Engineering and Construction,

AEPSC, on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8.

RECORDS 6.5.2.10 Records of NSDRC activities shall be prepared, approved and distributed as indicated below: a. Minutes of each NSDRC meeting shall be prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following each meeting. b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be

prepared, approved and forwarded to the Vice Chairman, Engineering and Construction, AEPSC, within 14 days following completion of the review.

C ~ I Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice Chairman, Engineering and Construction,

AEPSC, and to the management positions responsible for the areas audited within 30 days after completion of the audit.

6 ' FZPOF.ABLE EVENI'CTION 6.6.1 Each REPORTABLE EVENT requiring notification to the Commission shall be reviewed by the PNSRC and submitted to the NSDRC and the Vice President, Nuclear Operations. D. C. COOK - UNIT 2 6-12 Amendment No. 73

ADMINISTRATIVECONTROLS 6.8.3 Temporary changes to procedures of 6.8.1 above may be made provided: a. The intent of the original procedure is not altered. b. The change is approved by two members of the plant management staff, at least one of whom holds a Senior Reactor Operator's License on the unit affected. c. The change is documented, reviewed by the PNSRC and approved by the Plant Manager within 14 days of implementation. 6.9 REPORTING RE UIREMENTS ROUTINE REPORTS 6.9.1 In addition to the applicable reporting requirements of Title 10,- Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator unless otherwise noted. STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manu-factured by a different fuel supplier, and (4) modifications that may - have significantly altered the nuclear, thermal, or hydraulic perfor-mance of the plant. 6.9.1.2 The startup repcrt shall address each of the tests identified in,the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a compariscn of these values with design predictions and specifica-tions. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be in-cluded in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days follriwing resumpt'cn or'cnmenc~ment'f ccirmercial power operatic n, or (3) 9 months following 'nitial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of .startup test program, and resumption or co'mmencement of commercial D. C. COOK UNIT 2 6-14 Amendment No. 73

hDHINISTRATIVE CONTROLS pover operatioa) ~ supplementary reports shall be submitted at least every three months until all three events have been completed. hNNUhL REPORTS 6.9.1. 4 hnaual reports coverfng the acti vftfes of the unft as described belov for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted prior to March 1 of thc year folloving iaftial criticality. 6.9.1.5 Reports required oa an aanual basis shall include: a. h tabulatioa on an annual basis of the number of station, utility aad other personnel (iacluding contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to vork and job functions, e.g., reactor operations aal surveillance, fnscrvice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, aad ref ucliag. The ddsc assignment to various duty functioas may be estimates based on pocket dosimeter, TLD, or ff1m badge mcasuremcnts. Small exposures totalling less than 20X of the individual total dose need aot be accouated for. In the aggregate, at.least SOX of the total vhole body dose received from external sources shall be assigned to specific major vork functions. b. The complete results of steam generator tube inscrvicc inspections performed during the report period (ref ereace'pe cification 4.4.5..5. b) ~ c. Documentation of all challenges to the pressurizer pover operated relief valves (PORVs) or safety valves. 1 h single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all uaf Cs a't 'the station ~ 2 This tabulation supplemeats the requirements of 20.407 of 10 CFR Part 20 ~ D. C. COOK - UNIT 2 6-15 hmeadment No.

l II

ADMINISTRATIVECONTROLS The radioactive effluent release report to be submitted 60 days after January 1 of each year shall also include an assessment of radiation doses to the likely most exposed member of the public from reactor releases and other nearby uran'um fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. l. The radioactive effluent release report shall include the following information for each type of solid waste shipped offsite during the report period: a. Volume (cubic meters), b. Total curie quantity (specify whether determined by measurement or estimate), c. Principal radionuclides (specify whether determined by measurement or estimate), d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms), e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and f. Solidifcation agent (e.g., cement). The radioactive effluent release report shall include unplanned releases from the site'to unrestricted areas of radioactive materials in gaseous and liquid effluent on a quarterly basis. The radioactive effluent release reports shall include any change to the PROCESS CONTROL PROGRAM (PCP) and the OFFSITE DOSE CALCULATION MMJAL (ODCM) made during the reporting period. MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the Director, OffiCe Of Management 'and Program'Analysis, U.S. Nuclear Regulatory Commission, Washington, 'D.C'. 20555, with a copy to the Regional Office no later than the 15th of each month following the calendar month covered by the report. D. C. COOK - UNIT 2 6-18 Amendment No. 73

ADMINISTRATIVECONTROLS SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Regional Administrator within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification: a. ECCS Actuation, Specifications 3.5.2 and 3.5.3. b. Inoperable Seismic Monitoring Instrumentation, Unit No. 1g Specification 3.3.3.3. C ~ Inoperable Meteorological Monitoring Instrumentation, Unit No. 1, Specification 3.3.3.4. d. Fire Detection Instrumentation, Specification 3.3.3.8. e. Fire Suppression

Systems, Specifications, 3.7.9. 1, 3.7.9. 2, 3.7.9.3 and 3.7.9.4.

f. Seismic Event Analysis, Specification 4.3.3.3.2. g. Sealed Source leakage in excess of limits, Specification 4.7.8.1.3. D. C. COOK - UNIT 2 6-19 Amendment No. 73

ADMINISTRATIVECONTROLS 6.10 RECORD RETENTION 6.10.1 years: The following records shall be retained for at least five a. b. Co d. e.f. Records and logs of unit operation covering time interval at each power level. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. ALL REPORTABLE EVENTS submitted to the Commission. Records of surveillance activities, inspections and calibrations required by these Technical Specifications. Records of changes made to Operating Procedures. Records of sealed source and fission detection leak tests and results. Records of annual physical inventory of all sealed source material on record. 6.10.2 The following records shall be retained for the duration of the Facility Operating License: a ~ b. c ~ d. e. go h. k.l. m. n. Records and drawing changes reflecting unit design modifications made to systems and equipment described in the Final Safety Analysis Report. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories. Records of radiation exposure for all individuals entering radiation control areas. Records of gaseous and liquid radioact'ive material released to the environs. Records of transient or operational cycles for those facility components identified in Table 5.7-1. Records of reactor tests and experiments. Records of training and qualification for current members of the Plant staff. Records of in-service inspections performed pursuant to these Technical Specifications. Records of Quality Assurance activities required by the QA Manual. Records of reviews performed for changes made to procedures or equipmentor review of. tests and.experiments pursuant to 10 CFR 50.59 'ecordsof meetings of the PNSRC and the NSDRC. Records for Environmental Qualification which are covered under the provisions of paragraph 6.13. Records of radioactive shipments. Records of the service lives of hydraulic snubbers listed on Table 3.7-9 including the date at which service life commences and associated installation maintenance records. D. C. COOK - 'UNIT 2 6-20 Amendment No. 73

ADMINISTRATIVE CONTROLS F 11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be

approved, maintained and adhered to for all operations involving personnel radiation exposure.

6.12 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.203(c)(2) of 10 CFR 20, each high radiation area in which the intensity of radiation is 1000 mrem/hr or less shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit*. Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following: a. A radiation monitoring device which continuously indicates the radiation dose rate in the area. b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate level in the area has been established and personnel have been made knowledgeable of them. Ce An individual qualified in radiation protection procedures who is equipped with a radiation dose rate'monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit. 6.12.2 The requirements of 6.12.1, above shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem/hr. Zn addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty and/or the Plant Health Physicist. ~ Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they comply with approved radiation % protection procedures for entry into high radiation areas. D. C. COOK UNIT 2 6-21 Amendment No. 73

ADMINISTRATIVECONTROLS 6.13 ENVIRONMENTAL UALIFICATION 6.13.1 By no later than June 30, 1982, all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors "Guidelines for Evaluating Environment Qualification of Class 1E Electrical Equipment in Operating Reactors" (DOR Guidelines)g or, NUREG-0588 "Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment", December 1979. Copies of these documents are attached to Order for Modification of License No. DPR-74, dated October 24, 1980. 6.13.2 By no later than December 1, 1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, -further tested, or otherwise further qualified. l, D. C. COOK - UNIT 2 6-22 Amendment No.

0 0 t' 'g

ADMINISTRATIVE CONTROL/ 6'4 PROCESS CONTROL PROGRAM (PCP) 6.14.1 The PCP shall be approved by the Commission prior to implementation. 6.14.2 Licensee initiated changes to the PCP: 1. Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report fo the period in which the change(s) was made. This submittal shall contain: a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information; b. A determination that the change did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastesg and c. Documentation of the fact that the change has been reviewed and found acceptable by the PNSRC. 2. Shall become effective upon review and acceptance by the PNSRC. 6.15 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.15.1 -6.15.2 The ODCM shall be approved by the Commission prior to implementation. Licensee initiated changes to the ODCM 1. Shall be submitted to the Commission in the Semi-Annual Radioactive Effluent Release Report in the next report after the report period the change(s) was made effective. This submittal shall contain: a. b.'ufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and date box, together with appropriate analyses or evaluations justifying the change(s); A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and ce Documentation of the fact that the change has.beegr-reviewed and found acceptable by the PNSRC. D. C. COOK - UNIT 2 6"23 Amendment No. 73

ADMINISTRATIVECONTROLS 2. Shall become effective upon review and acceptance by the PNSRC. 6.15.3 Commission initiated changes: 1. Shall be determined by the PNSRC to be applicable to the facility after consideration of facility design. 2. The licensee shall provide the Commission'with written noti-fication of their determination of applicability including any necessary revisions to reflect facility design. 6.16 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SySTEMS (Li uid, Gaseous and Solid) 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid): Shall be reported to the Commission in the Annual Operating Report for the period in which the evaluation was reviewed by the (PNSRC). The discussions of each change shall contain: a. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR 50.59'. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental informationg c. A detailed description of the equipment, components and processes involved and the interfaces with other plant systemsg d. An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto> e. An evaluation of the change which shows the expected maximum exposure to individuals in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto'. comparison of the predicted releases of radioactive materials, in liquid and gaseous effluents and in solid &ste, to the actual releases for the period prior to when the changes are to be made'. C. COOK - UNIT 2 6-24 Amendment No. 73

ADMINISTRATIVE CONTROLS g. An estimate of the exposure to plant operating personnel a a result of the changes and 2 ~ h. Documentation of the fact that the change was reviewed and found acceptable by the PNSRC. I Shall become effective upon review and acceptance by the PNSRC. 6.16.2 Commission initiated changes: 1. The applicability of the change to the facility shall be determined by the (PNSRC) after consideration of the facility design. 2. The licensee shall provide the Commission with written notification of its determination of applicability including any necessary revisions to reflect facility design. D C. COOK - UNIT 2 6-25 Amendment No. 73}}

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