Text

Proposed Tech Specs Revising Figure 6.2-1 Re Offsite Organization & Reporting Requirements of Generic Ltr 83-43
	ML17321A357
Person / Time
Site:	Cook
Issue date:	12/17/1984
From:	; INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	;
Shared Package
ML17321A356	List: ML17321A355; ML17321A357;
References
	AEP:NRC:0659C, AEP:NRC:659C, GL-83-43, NUDOCS 8412200250
	Download: ML17321A357 (81)
	v • d • e

ATTACHMENT 2 TO AEP:NRC:0659C PROPOSED REVISED PAGES TO THE DONALD C.

COOK NUCLEAR PLANT UNIT NOS.

1 AND 2 TECHNICAL SPECIFICATIONS 4

84i2200250 84i217 PDR ADOCK 050003i5 P

PDR

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-mati'c 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 qual'itati;ve assessment of channel

~ '"'.

~ behavior during operation by observation; This determination shall

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

D, C.

COOK UNIT 1 I

1-2 Amendment No.

3/4

~ITIHG COHO ITIONS FOR OPERATION AN0 SURVEILLANCE RE UIRBfEHTS 3/4. 0 APPLICABILITY LIMITING CONOITION FOR OPERATION P

3.0.1 Lfaftfng Condftfons for Operation and ACTION requfrements shall be applicable during the OPERATIONAL NOOES or other conditions specified for each specification.

3.0.2 Adherence to the requirements of the Lfmftfng Condition for Oper ation and/or assocfated ACTION within the specfffed tfme interval shall constitute cosplfance with the specfffcatfon.

In the event the Lfaftfng Condition for Operation fs restored prior to expiration of the specified tfme interval, coapletfon of the ACTION statement is not required.

3.0.3 When a Limftfng Condition for'peration fs not met, except as provided fn the associated ACTION requirements, within one hour action shall be initiated to place the unit fn a MOOE in which the Specification does not apply by placfng ft, as applicable, fn 1.

At least HOT STAHGBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

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

At least COLO SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Nere corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time Ifmfts as measured from the time of failure to meet the Lfaftfng Condition for Operation.

Exceptions to these requirements are stated in the fndfvidual Specf fica'ti ons.

3.0.4 Entry into an OPERATIONAl. HCGE or other speci fied applicability condi-tion shall not be made unless the condftfons of the Lfmiting Condition for

'peration are met without reliance on pr'ovisions contained in the ACTION statements unless otherwise excepted.

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

3.0.5 Nen a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source fs inoperable, or solely because fts normal power source is inoperable, ft aay be ccnsidered OPERABLE for the purpose of satisfying the requirements of its applicab1e Limftfng Condition for Operation, provided:

.(1) its corresponding normal or emergency..power, source.,is OPERABLE; and (2) a11 of its'edundant.system(.s);,:

subsystem(s),

train(s),

component('s) and device(s) are OPERABLE, or likewise

". satisfy the requi. aments of this specification.

Unless both conditions (1) and (2)'re satis'ied, within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months action shall. be initiated to place

he unit in a HOOE in wh.'ch he aoplicable Limiting Condition for Operation does not apply by placing it as applicab1e in:

3..

At least HOT STANOSY within.l'e nex 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 2.

At least HOT SHUT"CNf within the following 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months , and 3.

At least COLO SHUTOG'<N within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

This Specification is not applicable in HOOES 5 or 6.

O.

C.

COOK - UNIT I 3/4 0"1

TABLE 4.3-1

.REACTOR TRIP SYSTEM INSTRUMENTATIOH SURVEILLANCE RE UIRENENTS 0

FUNCTIONAL UNIT 1.

Manual Reactor Trip 2.

Power

Range, Neutron Flux 3.

Power Range, Neutron Flux, High Posi tive'.Rate 4.

Power Range, Neutron Flux, High Negative Rate 6.

Intermediate =Range, Heutron Flux-Source

Range, Neutron Flux 7,

Overtemperature hT 8.

Overpower hT 9.

Pressurizer Pressure--Low 10.

Pressurizer Pressure--High 11.

Pressurizer Water Level--High 12.

Loss of Flow..Single Loop CHANNEL CHECK H.A.

N.A.

S S

CHANNEL CALIBRATION N.A.

O(2}, M(3}

and q(6)

R(e)

R(6)

R(e)

CHANNEL FUNCTIONAL TEST S/U(l)

S/U(l )

M and S/U(l)

MODES IN WHICH SURVEILLANCE RE VIREO H.A.

1, 2

0, 2and*

2(7), 3(7),

4 and 5

1, 2

TABLE 4.3-1 Continued NOTATION (1)

(2)

(3)

{4)

(5)

(6)

(7)

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

If not performed in previous.7 days.

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

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

Recalibrate if absolute difference )

3 percent.

Manual ESF functional input check every 18 months.

Each train tested every other month.

Neutron detectors may be excluded from CHANNEL CALIBRATION.

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

~

r4 0,

C.

COOK-UNIT 1

3/4 3-14 amendment No. 76

n TABtf 3.3-4 Continued

.ENGINEEREO SAFETY FEATURE ACTUATION SYSTEH INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT 6.

IIOTOR DRIVEN AUXILIARY FEEDWATER PUHPS ALLOWABLE VALUES a.

Steam Generator, Water Level -- Low-Low b.

4 kv Ous Loss of Voltage c.

Saf'ety Injectinn d.

Loss of tiain Feedwater Pumps 7.

TURBINE DRIVEN AUXILIARY FEEDWATER PUHPS 17% of narrow range instrument span each steam generator 3196 volts with a 2-second delay flot Applicable tlot Applicable 16K of narrow range instrument span each steam generator 3196, +10, -36 volts with a 2+.2 second delay Not Applicable Not Applicable O

V cia a.

Steam Generator Water Level -- Low-Low 4,

ileacto4 Coolant Puiap Ous Und~rvol tpge O,.

LOSS OF POWER a,

4 kv Ous Loss of Voltage b

4 kv Bus Degraded Voltage

> 17K of narrow range instrument span each steam generator

> 2750 Vol tseach bus 3196 volts with a 2-second delay 3596 volts with a 2.0 min. time delay

> 161 or narrow range instrument span each steam generator

> 2725 Volts--each bus 3196, +10, -36 volts with a 2+.2 second delay 3596, +36, -10 volts with a 2.0 minute

+ 6 second time delay

i I

1

<i' I

Pt~

0C Ct ZQKTZQE TESTABLE DURING ISOLATION TIME 51JHRHKW

~

0 0

M 0

0 0

0 ff

0rt m

4) gl I

0) 63.

64.

65.

66.

67; 68.

NCR-107 NCR-108 NCR-109 NCR-110.

NCR-252 PCR-40 73 ~

75.

76.

77

'8.

79.

80.

81.

82.

83.

84.

QCR-919 QCR-920 RCR-100 RCR-101 VCR-10 VCR-11 VCR<<20 VCR-21 XCR-100 XCR-101 XCR>>102 XCR<<103 69.

QCM-250 70.

QCM-350 71.

QCR-300 72.

QCR-301 PRZ Liquid Sample PRZ Liquid Sample PRZ Steam Sample PRZ Steam Sample Primary Water to Pressure Relief Tank Containment Service Air RCP Seal Water Discharge RCP Seal Water Discharge Letdown to Letdown Hx.

Letdown to Letdown Hx.

Demineralized Water Supply for Refueling Demineralized Water 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

. Glycol Supply from Fan Cooler Control Air to Containment Control Air to Containment-Isolation Control Air to Containment Isolation Control Air to Containment Cavity Cavity Yes Yes Yes Yes Yes Yes No No No No Yes Yes Yes Yes Yes Yes Yes Yes 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 R0 1.

2 ~

3 ~

4, 5.

6.

7 ~

8.

9.

10.

CCM-451 CCM-452 CCM-453 CCM-454 CCM-458 CCM-459 ECR-31 ECR-32 ECR-33 ECR-35 ECR-36 CCW from RCP Oil Coolers CCW from RCP Oil Coolers CCW from RCP Thermal Barrier CCW from RCP Thermal Barrier CCW to RCP Oil Coolers 4 Thermal Barrier CCW to-RCP Oil Coolers 5 Thermal Barrier Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor Containment Airborne Radiation Monitor No No No No No No No No No No No 60 60 30 30 60 60 10 10 10 10 10

x'

n00 I

CRH

'7 12.

VCR-205 13.

VCR-206 14.

VCR-207W W1 Upper Comp. Purge Air Inlet Upper Comp.

Purge Air Outlet Cont. Press Relief Fan Isolation

( )

TESTABLE DURINQ JLMGMPJBAXXQK Yes Yes Yes ISOLATION TIME

~~MHER 5

5 5

1.

ICM-111'.

ICM-129 3.

ICM-250 ICM-251 5.

ICM-260 6.

ICM-265 7.

ICM-305 8.

ICM-306 - "

9.

ICM-311

10. ICM-321 11.

NPX 151 VI'-

12.

PA>>3,43 13.

SF-151

14. SF-153

15. SF-159 16.

SF-160 17 ~ SI-171

18. SI-172 RHR to RC Cold Legs RHR Inlet to Pumps Boron Infection Inlet Boron InJeotion Inlet Safety Ingeotion Inlet Safety Ingeotion Inlet RHR Suotion from Sump RHR Suotion from Sump RHR to RC Hot Legs RHR to RC Hot Legs I

Dead Weight Tester Containment Servioe Air Refueling Water Supply Refueling Water Supply Refueling Cavity Drain to Purification System Refueling Cavity Drain to Purifioation System Safety Ingeotion Test Line Aooumulator Test Line Yes No Yes Yes Yes Yes Yes Yes~

Yes Yes Yes No Yes Yes Yes Yes Yes Yes NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA NA

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 0 2.

The pilot cell specific gravity, corrected to 77 F and full electrolyte level, is >1.200, b.

3.,

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

The overall battery voltage is

>250 volts.

At least once per 92 days by verifying that:

1.

The voltage of each connected cell is a 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.

c ~

At least once per 18 months by verifying that:

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.

~I' t

II I

a Pi

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 0 2.

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

~ 1.200, 3.

The pilot cell voltage is

> 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

> 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 abhormal 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 char'ger disconnected.

The battery terminal voltage shall be. maintained p'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.

4 1

rW;,

,V

6.0 ADMINISTRATIVECONTROLS 6. 1 'ESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in'riting the succession to this responsi-bility during his absence.

6.2 ORGANIZATION OFFSIDE 6.2.1 The offsite organization for facility management and technical support shall be as shovn on Figure 6.2-1.

FACILITY STAFF 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 ~

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

Ce At least tvo licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdovn and during recovery from reactor trips.

d.

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

e.

ALL CORE ALTERATIONS 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 go, A 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.

r 1\\

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 Amendment No.

77

CHAIRMANOF THE 8OARD AND CHIEF EXECUTIVEOFFICER AEPSC INDIANA& MICHIGANELECTRIC CO.

AND OTHER AEP SUBSIDIARIES VICE CHAIRMAN ENGINEERING AND CONSTRUCTION AEPSC AND VICE PRESIDENT INDIANA& MICHIGAN ELECTRIC COMPANY

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

EXECUTIVE VICE PRESIDENT AND CHIEF ENGINEER

~ ~ ~ ~ ee VICE PRESIDENT VICE PRESIDENT OPERATIONS AND INDIANA& MICHIGAN NUCLEAR ELECTRIC COMPANY MANAGEROF NUCLEAR OPERATIONS

~ ~ ~ W e

MANAGERS ENGINEERING DIVISIONS AEPSC 8 ~ ~ ~ ~ ~ ~ ~ ~ ~ 1 ~ ~ ~ ~ ~ ~

0

~ ~ ~ OQ MANAGER QUALITY ASSURANCE AEPSC PLANT MANAGER DONALDC. COOK NUCLEAR PLANT

~ ~ ~ ~ ~ ~ 4 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

AEPSC QA

.SUPERVISOR (ONSITE)

I

~ PLANT QC

'UPERINTENDENT I

,~

~ '

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ OO ~

ADMINISTRATIVE& FUNCTIONAL

~

P SU ERVISION TECHNICAL DIRECTION

~ oooo ~ oooo ~ o ~

TECHNICAL LIAISON D.C.COOK - UNXT 1

~ ~ ~ ~ ~ ~

~

FUNCTIONAL DIRECTION 6-2 "IGURE 6.2-1 ORGANIZATIONALRELATIONSHIPS WITHIN THE AMERICANELECTRIC POWER SYSTEM PERTAINING TO QA & QC AND SUPPORT OF THE DONALDC. COOK NUCLEAR PLANT Amendment Uo.

P~

ig

ADMINISTRATIVECONTROLS COMPOSITION 6.5.1.2 The PNSRC shall be composed of the:

Chairman:

Member:

Plant Manager or Designee Assistant Plant Manager Maintenance Assistant Plant Manager Operations Operations Superintendent Technical Superintendent Engineering Technical Supeiintendent Physical Sciences Maintenance Superintendent Plant Radiation Protection Supervisor QC Superintendent 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.

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 majority.

Membership changes resulting from title changes and/or reorganization of responsibilities may be made without prior NRC approval.

The Director, Office of Nuclear Reactor Regulation.shall be notified within 30 days of such changes..'.

D.G.

COOK UNIT 1 6-6 Amendment No.

I f

l

ADMINISTRATIVECONTROLS 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 forwarding of reports covering evaluation and recommendations to prevent recurrence to the Chairmaq of the NSDRC.

gi Review of all REPORTABLE EVENTS.

Review of facility operations to detect potential safety hazards.

h.

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

'he NSDRC.

I

\\

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

D. C.

COOK UNIT 1 6-7 Amendment No.

ADMINISTRATIVECONTROLS 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, IGMECo 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 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 6.5.2.4 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.

QUORUM I

~

I' 6.5.2:6' quorum of the NSDRC shall consist'f a majority of members,'of

whom, no more than a total of two (2) shall be Designated or Temporary Alternates.

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

No more than a minority of the quorum shall have line responsibility for operation of the facility..

Membership changes resulting from title changes and/or reorganization of responsibilities may be made without prior NRC approval.

The Director, Office of Nuclear Reactor Regulation shall be notified within 30 days of such changes.

D. C.

COOK UNIT 1 6-9 Amendment No.

ADMINISTRATIVECONTROLS REVIEW 6.5.2.7 The NSDRC is responsible for assuring that independent 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 in Technical Specifications or licenses.

I e.

Violations of applicable statutes,

codes, regulations,

orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

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

g, All REPORTABLE EVENTS.

h.

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

systems, or components.

1ndependent 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.

).

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 9uality 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 TROLS 6.8.3 Temporary changes to procedure'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 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 report shall address each of the'ests 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.

0 ~

r D. C.

COOK UNIT 1 6-14 Amendment No.

ADMINISTRATIVE CONTROLS power operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below 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 the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their assoc+fed man rem exposure according to work and gob functions, e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignment 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 accounted for.

In the aggregate, at least 80X of the tbtal whole body dose received from external sourcos shall be. assigned to specific ma)or work functions.

b ~

The complete results f

s team generator tube inservice inspections performed during the report period (reference Speci fication 4.4.5.5. b).

C ~

Documentation of all challenges to the pressurizer power operated relief valves (PORVs) or safety valves.

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

The submittal should combine those sections that are common to all units at the station.

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

h D. C.

COOK UNIT 1

6-15 Amendment No. 77

tq T

rt

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 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 Nuclear 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),

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 MANUAL (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 basi's to.'the Director, Office'f 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 foll'owing the calendar month covered by the report.

D. C.

COOK UNIT 1 6-18 Amendment No.

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 ~

b.

Inservice Inspection Program. Review, Specification 4.4.10.

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.

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.

V

~

D. C.

COOK UNIT 1 6-19 Amendment No.

ADMINISTRATIVECONTROLS 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a ~

b.

ce d.

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

a ~

b.

c ~

d.

e ~f.

g, h.

'3

~

'.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.

tl 4 4I lg

ADMINISTRATIVECONTROLS 6.11 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:

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 Radiation 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 1E Electrical Equipment in Operating Reactors" (DOR Guidelines); 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

'th'e environmental qualification method used for all safety-

~

'elated electrical equipment in sufficient detail to document 'the degree of compliance with the DOR Guidelines or NUREG-0588.

Thereafter, such

'ecords 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.

Y SP p

ADMINISTRATIVECONTROLS 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 for 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 wastes; and c.

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

2.

Shall become effective uponreview and acceptance by the PNSRC.

6.15 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.15.1 The ODCM shall be approved by the Commission prior to implementation.

6.15.2 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.

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 change wil'1 not reduce the accuracy or reliability of dose ca'lculations or setpoint determinations; and c.

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

D. C.

COOK.UNIT 1 6-22 Amendment No.

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'l 6.16.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous and solid):

1.

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; c ~

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; d.

An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quant'ity 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; A comparison of the predicted ielea'ses of'adioactive materials, in liquid'nd gaseous effluents 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.

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. 'hall 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.

W 4) p5l

/QO

\\g J 1j(

~

\\

i+

DEFINITIONS REPORTABLE EVENT 1.7 A 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, 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 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 shal3. 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.

'\\

,P "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.

V 'q

'JAN 1'Q 8';

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T 0 200 F 0

av LIMITINGCONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be ) 1.6% A,k/k.

APPLICABILITY:

MODES 1, 2,* 3, and 4.

ACTION:

With the SHUTDOWN MARGIN C,1.6% hk/k, immediately initiate and continue boration at 510 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 ~ 1.6% 8 k/k:

a 0 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.

If the inoperable control 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 untiippable control rod(s).

b.

c 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 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 SS criticality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

See.. Special Test Exception 3. 10. 1 With Keff >~1.0 58 With Keff C 1.0 D. C.

COOK UNIT 2 3/4 1-1

It I

TABLE 4.3-1

=AEACYOA TAIP SYSTEM INSYAUt)ENTATION SUAVEILLANCE AE UIAEtlENYS FUNCTIOtlhL UNIT 1.

manual Aeaqtor Trfp 2.

Power ltange,. Neutron.Flux OthtWEL CilfCK N.A.

CllhtltlEf.

CtthNttEL FUNCTIOtthL CALIOBATIOM TEST ft.h.

Slu(1)

O(2), H(a) and q(6)

BODES Itt 'NIICtf SUAVEILLhtlCE

'- AE UIAED H.A.

1, 2 5.

Power Range, tte'utron Flux, ltfgli J'osftfve Aate Power Itange, Heiitron Flux/

ftfply ttegat fve Rute Intermediate

Range, Neutron Flux

. H.A.

. H;A.

A(6)

R(6)

Slu(>)

1, 2

'1,2 1,

2 and

6.

Source

Range, Neutron Flux 7.

Over tempera ture,'T Q.

Overthrower aT 9.

Pressurfzer Pressure-Low 10.

Pressurfzer Pressure-fffgh ll; Pressurfzer Mater-Level-fffoh 12.

Loss of F'low - Sfngl'e Loop S

S:.

S S

R(e)

R H and S/U(l) 2(7).3(7),4 and 5

fk 1,2 1,2

1. 2

1. 2 1,2

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 in previbus 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 ENGINEEREO'SAFE IY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT.

ALLOWABLE VALUES 6.

MOTOR ORIVEN AUXILIARYFEEOWATER PUMPS a.

Steam Generator Wat.er Level --.Low-Low b.

4 kv Bus Loss of Voltage c.

Safety Injection d.

Loss of Main Feedwat.er Pumps

? 21K of narro~ range instrument. span each steam generator 3196 volts with a 2 second delay

. Not Applicable Hot Applicable

? 20K of narrow range instrument span each steam generator 3196, +lG, -36 volts with a 2 x 0.2 second delay Hot Applicable Hot Applicable 7..TURBINE ORIVEN AUXILIARY FEEOWATER PUMPS a.

Steam Generator Water Level -- Low-Low b.

Reactor Coolant Pump Bus Undervoltage 8.

LOSS OF POWER a.

4 kv. Bus Loss of. Voltage b.

4 kv Bus Degraded Voltage

? 21K of narrow range instrument span each

, steam generator

? 2750. Volts--each bus 3196 volts with a 2 second delay 3596 volts with a 2.0 minute time delay

? 20K of narrow range instrument span each steam generator

? 2725 Volts--each bus 3196, +lf, -36 volts with a 2..i:0..2 second delay 3596, +36,

,ls units with a 2.0 minute x 6 second time delay

n nOO I

~.I O

TABLE 3.3-6 RADIATIOH MONITORING INSTRUMENTATION NIGll ALARM/TRIP SETPOINT MEASUREMEHT IUlNGE MINIMUM CllhNNELS OPERABLE OPERATION MODE/INSTRUMENT ACTION 1

MODES 1 ~ 2 ~ 3~'

4 AREh MONITOR

i. Upp'er.Containment PROCESS'lONITORS

i. Particul'ate a.

-1 4

read ) ng 10 to 10 mR/hr 19

-4 read)ng1.5x10 to 1.5 uCi 20

--7

-2

'.read) ng 10 to 10 uCi/cc 20

~ 2 x normal channel x normal channel 2 x normal channel b.

ii. Noble Gas 2.

MODE 6 a.

TRAlH h.

i. Containment Area Radiation Channel-VRS-2.101 ii. Particulate Channel-ERS-. 2301 iii. Noble'Gas Channel-ERS-2305 22 any 2/3 Channels

<i~

-1 4

< x normal channel reading 10 to 10 mR/hr

<2 x normal channel read)ng 1.5x10 to 1.5 uCi

-7

-2

<2 x.normal channel reading 22 any 2/3 Channels I

b.

TRAIN B

i. Containment brea Radiation Channel-VRS-2201 ii. Particulate Channel-ERS-2401 iii. Hobl'e Gas Channel ERS-2405 Same as 2.a Same aa 2. a Same as 2.a Samp as 2.a Same as 2.a Same as 2.a 3

4 a.

Spent Fuel Storage 10 to 10 mR/hr lg

-1 4

~15 mR/hr

~ With fuel in storage pool or building.

~"

TABLE 4.3-3

.RADIATION"MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS OPERATING MODE/INSTRUMENT 1.

MODES lo 2o-,3o

& 4 a.

AREA MO!:ITOR

i. Upper Containment'.

PROCESS i10NITORS CHANNEL CHANNEL CHECK CALIBRATIOH S

CHANNEL FUNCTIONAL TEST MODES FOR WHICH SURVEILLANCE IS RE UIRED 1,2, 3&4

i. Particulate ii. Noble Gas S

S R

R M

M 1, 2, 3

& 4 1.2l3&4 2,

MODE 6 I

a.

TRAIN'

~

i. Containment Area Radiation Channel ii. Particulate Channel iii. Noble Gas Channel S

S S

R R.

R M

M M

6 6

6 b.

TRAIN B i; Containment brea Radiation Channel ii. Particulate Channel iii. Noble Gas Channel S

S S

R R

R M

M M

6 6

6 3 ~

SPENT Fl:EL STORAGE

With fuel in the. storage pool or building.

4 q(

4I l1

\\

ELECTRICAL POWER SYSTEMS SURVEILLANCE RE UIREMENTS (Continued) 0 2.

The pilot cell specific gravity, corrected to 77 F and full electrolyte level, is > 1.200, b.

3.

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

The overall battery voltage is > 250 volts.

At least once per 92 days by verifying that:

2.

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, 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.

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, and coated with anti-corrosion material.

3.

The battery charger will supply at least 140 amperes at

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

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.

'I

~

t ~

At least'nce per 60'onths, during'hutdown, by verifying that the battery'apacity is at leas't 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.

k 4 )1 I

'L f

C fg A'g

6.0 ADMINISTRATIVECONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this respon>>

sibility during his absence.

6 ~ 2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support shall be as shown in Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a ~

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

b.

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

C ~

At least tvo licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.

e.

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

ALL CORE ALTERATIONS 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.

A 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 mirdmum shift crew necessary for safe shut-dovn of the unit or any personnel required for other essentiiQ.

functions during.a fire emergency.

go The amount of overtime vorked by plant staff members performing safety-related functions must be limited in accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).

D. C.

COOK - UNIT 2 6-1 Amendment No.

r+

j li~ 4 ly

CHAIRMANOF THE BOARD AND CHIEF EXECUTIVEOFFICER AEPSC INDIANA& MICHIGANELECTRIC CO.

AND OTHER AEP SUBSIDIARIES VICE CHAIRMAN ENGINEERING AND CONSTRUCTION AEPSC AND VICE.PRESIDENT INDIANA& MICHIGAN ELECTRIC COMPANY VICE P RESIDENT VICE PRESIDENT OPERATIONS AND INDIANA& MICHIGAN NUCLEAR ELECTRIC COMPANY MANAGEROF NUCLEAR OPERATIONS

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ N EXECUTIVE VICE PRESIDENT AND CHIEF ENGINEER MANAGERS ENGINEERING DIVISIONS AEPSC

~ ~ ~ ~ OO

+ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ tQ MANAGER QUALITY ASSURANCE AEPSC PLANT MANAGER DONALDC. COOK NUCLEAR PLANT

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

AEPSC QA SUPERVISOR (ONSITE)

PLANT QC SUPERINTENDENT

~

P

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ I ~ ~ ~ ~ ~ OO ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 0 ~ ~ 1 ~ ~

ADMINISTRATIVE& FUNCTIONAL SUP ERV IS I ON

'ECHNICALDIRECTION

~ ~ ~ ~ + ~ ~ ~ ~ ~ ~ ~ ~

TECHNICAL LIAISON

~

FUNCTIONALDIRECTION D.C.COOK -UNIT 2 6-2 FIGURE 6.2-1 ORGANIZATIONALRELATIONSHIPS WITHIN THE AMERICANELECTRIC POWER SYSTEM PERTAINING TO QA & QC AND SUPPORT OF THE DONALDC. COOK NUCLEAR PLANT Amendment No.

hp 4

hh

6.0 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 be maintained under the direction of the Training Coordinator and shall meet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1971 and 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 be composed* of the:

Chairman:

Member:

. Member:

, Member.:

Plant Manager or Designee Assistant Plant Manager Maintenance Assistant Plant Manager Operations Operations Superintendent Technical Superintendent - Engineering Technical Superintendent

- Physical Sciences Maintenance Superintendent Plant Radiation Protection Supervisor,,

QC Super'.ntendent e

Membership changes resulting from title changes and/or reorganization of responsibilities may be made without prior NRC approval.

The Director, Office of Nuclear Reactor Regulation shall be notified within 30 days of such changes.

D.C.

COOK UNIT 2

6-5 Amendment No.

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'ny 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.

~UURUM 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.

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.

t h

t

. D. C.

COOK UNIT 2 6-6

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.

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 recommen-dations to prevent recurrence to the NSDRC.

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.

~

W

~

c ~

Provide written,notification within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the NSDRC of disagreement between the PNSRC and the Plant Manager;

however, the Plant Manager.shall 5ave. responsibility', for; resolution of.

'uch 'disagieements pursuant t'o 6.1;1 above.

'..C.

COOK UNIT 2 6-7 Amendment No.

I I

eaS

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

1. 'ice Chairman, Engineering and Construction 2.

Executive Assistant to the President, IGMECo 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 11.

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 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 6.5.2.4 Consultants shall be utilized as determined by the NSDRC Chairman. to provide expert advice to the NSDRC.

6.5.2.5, The NSDRC shall meet at least once per six months.

"Membership changes resulting..from. title changes and/or reorganization of responsibilities may be made without prior NRC approval; The Dir'ector, Office of Nuclear Reactor Regulation shall be notified within 30 days of such changes.

. D. C.

COOK UNIT 2 6-9 Amendment No.

~!

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.

COOK UNIT 2 6-12 Amendment No.

ADMINISTRATIVECONTROLS QUORUM 6.5.2.6 A quorum of the NSDRC shall consist of a majority of 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.

No more than a minority of the quorum shall have line responsibility for operation of the facility.

REVIEW 6.5.2.7 The NSDRC is responsible for assuring that independent*

reviews of the following are performed:

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.

Ct 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.

Violaticns of codes, regulations,

orders, Technical Specifications, license requirements, or of internal procedures or instructions having nuclear safety significance.

Significant operating abnormalities or deviations from 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, or components.

t

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.

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 6.9.1.3.

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 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 Start'up reports sha'll be submitted within (1) 90 days following

'. completion of, the startup test program,'2) 90 d'ays following resumption or'conmencement'f coimercial power operation, or (3)'

months following

'nitial cf.iticality, whichever is earliest.

If the Startup Re'port does not cover all three events (i.e., initial criticality, completion of

.startup test program, and resumption or commencement of commercial D. C.

COOK UNIT 2 6-14 Amendment No.

y

.ADMINISTRATIVE CONTROLS power operation),

supplementary reports shall be submit ted at least every three months until all three events have baen completed.

ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below 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 the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their associated man rem exposure according to work and

)ob functions, e.g.,

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The ddsc assignment 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 accounted for.

In the aggregate, at, least 80X of the total whole body dose received from external sourcos shall be assigned to specific ma)or work functions.

b.

The complete results of s team generator tube inservice inspections performed during the report period (reference Specification 4.4.5..5.b).

C ~

Documentation of all challenges to the pressurizer power operated relief valves (PORVs) or safety valves.

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

The submittal should combine those sections that are common to all units at the station.

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

J I

D..C.

COOK - UNIT 2 6-15 Amendment No.

68

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 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 Nuclear 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),

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 MANUAL (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 P'rogiam'Analysis, U..S. Nuclear Re'gulatory

'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.'.

COOK UNIT 2 6-18 Amendment No.

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. 1, 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.

ADMINiSTRATIVECONTROLS 6.10 RECORD RETENTION 6.10.1 years:

The following records shall be retained for at least five

'a ~

b.

c d.

e.f.

g 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 e b.

c ~

d.

e.

g h.

i ~

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 radioactive 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 equipment. or 'review. of tests, and.experiments pursuant to 36 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 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.

D 6

ADMINISTRATIVECONTROLS 6.11 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.

c ~

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.

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 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.

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); 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.

'I

\\

D C.

COOK UNIT 2 6-22 Amendment No.

f V

fJ

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:

Shall be submitted to the Commission in the semi-annual Radioactive Effluent Release Report for 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 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.

Licensee initiated changes to the ODCM!

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.

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);

F b.'

A'etermination that the. change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c.

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

D. C.

COOK UNIT 2 6-23 Amendment No.

r<

I 4t

}t

~r tf

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; b.

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

A detailed description of the equipment, components and processes involved and the interfaces with other plant systems; 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;,

\\

f;

. A comparison of'he predicted releases of radioactive materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made; D. C.

COOK UNIT 2 6-24

, Amendment No.

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 2 6-25 Amendment No.

ML17321A357

Text

Navigation menu

Search