Text

Proposed Revs to Append a Tech Specs of Subj Facil. Contains 16 Changes W/Appropriate Rev Pages Fire Detec Instru,Flow Rates for Boron Injec Sys,Spray Additive Sys Tech Specs,Applicability of Tech Specs & Editorial Changes
	ML17317A840
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Site:	Cook
Issue date:	12/22/1978
From:	; INDIANA MICHIGAN POWER CO.
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ATTACHMENT 'A'O AEP:NRC:00109 PROPOSED REVISIONS TO THE DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 APPENDIX 'A'ECHNICAL SPECIFICATIONS V812290 y 3~

CHANGE NO.

1 Revision to Table 3.3-10.; "Fire Detection Instrumentation" This change involves a revision to Table 3.3-10 entitled, "Fire De-tection Instrumentation" on page 3/4 3-52.

The minimum number of thermi'stor detectors specified for the Containment quadrants 1, 2, 3 and 4 do.not agree with the as built installation of the thermistor detection system for the containment-cable trays.

This change has been discussed with members of the NRC staff and is consistent with the requirements of the fire protection program for the Donald C.Cook Nuclear Plant.

This change will not adversely affect the health and safety of the public.

CHANGE NO.

2 This change involves a revision to surveillance requirement 4.5.2.f.

We are requesting that the flow rates listed for the Boron Injection System (single pump),

and Safety Injection System (single pump) be revised to assure consistency between the pump design capacities, the plant safety analysis and the technical specifications; These changes will not adversely affect the health and safety of the public.

CHANGE NO.

3 EDITORIAL This change involves a revision to the Bases Section page 8 3/4 4-4.

The reason for this change is to provide a clarifying statement as to how the 52 gpm controlled leakage limitation was accounted for in the accidhnt analysis for the Donald C.

Cook Nuclear Plant.

'CHANGE NO.

4 This change involves revising Surveillance Requirement 4.6.2.2,d of the Spray Additive System Technical Specification (Page 3/4 6-13; Unit 1}

The current surveillance requirement is unworkable as written and this revision will provide better consistency between the intent of the sur-veillance requirement and the design capability of the Spray Additive System.

In addition, the revised flow rates included in the attached page 3/4 6-13 will provide consistency with the flow and pH requirements used in the safety. analysis and also assure that the contents of the Spray Additive Tank are added to the system at the proper rate.

~Ana1 ses have been erformed to show that with a flow,rate from the spray additive tank of 20 to 50 gpm, the pH of the spray solution, will be in accordance with the requirements for the accident analysis in the FSAR.

This change

CHANGE NO.

4 CONT'D is consistent with the functional requirements of the spray additive system included in the safety analysis and will not adversely affect the health and safety of the public..

CHANGE NO.

5 This change involves a revision to the Applicability of Technical Speci-fication 3.9.9 We are requesting that the Applicability be changed from "Node 6" to "During Core Alterations or movement of Irradiated Fuel within the Containment."

The reason for this change is for con-sistency with Specification 3.,9.4 in that 3.9.4 allows certain building penetrations (air locks) to be open while not moving irradiated fuel during trode 6.

Further, since it is not possible to establish contain-ment integrity with the air locks open, both Specifications 3.9.4 and 3.9.9 should be consistent with regard to their Applicability.

This change is consistent with the intent of the Technical Specifications and will not have any adverse affect on the health and safety of the public.

CHANGE NO.

6 This change involves a revision to the definitions section on page 1-5.

Definition 1.22 measures the Reactor Trip System

Response

Time by using the loss of stationary gripper coil voltage.

However, this loss of vol-tage is a result of the reactor trip breakers opening.

We are requesting measurement of the time interval by using the opening of the reactor trip breakers as shown in the attached revised page 1-5.

This change provides better consistency between the Technical Specifications and how the response time interval is actually measured.

When the reactor trip breakers

open, we get a status light indication of no voltage at the stationary gripper coil.

This change has no adverse affect on the health and safety of the public.

CHANGE NO.

7 This change involves a revision to Surveillance Requirement 4.3.3.6.1.

We are presently required to update the incore flux map every 31 days.

However, since the flux is burnup dependent, we request that this be changed to 31 EFPD as shown in the attached revised page 3/4 3-49.

The reason for this change is that a flux map taken every 31 EFPD will be more meaningful in terms of the dependence on accumulated core burnup and the requirements for taking a meaningful flux map.

This change will not adversely affect the health and safety of the public.

CHANGE NO.

8 This change involves a revision to Table 3.6-1.

We have installed an automatic trip (isolation) valve on the return line to the con-tainment from the Containment Air Particulate/Radio Gas Monitors (R-ll & R-12).

This valve is a Phase "B" Containment Isolation

'alve and should be included in the Technical Specifications as shown in the attached revised page 3/4 6-18.

Also note that in order to have the proper numbering sequences, we have revised the valve num-bering'n page 3/4 6-18 thru 6-20 This change will not adversely affect the health and safety of the public.

CHANGE NO.

9 EDITORIAL This change involves a revision to Table 2.2 -1 on page 2-9.

Note 3 indicates "4 percent" and this is an editorial error.

The attached

, revised page 2-9 has corrected this editorial error to read "2 percent."

This change will not adversely affect the health and safety of the public.

CHANGE NO.

10 EDITORIAL This change involves a revision to Table 3.3-1 on page 3/4 3-4.

The words "same loop" should be added to Item 15, under Total No. of Channels as shown in the attached revised page 3/4 3-4.

This change will not adversely affect the health and safety of the public.

CHANGE NO.

11 EDITORIAL This change involves a revision to Bases section 3/4 2.2 on page B3/4 2-4.

In paragraph a.

the word "rod" should be changed to "rods" as shown in the attached revised page B3/4 2-4.

This change is editorial.

This change will not adversely affect the health and safety of the public.

CHANGE NO.

12 This change involves a revision to Figure 6.2-2 "Facility Organization" on page 6-3.

We have had a system wide change in the titles of the staff members of our operating plant.

Certain "Foreman" are now "Supervisors" and "Supervisors" are now "Superintendents."

This change affects the Technical Specifications for the-Cook Nuclear Plant as shown in the attached revised pages 6-3 and 6-4.

4 CHANGE NO.

13 This change revised Technical Specifications 6.5.2.2, 6.5.2.6, 6.5.2.9, 6.5.2. 10,and 6.6.

1 (Pages 6-9, 6-11 and 6-12).

These speci-fications will be amended to indicate the revised NSDRC membership, the number of members/alternates required to constitute a quorum of the

NSDRC, and to clear up minor (editorial) inaccuracies with respect to AEPSC management titles.

The above changes will not adversely effect the health and safety of the public.

CHANGE,NO.

14 CONTAINi~1ENT AIR RECIRCULATION SYSTEMS This change revised Technical Specifications 4,6.5.6(a) and (d) on page 3/4 6-35.

The delay times for the containment air recirculation fan auto-.start and the suction line valve opening time (on auto-start signal) will be changed to 9 +

1 minutes.

We have been informed by Westinghouse that a value of seven minutes was used in the safety analysis for fan-auto start delay time.

The present hydrogen analysis for Unit 2 (FSAR Section 14.3.6-Unit 2 Yellow Pages) assumes a maximum auto;start delay time of ten minutes.

Therefore, the above indicated changes will provide additional margin, in the conservative direction, between the values assumed in the safety analyses and the Technical Specification values.

Page 3/4 3-29 has also been revised to indicate the "new" maximum fan delay time.

The above changes will not adversely effect the Health and Safety of the public.

CHANGE NO.

15 This change deletes specification 6.10.2.C on page 6-20.

Specification 6.10.2.C requires that facility radiation and contamination survey records be retained for the duration of the Facility Operating License.

Deletion of the specification will bring the Cook Plant Technical Specifications in lin'e with the present standardized technical specification (STS) format.

This change will not adversely effect teh health and safety of the public.

CHANGE No.

16 This change revised the footnote to Table 3.2-1.

The work 'increase

'as been replaced with the word 'change'n two locations.

This change will allow for rapid power decreases without violating the pressurizer pressure 'limit'fthe table 3.2-1 (an event which requires an LER be submitted to the NRC).

This change will not adversely effect the health and safety of the public.

ATTACHMENT 'B'O AEP:NRC:00109 PROPOSED REVISIONS TO THE DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 APPENDIX 'A'ECHNICAL SPECIFICATIONS

AEP:HRC:00109 CHANGE NO. A

INSTRUMENTATION FIRE DETECTION INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.7 As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 3.3-10 shall be OPERABLE.

APPLICABILITY: Whenever equipment in that fire detection zone is required to be OPERABLE.

ACTION:

With the number of OPEPABLE fire detection instruments less than required by Table 3.3-10:

Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , establish a fire watch patrol to inspect the zone(s) with the inoperable instrument(s) at least once per hour, and 2.

3.

Restore the inoperable instrument(s) to OPERABLE status within 14 days or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pur suant to Specification 6.9.2 within the next 30 days outlining the action taken, the cause of the inoperability and the plans and schedule for restoring the instrument(s) to OPERABLE status.

The provisions of Specifications 3.0.3 and 3.0.4 are not appl icabl e.

SURVEILLANCE REOU IREMENTS 4.3.3.7.1 Each of the above fire detection instruments shall be demon-strated OPERABLE at least once per 6 months by performance of a CHANNEL FUNCTIONAL TEST.

4.3.3.7.2 'he NFPA Code 72D Class B supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall be demonstrated OPERABLE at least once per 6 months.

D. C.

COOK - UNIT 1 3/4 3-51 Amendment No.

22

1 TABLE 3.3-10 FIRE DETECTIOH INSTRUMENTATION INSTRUMENT LOCATION

'I MINIMUM INSTRUMENTS OPERABLE 1.

Containment Zone 6, Quadrant 1 Cable Tunnel Zone 7, Quadrant 2 Cable Tunnel Zone 8, Quadrant 3N Cable Tunn'el Zone 9, Quadrant 3M Cable Tunnel Zone 10, Quadrant 3S'Cable Tunnel Zone ll, Quadrant 4 Cable Tunnel Quadrant 1

Quadrant 2

Quadrant 3

Quadrant 4

2-HV-CFT-1 Charcoal Filters 1-HV-CFT-2 Charcoal Filters 2.

Control Room Zone 22, Control Room 3.

Cable Spreading Room Zone 15, Switchgear Cable Vault Zone 16, Auxiliary Cable Vault Zone 17, Control Room Cable Vault Zone 18, Control Room Cable Vault 4.

Diesel Generato~

Diesel Generator Room lAB Diesel Generator Room 1CD 5.

Diesel Fuel Oil Room SMOKE IOHIZATION 3

5 3

3 2

5

.10 5

24 25 HEAT THERMISTOR a4 3

as aa 1

1 D. C.

COOK - UNIT 1

3/4 3-52

AEP:HRC:00109 CHANGE NO.W

EMERGENCY CORE COOLING SYSTEitS SURVEILLANCE REQUIREHENTS (Continued)

By performing a flow balance test during shutdown following completion of modifications to the ECCS subsystem that alter the subsystem flow characteristics and verifying the following '

flow rates:

Boron Injection System Safety Injection Systen Sin le Pumo "

Sin le PumN*

Loop 1 Boron Injection Fl ow 117. 5 gpm Loop 2 Boron Injection Flow 117.5 gpm Loop 3 Boron Injection Fl ow 117. 5 gpm Loop 4 Boron Injection Flow 117.5 gpm Loop 1 and 4 Cold Leg Flow ~ 300 gpm Loop 2 and 3 Col d Leg FlOW ~

300 gpm The flow rate, in each Boron Injection (BX) line should be adjusted to provide 117.5 gpm (nominal)'low into each loop.

Under these conditj.ons there is zero mini-flow and 80 gpm simulated RCP seal injection line flow.

The actual flow in each BX line may deviate from the

- nominal so long as the difference between the highest and lowest flow is 10 gpm or less and the total flow to the four branch lines does not exceed 470 gpm.

5linimum flow (total flow) required is 345.8 gpm to the three

'ost conservative (lowest flow) branch lines.

- ,Total SIS (single pump) flow, including miniflow, shall not exceed 650 gpm.

O.

C.

COOK - UNIT 1

3/4 5-Ga Amendment No..23

AEP:NRC:00109 CHANGE NO. 3

~

S'

REACTOR COOLANT SYSTEH BASES adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents.

Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown.

Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with the all volatile treatment (AVT) of secondary coolant.

However, even if a defect of similar type should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required of all tubes with imperfections exceeding the plugging limit which, by the definition of Specification 4.4.5.4.a is 405 of the tube nominal wall thickness.

Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 205 of the original tube wall thickness.

Mhenever the results of any steam generator tubing inservice inspec-tion fall into Category C-3, these results will be promptly reported to the Commission pursuant to Specification 6.9.1 prior to resumption of plant operation.

Such cases will be considered by the Commission on a

case-by-case basis and may result in a requirement for analysis, labora-tory examinations,

tests, additional eddy current inspection, and revision of.the Technical Specifications, if necessary.

3/4.4.6 REACTOR COOLANT SYSTEH LEAKAGE 3.4.4.6.1 LEAKAGE OETECTION SYSTEHS

~

The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.'hese detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Oetection Systems",

Hay 1973.

3/4.4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the

RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1

GPH.

This threshold value is sufficiently. low to ensure early detection of additional leakage.

D.

C.

COOK-UNIT 1

B 3/4 4-3

~r" e REACTOR COOLANT SYSTEM BASES The 10 GPM IDENTIFIED LEAKAGE limitation provides allowance for a limited amount'of leakage from known'ources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

0 The CONTROLLED LEAKAGE limitation restricts operation when the total flow supplied to the reactor coolant pump seals exceeds 52 GPM..

This limitation is based on the maximum seal injection flow capability of the Reactor Coolant Pumps and ensures a

maximum safety injection flow assumed in the accident analysis.

The total steam generator tube leakage limit of 1

GPM for all steam generators not isolated frm the RCS ensures that the dosage contribution from the tube leakage will be limited 'to a small fraction of Part 100 limits in the event of either a steam generator tube rupture or steam line bWeak, The 1

GPM limit is consistent with the assumotions used in the analysis of these accidents.

The 500 gpd leakage limit per steam generator ensures that steam generator tube integrity is maintained in the event of a main steam line rupture as under LOCA conditions.

PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may be indicative of an impending gross failure of the pressure boundary.

Should, PRESSURE BOUHDAPY LEAKAGE occur through a component which can be isolated from the balance of the Reactor Coolant: System, plant operation may continue provided the leaking component is promptly isolated from the Reactor Coolant System since isolation removes the source of potential failure.

3/4.4.7 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or, failure due to stress corrosion.

Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. 'he associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent.

Corrosion studies show that operation may be, continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified. limited time intervals without. having a significant effect on the structural'.integrity of the Reactor Coolant System.

The time'interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

D-. C.

COOK-UNIT 1 B 3/4 4-4

AEP:HRC:00109 CHANGE NO.

CONTAIi'li~lENT SYSTEMS SURVEILLANCE REOUIREllENTS (Continued) c.

At least once per 18 months during shutdown, by:

1.

Cycling each power operated (excluding automatic) valve in the flow path that is not testable during plant operation, through at least one complete cycle of full travel.

2.

Verifying that each automatic valve in the flow path actuates to its correct position on a Containment Pressure High-High signal.

d.

At least once per 5 years by verifying a water flow rate of

~ 20gpm and ~ 50gpm from the spray additive tank to each containment spray system with the spray oump operating in the" recirculation mode ih th a'ump dis-charge pressure 0 225 psig.

D.

C.

COOK-UNIT 1

3/4 6-13,

~

t(

~l CONTAINHEHT SYSTEMS 3/4.6.3 CONTAINMEHT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.3.1 The containment isolation valves specified in Table 3.6-1 shall be OPERABLE with isolation times as shown in Table 3.6-1.

APPLICABILITY:

MODES 1, 2, 3 and 4.

r',

ACTION:

With one or more of the isolation valve(s) specified in Table 3.6-1 inoperable, either:

a.

Restore the inoperable valve(s) to OPEPABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one deactivated automatic valve secured in the isolation position, or co Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one closed manual val.ve or blind flange; or Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in'COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE RE UIRB1EHTS 4.6:3.1.1 The isolation valves soecified in Table 3.6-1 shall be demonstrated OPERABLE:

a.

. At least once per 92 days by cycling each OPERABLE power operated or automatic valve tes.able during plant operation through at least one complete cycle of full travel.

b.

Imnediately prior to returning the valve to service after maintenance, repair or replacement work is performed. on the 0.

C.

COOK-UNIT 1

3/4 6-14

AEP:t<RC:00109 CHAHGE NO.~

I

~

0

~

REFUELING OPERATIONS COOLANT CIRCULATION LIMITING CONDITION FOR OPERATION 3.9.8 At least one residual heat removal loop shall be in operation.

APPLICABILITY:

gODE g ACTION:

1 r'ith less than one residual heat removal loop in operation, except as provided in b. below, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System.

Close al'1 containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

b.

The.residual heat removal loop may be removed from operation for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period during the performance of CORE ALTERATIONS in the vicinity of the reactor pressure vessel hot legs.

c.

The provisions of Specification 3.0.3 are not applicable; SURVEILLANCE REQUIREt'EHTS 4.9.8 A residual heat removal loop shall be determined to be in operation and circulating reactor coolant at a flow rate of > 3000 gpm at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

0.

C.

COOK -'NIT 1 3/4 9-9

REFUEL IHG OPERATIONS CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIl1ITIHG CONDITION FOR OPERATION

.3.9.9 The Containment Purge and Exhaust isolation system shalT be OPERABLE.

APPLICABILITY:

During CORE ALTERATIONS or movement of irradiated fuel within the containment.

ACTION:

With the Containment Purge and Exhaust isolation system inoperable, close each of the Purge and Exhaust penetrations providing di'rect access from the containment atmosphere to the outside atmosph'ere.

The provision of Specification 3.0.3 are not applicable.

~ SURYEILLAHCE REQUIREMENTS 4.9.9 The Containment Purge and Exhaust isolation system shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment Purge and 'Exhaust isolation occurs on manual initiation and on a high radiation signal from each of the containment radiation monitoring instrumentation channels.

D. C. 'COOK - UNIT 1

3/4 9-10

AEP: t<RC: 00109 CHANGE NO.~

DEFINITIONS 0

FRE UENCY NOTATION 1.21 The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 1.2.

REACTOR TR!P SYSTEN RESPONSE TINE 1.22 The REACTOR TRIP SYSTEN

RESPONSE

TINE shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until the reactor trip breakers open.

ENGINEERED SAFETY FEATURE RESPONSE TINE 1.23 The ENGINEERED SAFETY FEATURE RESPONSE TINE shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their re-quired positions, pump discharge pressures reach their required values, etc.).

Times shall include diesel generator star ting and sequence loading delays where applicable.

AXIAL FLUX DIFFERENCE 1.24 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals betwe n the top and bottom halves of a two section excore neutron detector.

PHYSICS TESTS 1.25 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear charac.eristics of the reactor core and related instrumentation and

1) described in Chapter 13.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Comnission.

E - AVERAGE DISINTEGRATION ENERGY 1.26 E shall be the average (weighted in proportion to the concentration'f each radionuclide in the reactor coolant at the time of'ampling) of the sum of the average beta and gamma energies per disintegration (in NeV) for isotopes, other than iodines, with half lives greater than 15

minutes, making up at least 95~ of the total non-iodine activity in the coolant.

D..C.

COOK - UNIT 1

TABLE 1.1 OPERATIONAL MODES NOOE 1.

POWER OPERATION 2.

STA/TUP 3.

HOT STANDBY 4.

HOT SHUTDOWN 5.

COLD SHUTDOWN 6.

REFUELING~.

REACTIVITY CONDITION, K

I

> 0.99

< 0.99

< 0.95

'K RATED THERMAL POWER*

> 5X 0

0 0

AVERAGE COOLANT TEMPERATURE

> 350'F

> 350'F 350'F

> Tay

> 200'F

< 200'F

< 140'F g

YhA.

~ Reactor vessel head unbolted or removed and fuel in the vessel.

~ ~

D. C.

COOK - UNIT 1

1-6

1 Aep: tee: ooio9 CHANGE NO.~

s s

s INSTRUI 'IEHTATIOH AXIAL POWER DISTRIBUTION YIOHITORIING SYSTEi'l LINITIHG CONDITION FOR OPERATION 3.3.3.6 The axial power distribution monitoring system (APDHS) shall be OPERABLE with:

a.

At least two detector thimbles available for which R has been determined from full incore flux maps.

These two thimbles shall be those having the lowest uncertainty, a, covering the full configuration of permissible rod patterns permitted at RATED THER'"AL POllER.

b.

At least two movable detectors, with associated devices and readout equipment, available for mapping F (Z) in the above required thimbles.

APPLICABILITY:

When the APDNS is used for monitoring the axial power ACTIDII: Mith the APDIIS inoperable, do not use the system for determining the Axial Power Distribution:

The provisions of Specification 3.0.3 are not appl icabl e.

SURYEILLAHCE RE UIRENEHTS 4.3.3.6.1 The full incore flux maps used to d termine g and for monitor-ing F.(Z) shall be updated at least once per 31 EFPD.

The continued accuracy and representativeness of the selected thimbles shall be verified by using their latest flux maps to update the R for each representative thimble.

The orignial uncertainty, a, shall not be updated, except. as follows:

"Except as provided in Specification 4.2.6.1.b.

3TThe APDt<S may be out of service:

1) when incore maps are being taken as part of the Augmented Startup Test Program, or 2) when surveillance for determining power distribution maps is being performed.

'I D. C.

COOK-UNIT 1 3/4 3-49

INSTRUMENTATION SURVEILLANCE RE UIRBIENTS (Continued) a ~

~

~ ~

~

If the abso1ute va1ue of is greater than Za., another j

s J

map shall be completed to verify the new K.. If the second map showgthe first to be in error, the first hap shall be dis-

'egarded.

-If the second map confirms the new R., four more maps

.. (including rodded configurations allowed by the~insertion limits) will be completed so'that a new R

and a

can be defined from the six new maps.

4.3.3.6.2 The APONS shall be demonstrated OPERABLE:

a.

By performance of a CHANNEL FUNCTIONAL TEST within 7 days prior to its use and at least once per 31 days thereafter when used for monitoring.F.(Z).

b.

At least once per 18 months, during shutdown or below Gll of RATED'THERMAL PO';lER, by performance of a CHANNEL CALIBRATION.

0.

C.

COOK-UNIT 1

3/4 3-60

AEP: HRC: 00109 CHANGE NQ.~

VALVE HUMBER PHASE "A" ISOLATIOH FUHCTIOH Continued r'ABLE 3.6-1 Continued I

. TESTABLE DURIHG ISOLATIOH TI)1E PLAt<T OPERATIOt)

It)

SECONDS'7.

28.

29.

30.

31.

32.

33.

34.

35.

36.

37.

38.

39.

40.

41.

42.

43.

44.

45.

46.

47.

48.

49.

50.

51.

52.

53.

54.

55.

56.

ECR-10 ECR-11 ECR-12 ECR-13 ECR-14 ECR-15 ECR-16 ECR-'l7 ECR-18 ECR-19 ECR-20 ECR-21 ECR-22 ECR-23 ECR-24 ECR-25 ECR-26 ECR-27 ECR-28 ECR-29 GCR-301 GCR-314 ICR-5 ICR-6 t1CR-251 MCR-252 h1CR-253 MCR-254 HCR-105 t<CR-106 Cont.

Hz Sample Return Cont.

kl2 Samole - Air to Rec.

E Cont.

H Sample - Air from Rec.

E Cont.

kiz Sample - Low. Cont. Vol.

Cont.

ll2 Sample - Low Cont. Vol.

Cont.

H2 Sample - Up Cont. Vol.

Cont.

klan Sample - Up Cont. Vol.

Cont.

H2 Sample - Air to Rec. lt Cont.

H2 Sample - Air from Rec.

W Cont.

ll2 Samole - Cont.

Dome Vol.

Cont.

Il2 Sample-Return Cont.

H2 Sample - Air to Rec.

E.

Cont.

ll2 Sample - Air fr. Rec.

E Cont.

ll2 Sample - Lo~( Cont. Vol.

Cont.

ll2 Sample - Low Con'.. Vol.

Cont.

ltd Sample - Un Coni. Vol.

Cont.

ll2 Samole - Up Cont. Vol.

Cont.

Il2 Sample - Air to Rec.

H.

Cont.

H2 Samnle - Air Fr. Rec.

H.

Cont.

ktz Sample - Cont.

Dome Vol.

H2 Supply to Pressurizer Relief Tank H2 Supply to Accumulators Accumulators Sample Accumulators Sample Samnle Line from Steam Gen. Outlet IIl Sample Line from Steam Gen. Outlet It'2 Sample Line from Steam Gen. Outlet 83 Sample Line from Steam.Gen.

Outlet II4 llot Leg Sample.

Hot Leg Sample Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes 10 10 10 10 10 10 10 10 10 10 10 10.

10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10

('D n

VALVE NUMBER FUNCTIOH TABLE 3.6-1 (Continued TESTABLE OURINCi PLANT OPERATION ISOLATION TINE IN SECONOS CD CD I

10.ll.

12.

- 13.

14.

15..

16.

17.

18.

19..

21.

22.

23.

24.

25.

26.

27.

28.

29.

30.

31.

32.

33.

34.

35.

36.

37.

PHASE "B" ISOLATION HCR-901 llCR-903 klCR-905 HCR-907 llCR-909 WCR-911 HCR-913 l<CR-915

>lCR-921 HCR-923 tiCR-925 HCR-927 HCR-929 tlCR-931

'klCR-933 HCR-935 HCR-945 tlCR-946 HCR-947 1lCR-948 MICR-951 HCR-952 HCR-953 llCR-954 LKR-955 lfCR-956 1lCR-957 WCR-958 Continued HESll to Low Containment Vent ¹1 NESll from Low Containment Vent ¹1 HESH to Low Containment Vent ¹2 NESTS! from Low Containment Vent lI2 HESH to L'ow Containment Vent ¹3 HESH from Low Containment Vent ¹3 HESll to Loiv Containment Vent ¹4 HESl< from Low Containment Vent ¹4 HESH to Up Containment Vent ¹1 HESll from Up Containment Vent ¹1 HESLl to Up Containment Vent ¹2 HESLI from Up Containment Vent ¹2 HESW to Up Containment Vent ¹3 HESH from Up Containment Vent ¹3 HEStl to Up Containment Vent ¹4 HESH from Up Containment Vent ¹4 HESll from PCP Motor Air Cooler HESH from RCP tlotor Air Cooler HESll from RCP Motor Air Cooler HEStl from RCP Potor Air Cooler HESH to RCP llotor Air Cooler Vent ¹1 HESH to RCP Motor Air Cooler'Vent ¹2 HESll to RCP Motor Air Cooler Vent ¹3 HESH to RCP Motor Air Cooler'ent

¹4 HFSH from RCP Motor Air Cooler Vent ¹1 HESH from RCP Motor Air Cooler Vent ¹2 HESll from RCP Motor Air Cooler Vent ¹3 HESH from RCP ltotor Air Cooler Vent ¹4 Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10 10

CD n

CD CD I

VALVE HUtSER PtlASE "8" ISOLATION FUNCTION Continued TABLE 3.6-1 (Continued TESTABLE DURING ISOLATION TIME PLANT OPERATIOn IH SECONDS Ch I

CD 38.

39.

$ 0.'l.

42.

43.

44.

45.

46.

47.

48.-

49.

50.

51.

HCR-961

'l!CR-963 HCR-965 HCR-967 HCR-902 HCR-906 l!CR-910 HCR-974 HCR-922 HCR-926

'HCR-930 llCR-934 WCR-962 lKR-966 ttESlt to Instr.

Rm. East Vent HESH from Instr.

Rm. Hest Vent tlESH to Instr, Rm. Fast Vent t(ESH from Instr.

Rm. Hest Vent HESH from Lower Containment Vent.81 HES'lt from Lower Containment Vent 82 HESlt from Lower Containment Vent //3 HESlt from Lower Containment Vent I/O

=

NESH from Upoer Containment Vent /Il ttESlt from Upper Containment Vent 82 HESH from Upper Containment Vent //3 HESlt from Upner Contain;!!ent Vent //4 HESH from Instrument Room East Vent flESlt from Instrument Room Hest Vent Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes 10 10 10

'0 10 10 10 10 10 10 10 10 10 10 C.

CottTAIHtlENT.PURGE AHD EXtWUST l.

2.

3.

5.

6.

~

7.

0.

9.

10.

11.'CR-101 VCR-102 VCR-103 VCR-104 VCR-105 VCR-106 VCR-107*

VCR-201 VCR-202 VCR-203 VCR-204 Instr.

Room Instr; Room Lower Comp.

Lower Comp.

Upper Comp.

Cont. Press.

Instr.

Room Instr..Room Lower'Comp.

Lower Comp.

Purge Air Inlet Purge 'Air'Outlet Purge Air Inlet Purqe Air Outlet Purqe Air Inlet Purge Air Outlet Relief Fan Isolation Purge Air Inlet Purqe Air Outlet Purqe Air Inlet Purge Air Outlet Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes Yes 10 10

't0 10 10 10 10 10 10 10 10

AEP:NRC:00109 CHAli'GE NO.

Note 2:

TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUHENTATION TRIP SETPOINTS NOTATION Continued v3S Overpower eT

< eT [K4-K<

>~,

< T - K6 (T-T")-f>(dI)]

'"3'here:

4T

=

Indicated LT at rated power 0

T

=

Average temperature,

'F T"

~

Indicated T

at RATED THEfML POWER

< 567.8'F K4 K5 K6 v3S

~+v3 f2(aI) 1.075 0.0177/'F for increasing average temperature and 0 for decreasing average temperature 0.0012 for T > T"; K6 = 0 for T < T" The function generated by the rate lag controller for T dynamic compensation avg Time constant utilized in the rate lag controller for T

= 10 secs.

3 avg Laplace transform operator fl (hI) as defined i'n Note 1 above.

Note 3:

The channel's maximum trip point shall not exceed its computed trip point by more than 2 percent

,excluding transmitter.

AEP:HRC:00109 CHAf<CiE NO. /d.

TABLE 3.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION

./

FUNCTIONAL UNIT 8.

Overpower hT Four Loop Oper ation Three Loop Operation 9.

Pressurizer Pressure-Low 10.

Pressurizer Pressure

High ll.

Pressurizer plater Level--High 12.

Loss of Flow - Single Loop (Above P-8) 13.

Loss of Flow - Two Loops (Above P-7 and below P-8)

TOTAL NO.

OF CHANNELS 3/loop 3/loop CHANNELS TO TRIP 2/loop in any oper-ating loop 2/loop in two oper-ating loops HINIHUM CHANNELS APPLICABI E OPERABLE NODES 1,

2 1.

2 1,

2 3

1, 2

2 1,

2 2/loop in 1

each oper-ating loop 2/loop 1

each oper-ating loop ACTION 2

O 6

7

r '\\ll

'" ~

~

P d

)

~

~po il II a

~

ll

~

a e

aa

~

~ e yt 4

~

~ a a

I

~,

e ~

~

I'pl

~ a

're ta a M

<'~e'~;,f

~

/

~ a

~

~tI

~ o p a

~ v

~

P 1

1 TASLE 3. --1 (Contira<<ea'i.

RERiNII TR'.II ET'".TEI IIISTIUll'.i>'ilITIOII Q

FUHCTIOHAL UHIT 14.

Steam Generator Mater Level--Low-Low.

TOTAL HO.

OF CUAHHELS 3/loop a'h CHAHHELS TO TRIP 2/loop in any oper-ating loops

~

\\s a

15.

Steam/Feedwater Flor~',

2/1 oop-1 evel 1/ loop-1 evel Nismatch and LoH Stc.,m an<i.

coirici.ent Generator Water Lev"';",

2/1 oop-f n

<;.i ':h mismatci":,~n 1/1 c oj;-',i 1 o,>;

if Q, I

~ 'l.ia, 1 g.

~

1 e, 6

', ',t.;BC'L'll'ia;lg~mP'T.,m) S)',t>$ ~,,

. )l

~

P

~,

a ~

. )l

I"

~ ~

~

Vq I ~

'a

~

a t

1)a r

~'I

)6.

Undenvalta'ge aRea-ttotT)oo'lant 4-1/boa 2

Pumps 17.

Underfrequency-Reactor Coolant 4-1/bus 2

Pumps 18.

Turbine Trip

. A.

Low Fluid Oil Pressure 3

B.

Turbine Stop Valve Closure 4

HIH If)UN

~

CHAHHELS ='.APf'LICABLE OPERABLE NODES 2/loop iH 1,

2 each oper'.-'

'ating loop 1/loop-1 evael

'1, 2

and

.'

'/1

~top-f-,;~.J; ml;=,().ItGj'i:f;+

f 9/$ l Qa)pa 'hi+1 e "s

~ ~p Iw ( ',

1/loop-f]a.)

mismatch 1.

P.a.

2 4

7 8

-a l

~

19.

Safety Injection Input from ESF 2

1, 2

AEP: NRC: 00109 CHANGE NO.~l

Percen Therma 100.o t of Rated I Power

"'I+

5%

9CA I g 80%

70%

V t

T 30~

-20%

-10'.f 0

+10'30%

INDICATED AXIALFLUX DIFFERENCE Figure 0 3/4 2.1 TYPICAI INDICATEO AXIALFLUX DIFF:: '.'!CE VERSVS THERMAL POWER AT POL D. C. COOI'-UNlT 't B 3/4 2-3

POWER OISTRIBUTIOH LIHITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AHO HUCLEAR Ei'/THALPY HOT CHAHilEL FACTORS-F (Z) and F,H The limits on heat flux and nuclear enthalpy hot channel factors ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200'F ECCS acceptance criteria limit.

Each of these hot channel factors are measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the hot channel factor limits are maintained provided:

a.

Control rodsin a single group move together with no individual rod insertion differing by more than + 12 steps from the group demand position.

b.-

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.5.

Ca d.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the limits.

The relaxation in F

H as a function of THERtCL POWER allows chanqes in the radial power.shape for all permissible rod insertion limits.

F'ill be maintained within its limits provided conditions a thru d abokL, are maintained.

When an F

measurement is taken, both experimental error and man-ufacturing tolerance must be 'allo~ed for.

5< is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3Ã is the appropriate allowance for manufacturing tolerance.

When F<H is measured, experimental error must be allowed for and 4/

is the appropriate allowance for a full core>map taken with the incore detection system.

The specified limit for F'

also contains an 8X allo'~~ance for uncertainties which mean that normal operation will result in F<H < 1.51/1.08.

The 8'~ allowance is based on the following considera-tions:

0.

C.

COOK-UNIT, 1 B 3/4 2-4

AEP:NRC:00109 CHANGE NO. 1Q

Cl n

~ pLANT 51AflAG AOMltJISTAATIVE MAINTENANCE ~

OPEAATION ~

sUPERvlsoA Supt.

Suot.

SOL TfCIDJ ICAL Supt.

STAFF STAFF PROD. SUPV.

OPEAATIONS SOL OA SUPERVISOR TAAINING COOADINATOA SIII F'T OPKAATING ENG.

OP E BATING ENGIIJEEA OL

'UCLEAR

~

KNGItiE'KA PE:IFCBthANcE coNTPDL AND P lant ']ant

~

sUPEBYIMA INsTBMIEIJTATIDN

)heAIlca1 ad'tati on lNGI'JKEI'IJGltJ<-+A upv.

Prot.Supv LKGEND:

COL SEti!OA OPEAATOA LICENSE OL - OPERATOR LICENSE i - KKYSUPEAVISORY I'EASONNEL UtJIT Sgtlu.

PEBFOB!>IAIJCE E<JGINEEB PERFCR:<IAIIC ENGltifEA ItJSTRUIZCNT t<IAliJTKIJAtiCE Sl'Inv.

CIIKMIST BACIATIOti PROTECTION Supv.

EOUIPXIKNT CPE BATOA TECIJIJICIANS TECIII JIG l ANS O

AUK<ILIAAY EOVIP><IKtJT OPERATOR FlGURE 6.2-2 Fcdllty Orgcn!athn Dontl'8 C. Cook Unit No. 0

~ ~

I W

~ ~

AEP:NRC:00109 CHANGE NO. /~~

AOl1INISTRATIVE CONT COMPOSITION 6.5.2.2 The NSDRC shall be composed of the:

Vice Chairman'Engineering and Construction Senior Executive Vice President Engineering Senior Vice President Construction Executive Vice President Indiana

&.t1ichigan Electric Company Vice President Electrical Engineering Vice President Mechanical Engineering Assistant Vice President and Chief Civil Engineer

, Chief Nuclear Engineer (Chairman)

Chief Design Engineer Plant Manager, Donald C.

Cook Plant Head Environmental Engineering Division

Head, Nuclear Safety

& Licensing Section (Secretary)

Alternate:

Executive Assistant to the Vice Chairman Engineering

& Construction Alternate: Assistant Division Head, Project Control and Support Division Alternate:

Executive Assistant to the Executive Vice President I

& M Alternate: Assistant Chief flechanical Engineer Alternate: Assistant Chief Civil Engineer Alternate: Assistant Division Head, Nuclear Engineering Division Alternate:

Head, Electrical Plant Design Section Alternate: Assistant Plant Manager, Donald C.

Cook Plant Alternate:

Senior Staff Engineer, Environmental Engineering Division Alternate:

Engineer, Nuclear Safety

& Licensing Section Alternate:

AEPSC flanager of equality Assurance Alternate: Assistant Chief Electrical Engineer ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NSDRC Chairman to serve on a temporary basis;

however, no more than two alternates shall participate as voting m mbers in NSDRC activities at any one time.

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

MEETING.FREOUENCY 6.5.2.5 The NSDRC shall meet at'east once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.

gila RUM 6.5.2.6 A quorum of NSORC shall consist of the Chairman or his designated.

alternate and,at least 4

NSORC meinbers including alternates.

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

D. C.

COOK -'NIT 1 6-9 f

0

.AO.~INISTRATIYE CONTPOLS REYI EM 6.5.2.7 The NSDRC shall review:

a 0 b.

C.

d'.

e.

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

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.

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

Proposed changes in Techni cal Speci ficati ons or 1 icenses.

Violations of applicable statutes,

codes, regulations,

orders, Technical Specifications, license requirem nts, or of interna'l procedures or instructions having nuclear safety significance.

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

REPORTABLE OCCURRENCES requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . notification to the Cormission.

h..

All recognized indications of an unanticipated deficiency in some aspect of design or. operation of safety related st'ructures,

systems, or components.

Reports and meetings minutes of the PNSRC.

D.

C.

COOK - UNIT 1 6-10

ADl)INISTRATIV'E CONTROLS REVI EM 6.5.2.7 The NSDRC shall review:

a ~

b.

C.

d.

e.

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

I

~

Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section 50;59, 10 CFR.

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

Proposed changes in Technical Specifications or licenses.

Yiolations 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.

REPORTABLE OCCURRENCES requiring 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification to the Cooxni ssion.

h.

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

systems, or components.

'.Reports and me tings minutes of the PNSRC.

0.

C.

COOK - UNIT 1 6-10

~ g *

,~

0 ADl!Ir')ISTRATIYE CONTROLS

'UOITS I

6.5.2.8 Audits of facility activities shall be per formed under the.

cognizance of'the NSORC.

'These.'audits shall encompass:

a.

The conformance of facility ooeration to provisions contained within the Technical Specifications.

nd applicable license conditions at least once per 12 months.

b, The performance, training and qualifications of the entire facility staff at least once per 12 months.

c.

The results of actions taken to correct deFiciencies occurrir.g in facility equipment, structures, systems or method of >peration that af.ct nuclear safety at least once per-6 monthse d.

e.

The performance of activities required by the guality Assur ance Program to meet the criteria of Appendix "8", 10 CFR 50, at

.least once per 24 months.

The Facility Eroergency Plan and implementing procedures at least once per. 24 months.

The'Facility.Security Plan and implementing procedures at least once per 24 months'.

Any other area of facility operation considered appropriate by the

HSDRC, h.

3 ~

The Facility Fire Protection Program and implementing. procedures at least orce per 24 months.

An independent fire protection and loss prevention program inspection and audit shall be per ormed at least once per 12 months utilizing either qualified offsite licensee personnel or an outside fire protection firm..

An inspection and audit of the fire protection ard loss pre-vention program shall bc performed by a qualified outside fire consultant at least once per 36 months.

AUTl)ORITY 6.5.2.9 The HSORC shall report to and advise the Yice

Chairman, Engineeri.na and Constrrrrtinn, Ar-:PSC nn thnse areas of respor~sibility specified in Sections 6.5.2.7 and 6.5.2.8.

D, C.

COOK' UllIT 1 6-11 'mendment No.

22

0 I I

~

I

~

I I

I

~

~ ~

~

I

~

I

~

I

~

I

~

~ ~

~

I

~

I

~

I o

I I

~

I

~

I

~

~ ~

~

I I

~ ~

~

I

~

I

~ ~

I I

I

~

~ I

~

~ ~

~

~ l

~

I

~

~ ~ 0 g ~

~

4

~

AEP: HRC: 00109 CHANGE NO.+/~:

COHTAIHMEHT SYSTEMS COHTAIHljEHT AIR RECIRCULATION SYSTEMS LIMITIHG CONDITION FOR OPERATION 3.6.5.6 Two independent containment air recirculation systems shall be OPERABLE.

APPLICABILITY:

MODES 1; 2, 3 and 4.

ACTION:

With one containment air r'ecirculaiion system inoperable, restore the inoperable system to OPERABLE status within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s or be in at least HOT STAHDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REOU IREMEHTS 4.6.5.6 Each containment air recirculation system shall'e demonstrated OPERABLE at least once per 3 months on a STAGGERED TEST BASIS by:

a ~

b.

Ci d.

Verifying that the return air fan starts on an auto-start signal after a 9 +

1 minute delay and operates for at-least 15 minutes, Verifying that with the return air fan dampers

closed, the fan motor current is 56 + 5 amps when the fan speed is 880 + 20

RPM, Verifying that with the fan off, the return air fan damoer opens when a force of < 11 lbs is applied to the counter-weight, and Verifying that the motor operated valve in the suction line to the containment's lower compartment opens after a

9

+

1 minute delay.

D. C.

COOK-UNIT 1

3/4. 6-35

CONTArtnENT SYSTEHS FLOOR-DRAINS LIMITING CONDITION FOR OPERATION 3.6.5.7 The ice condenser floor drains shall be OPERABLE.

APPLICABILITY:

ftOOES 1, 2,.3 and 4.

ACTION:

With the ice condenser floor drain inoperable, restore the floor drain to OPERABLE status prior to increasing the Reactor Coolant Svstem tem-perature above 200 F.

SURVEILLANCE REOtJIREt<ENTS 4.6.5.7 Each ice condenser floor drain shall be demonstrated OPERABLE at least once per 18 months during shutdown by:

'a ~

b.

Ce

,. d.

Verifying that valve gate opening is not impaired by ice, frost or debris, Verifying that the valve seat is not damaged, Verifying that the valve gate opens when a force of < 100 lbs is applied, and Verifying that the 12 inch drain line from the ice condenser floor to the containment lower compartment is unrestricted.

D. C.

COOP-UNIT 1

3/4 6-36

TABLE 3.3-5 Continued EHGIHEERED SAFETY FEATURES

RESPONSE

TIt'jES INITIATINGSIGNAL AND FUNCTION 6.

Steam Flow in Two Steam Lines-High Coincident with Ste="m.Line Pressure-Low

RESPONSE

TIME IN SECONDS a 1 b.

c d;

Safety Injection (ECCS)

Reactor Trip (from SI)

Feedwater Isolation Containment Isolation-Phase "A"

e.

Containment'urge and Exhaust Isolation f.

Auxiliary Feedwater Pumps g.

Essential Service Mater System h.

Steam Line Isolation'

13. 08/23. 0;ir"

< 3.0

< 8.0 18.0;,/28.0:a Not Applicable Not Applicable 14.08/%.088

< 8.0 Containment Pressure High-Hiqh a.

Conta'inment Spray b;

Containment Isolation-Phase "B"

c.

Steam Line Iso1ation d.

Containm nt Air Recirculation Fan.

< 45.0 Not Applicable

< 7.0

< 600.0 8.

Steam Generator Mater LevelHi h-Hi h a.

Turbire Trip-Reactor'rip b.

Feedwater Isolation

< 2.5

< 11.0 D.

C.

COOK-UNIT 1

3/4 3-29 'mendment. Ho.'8

TABLE 3.3-5 Continued TABLE NOTATION Diese1 generator starting and sequence loading delays inc1uded.

Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps, SI and RHR pmps.

Diesel generator starting and sequence loading delays not included.

Offsite porkier available.

Response

time limit includes opening of valves to establish SI path and attainment of disci arge pressure for centrifugal charging pumps.

N Diesel generator starting and sequence loading delays included.

Response

time limit includes opening of valves to 'establish SI path and at.ainment of discharge pressure for centrifugal charging pumps.

~

I D. C.

COOj:-UHIT 1 3/4 3-30

AEP:NRC:00109 CHANGE NO. /5

ADD>HISTRATIYE CONTROLS 6.9 REPORTING REQUI RENE}NTS {Continued) e.

Seismic event analysis, Specification

4. 3. 3. 3.2.

f.

Sealed Source leakage on excess of limits, Specification 4.7.9.1.3.

g.

Fire Detection Instrumentation, Specification 3.3.3.7.

h.

Fire Suppression

Systems, Specificaiions 3.7.9.1, 3.7.9.2, 3.7.9.3 and 3.7.9.4.

';10 RECORD RETENTIO}N 6.10.1 The following records shall be retained for at leas five years:

a.

Co Records and logs of unit, operation covering time interval ai each power }eveI.

Records'nd logs of principal maintenance activities, inspec-tions, repair and replacement of principal items of equipment related to nuclear safety.

All REPORTABLE OCCURRENCES submitted to the Commission.

d.

go h.

Records of surveillance activities, inspections and calibra-tions required by these Technical Specifications.

Records of reactor tests and experiments.

Records of changes made to the procedures required by Specifica-tion 6.8.1.

Records of radioactive shipments.

Records of sealed source leak tests and results.

Records of'nnual physical inventory of al.l sealed source material of record.

a.

Records and drawing changes reflecting unit design modifi-cations made to systems and equipm nt described in the Final Safety Analysis Report.

b.

Records of new and irradiat d fuel inventory, fuel transfers

,.and assembly burnup histories.

0.

C.

COO'r( - UNIT 1

6-19 Amendli'Bni No i jg }

23 6.10.2 The fo11owing records shall be retained for the duration of the Facility Operating License:

~

~ ~

(>>

~e ADMINISTRATIVE CONTROLS c.

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

d.

Records of gaseous and liquid radioactive material released to the environs.

e Records of transient or operational cycles for those facility

. components identified in Table 5.9-1.

Records of training and qualification for current m mbers of the plant staff.

g..

Records of in-service inspectin

>s perfor ed pursuant to these Technical Specifications.

h.

Records of guality Assurance activities required by the gA Hanual.

Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50,59.

j.

Records of meetings of the PNSRC and the NSDRC.

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.

I 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the "control device" or "alarm signal" required by paragraph 20.Z03(c)(2) of 10 CFR 20:

a 0 A High Radiation Area in which the intensity of -radiation is greater than 100 mrem/hr but less than 1000 mremlhr 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 o" individuals pemitted to enter such areas shall be provided with a radiation monitoring device, which continuously in-dicates the radiation dose rat in the area.

n I'hA'/

- I IllTT 6-90 An ndment No.

23

AEP:NRC:00109 CHANGE NO.JP

POllER DISTRIBUTIOH LIMITS

- DHB. PM%METERS LIMITIHG CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

a.

Reactor Coolant System T

b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate APPLICABILITY:

MODE 1

ACTION:

With any of the above parameters exceeding its limit, restore the param-eter to within its limit within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ". or reduce THEfVSL POllER to less than 5Ã of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE UIREMEHTS 4.2.5.1 Each of the. parameters of Table 3.2-1 shall be verified to be within their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18-months.

0.

C.

COOK - UNIT 1

3/4 2-13

n n

C)

PARAMETER reactor Coolant System Tavg Pressurizer Pressure Reactor Coolant System lotal Flow Rate TABLE 3.2-1 DNB PARANETERS 4 Loops In

~Otal

< 571.8'F LIMITS

> 2220 psia*

> 1.350x10 lbs/hr 3 Loops in 0 eration

< 571.8'F

> 2220 psia*

> 0.9917xl0 lbs/hr "Limit not applicable during either THERNL POWER ramp changes or THERNL POWER step changes in excess of 10Ã RATED THERNL POWER."

ML17317A840

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