Text

Amends 63 & 45 to Licenses DPR-58 & DPR-74,respectively, Changing Tech Specs Re Surveillance Requirements for Boron & Safety Injection Sys & Updating Core Flux
	ML17319B601
Person / Time
Site:	Cook
Issue date:	10/04/1982
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17319B602	List: ML17319B601; ML17319B604;
References
	NUDOCS 8210250393
	Download: ML17319B601 (79)
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UNITEDSTATES t NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN ELECTRIC COMPANY DOCKET.NO. 50-315 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No.

DPR-58 1.

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

A.

B.

C.,

D.

E.

The applications for amendment by Indiana and Michigan Electric Company (the licensee) dated December 22, 1978, February 22, 1980 and May 26, 1981 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted witho'ut endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

82i0250393 821004 PDR ADOCK 05000315 P

PDR

C) 2.

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

DPR-58 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license.

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

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

F R THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 4, 19B2 a

ef n

1 Operating Reacto s

B anch No.

Division of Licen ing

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 63 TO FACILITY OPERATING NO. DPR-58 DOCKET NO. 50-315 Revise Appendix A as follows:

Remove'a es 1-1 1-la 3/4 2-13*

3/4 2-14 3/4 3-3*

3/4 3-4 3/4 3-49 3/4 3-50*

3/4 5-6a 3/4 6-13 3/4 6-14*

3/4 6-35 3/4 6-36*

3/4 9-9*

3/4 9-10 B 3/4 2-3*

B 3/4 2-4 B 3/4 4-3 6-3 6-4 6-9 6-10*

6-19 6-20 Insert Pa es 1-1 3/4 2-13*

3/4 2-14 3/4 3-3*

3/4 3-4 3/4 3-49 3/4 3-50*

3/4 5-6a 3/4 6-13 3/4 6-14*

3/4 6-35 3/4 6-36*

3/4 9-9*

3/4 9-10 B 3/4 2-3* '

B 3/4 2-4 B 3/4 4-3 6-3 6-4 6-9 6-10*

6-19 6-20

, *Included for convenience

I

1. 0 DEFINITIONS DEFINED TERMS "

1.1 The DEFINED TERMS of this section appear in capitalized type and are applicable throughout these Technical Specifications.

THERMAL POWER 1.2 THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

RATED THERMAL POWER 1.3

. RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3250 MWt.

OPERATIONAL MODE 1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1.1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary state-ments to each principle specification and shall be part of the specifications.

OPERABLE - OPERABILITY 1.6 A system, subsystem,

train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s).

Implicit in this definition shall be the assumption that all necessary attendant instru-mentation, controls, normal and emergency electric power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the

system, subsystem, train, component or device to perform its function(s) are also capable of performing their related support function(s).

D.

C.

COOK " UNIT 1 AHEWDI1ENT NO 63

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING 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 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 51 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 UIREMENTS 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.

D.

C.

COOK - UNIT 1

3/4 2-13 AMENDMENT NO.

6'3

TABLE 3.2-1 DNB PARAMETERS PARAMETER 4 Loops In

~0eration LIMITS 3 Loops in

~0eration Reactor Coolant System Tavg Pressurizer Pressure Reactor Coolant System Total Flow Rate

< 571.8 F

> 2220 psia*

> 2220 psia"

> 1.350xl0 lbs/hr

> 0.9917x10 lbs/hr Limit not applicable during either a THERMAL POWER ramp in excess of 5X RATED THERMAL POWER per minute or a THERMAL POWER step in excess of lOX RATED THERMAL POWER.

TABLE 3.3-1 Continued REACTOR TRIP SVSTEM INSTRUMENTATION FUNCTIONAL UNIT 8.

Overpower hT Four Loop Operation Three Loop Operation TOTAL NO.

OF CHANNELS CHANNELS TO TRIP MINIMUM CHANNELS OPERABLE APPLICABLE MODES 1,

2 1,

2 ACTION 2

9 9.

Pressurizer Pressure-Low

, 10.

Pressurizer Pressure High 11.

Pressurizer Water Level--High 12.

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

Loss of Flow - Two Loops (Above P-7,and below P-8) 3/loop 3/loop 2/loop in any oper-ating loop 2/loop in two oper-ating loops 3

1, 2

2 1,

2 2/loop in 1

each oper-ating loop 2/loop each oper-ating loop

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 14.

Steam Generator Mater Level--Low-Low TOTAL NO.

OF CHANNELS 3/loop CHANNELS TO TRIP 2/loop in any oper-ating loops MINIMUM CHANNELS APPLICABLE OPERABLE MODES 2/loop in 1,

2 each oper-ating loop ACTION 15.

Steam/Feedwater Flow Mismatch and Low Steam Generator Mater Level 16.

Undervoltage-Reactor Coolant Pumps 17.

Underfrequency-Reactor Coolant Pumps 18.

Turbine Trip A.

Low Fluid Oil Pressure B.

Turbine Stop Valve Closure 19.

Safety Injection Input from ESF 2/1 oop-1 evel and 2/loop-flow mismatch in same loop 4/1/bus 4-1/bus 1/loop-level coincident with 1/1 oop-flow mismatch in same loop 1/loop-level and 2/1 oop-flow mismatch or 2/loop-level and 1/1 oop-flow mismatch 1,

2 1,

2

INSTRUMENTATION AXIAL POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION 3.3.3.6 The axial power distribution monitoring system (APDMS) 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, o, covering the full configuration of permissible rod patterns permitted at RATED THERMAL POWER.

.b.

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

j APPLICABILITY:

When the APDMS is used for monitoring the axial power dsstribut>on 8.

ACTION:

With the APDMS inoperable, do not use the system for determining the Axial Power Distribution.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.6.1 The full incore flux maps used to determine R and for monitoring 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 original uncertainty, a, shall not be updated, except as follows:

Except as provided in Specification 4.2.6. l.b.

8The APDMS 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.

D.

C.

COOK - UNIT 1

3/4 3-49 Amendment No. 63

1

INSTRUMENTATION SURVEILLANCE RE UIREMENTS (Continued) a

~

Ri -K,.

If the absolute value of~ is greater than 2o., another R.

map shall be completed to verify the new F... If the second map shows the first to be in error, the first hap shall be dis-regarded.

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 o

can be defined from the six new maps.

4.3.3.6.2 The APDMS 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 5X of RATED THERMAL POWER, by performance of a CHANNEL CALIBRATION.

D.

C.

COOK-UNIT 1

3/4 3-50 AMENDMENT NO.

63

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS 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 Sin le Pum Loop 1 Boron Injection Flow 117.5 gpm Loop 2 Boron Injection Flow 117.5 gpm Loop 3 Boron Injection Flow 117.5 gpm

, Loop 4 Boron Injection Flow 117.5 gpm Safety Injection System Sin le Pum ""

Loop 1 and 4 Cold Leg Fl ow > 300 gpm Loop 2 and 3 Cold Leg Flow > 300 gpm The flow rate in each Boron Injection (BI) line should be adjusted to provide 117.5 gpm (nominal) flow into each loop.

Under these conditions there is zero miniflow and 80 gpm simulated RCP seal injection line flow.

The actual flow in each BI 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.

Minimum, flow (total flow) required is 345.8 gpm to the three most conservative (lowest flow) branch lines.

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

D.

C.

COOK - UNIT 1

3/4 5-6a Amendment No.

63

CONTAINMENT SYSTEMS SURVEILLANCE RE UIREMENTS 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 signals d.

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

-20 gpm (> 20 gpm) but not to exceed 50 gpm (< 50 gpm) from the spray additive tank test line to each containment spray sys'em with the spray pump operating on recirculation with a pump discharge pressure

> 255 psig.

D.

C.

COOK - UNIT 1

3/4 6-13 Amen'dment No.

CONTAINMENT SYSTEMS 3/4.6. 3 CONTAINMENT ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6 '.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.

ACTION:

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

a.

Restore the inoperable valve(s) to OPERABLE 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 I

c.

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 valve or blind flange; or de 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 REQUIREMENTS 4.6.3.1.1 The isolation valves specified in Table 3.6-1 shall be demonstrated OPERABLE:

a.

At least once per 92 days by cycl,ing each OPERABLE oower ooerated or automatic valve testable during plant operation through at least one complete cycle of.,full travel.

b.

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

C.

COOK-UNIT 1

3/4 6-14 AMENDMENT NO.'63 '

CONTAINMENT SYSTEMS CONTAINMENT AIR RECIRCULATION SYSTEMS LIMITING 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 recirculation system inoperable, restore the inoperable system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 .

K SURVEILLANCE REOUIREMENTS 4.6.5.6 Each containment air recirculation system shall be demonstrated OPERABLE at least'once per 3 months on a

STAGGERED TEST BASIS by:

a.

Verifying that the return air fan starts on an auto-sCarC signal after a

10 +

1 minute delay and operates for at least 15 minutes, P

b.

Verifying that with the return air fan discharge backdraft damper locked closed and the fan motor energized, the static pressure between the fan discharge and the backdraft damper is

> 4.0 inches, water gauge.

c.

Verifying that with the fan off, the return air fan damper

'pens when a force of <

11 lbs is applied to the counter-

weight, and d.

Verifying that'he 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 Amendment No.'3

CONTAINMENT SYSTEMS FLOOR DRAINS LIMITING CONDITION FOR OPERATION 3.6.5.7 The ice condenser floor drains shall be OPERABLE.

APPLICABILITY:

MODES 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 System tem-perature above 200'F.

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

a.

Verifying that valve gate opening is not impair'ed by ice, frost or debris, b.

Verifying that the valve seat is not damaged, C.

d.

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.

COOK-UNIT 1

3/4 6-36 AMENDMENT NO.

63

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

APPLICABILITY:

MODE 6.

ACTION:

a

~

b.

C.

With 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.

C'lose all 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 .

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.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 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 .

D.

C.

COOK - UNIT 1

3/4 9-9 AMENDMENT NO. 63

REFUELING OPERATIONS CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Purge and Exhaust isolation system shall be OPERABLE.

APPLICABILITY: During Core Alterations or movement of irradiated fuel within ACTION:

With the Containment Purge and Exhaust isolation system inoperable, close each of the Purge and Exhaust penetrations providing direct access from the containment atmosphere to the outside atmosphere.

The provision of Specification 3.0.3 are not applicable.

SURVEILLANCE RE UIREMENTS 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 initiationand on a high radiation signal from each of the containment radiation monitoring instrumentation channels.

D.

C.

COOK - UNIT 1

3/4 9-10 Amendment No.

63

r cI 4c I ~ o v/

~ tv\\

ai Power Therm 100%

~ I i ~

90%

80o%%d

~

70%

60o/o Target Flux Difference

~ /

50o%%d 40%

30%

20%

10%

~30%

.20'/o

~ 10%o 0

+10%

+20o%%d INDICATED AXIAL FLUX DIFFERENCE Figure B 3/4 2.1 TYPICAL INDICATED AXIALFLUX DIFFERENCE VERSUS THERMAL PO(VER AT BOL 30%

0.

C.

COOK-UNIT 1

B 3/4 2-3 AMENDMENT NO.

63'

POWER DISTRIBUTION LIMITS BASES 3/4. 2. 2 and 3/4. 2. 3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS-

~Fo>

d 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 wi 11 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

'.2.9.

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

a.

Control rods in 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. l. 3. 5.

c.

The control rod insertion limits of Specifications

3. 1.3.4 and 3.1.3.5 are maintained.

d.

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 THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

F will be maintained within its limits provided conditions a thru d aboN, are maintained.

When an F

measurement is taken, both experimental error and man-ufacturing tolerance must be allowed for.

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

N When F

is measured, experimental error must be allowed for and 4X is the appr'seriate allowance for a fu11 coreNmap taken with the incore detection system.

The specified limit for F also contains an SX allowance for uncertainties which mean that kormal operation will result in F

H < 1.51/1.08.

The SX allowance is based on the following considera-tion)":

D.

C.

COOK - UNIT 1

B 3/4 2-4 Amendment No. 63

REACTOR COOLANT SYSTEM BASES 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.6.

1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the Reactor Coolant Pressure Boundary.

These detection systems are consistent with the recommendations of Regulatory Guide 1.45, "Reactor Coolant Pressure Boundary Leakage Detection Systems,"

Hay 1973.

3/4. 4. 6. 2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is

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

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

The 10 GPH IDENTIFIED LEAKAGE limitations provides allowance for a limited amount of leakage from known sources whose presence will not interfere with the detection of UNIDENTIFIED LEAKAGE by the leakage detection systems.

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.

D.

C.

COOK - UNIT 1 B 3/4 4"3 Amendment No, giP 63

I C.

PLANT MANAGER ADMINISTRATIVE SUPERVISOR MAINTENANCE OPERATION SUPERINTENDENT SUPERINTENDENT SOL TECHNICAL SUPERINTENDENT STAFF TRAINING COORDINATOR STAFF PROD. SUPV.

OPERATIONS SOL SHIFT OPERATING ENG.

SOL QA SUPERVISOR OPERATING ENGINEER OL NUCLEAR PERFORMANCE ENGINEFR SUPERVISOR ENGINEER CONTROL AND +

INSTRUMENTATION ENGINEER PLANT CHEMICAL SUPERVISOR PLANT

~

RADIATION PROT. SUPV.

LEGEND:

OL - SENIOR OPERATOR LICENSE OL - OPERATOR LICENSE

~ - KEY SUPERVISORY PERSONNEL UNIT SUPERVISOR OL PERFORMANCE PERFORMANCE ENGINEER ENGINEER INSTRUMENT MAINTENANCE SUPERVISOR CHEMIST RADIA PROTE SUPERVISOR 3

(D CL Gl C+

O EQUIPMENT OPERATOR AUXILIARY EQUIPMENT OPERATOR TECHNICIANS TECHNICIANS Figure 624 Facility Organization - Donald C. Cook - Unit No.

1

TABLE 6.2-1 HINItlUH SHIFT CREM CCtlPOSITIOM8 LICENSE CATEGORY SOL OL NON-Licensed Shift Technical Advisor APPl ICABLE NODES 1, 2, 3

8 4

2 Mone Required Woes not include the licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIOMS after the initial fuel loading.

dShift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on duty shift crew members provided itnnediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.

shared with 0.

C. Cook Unit 2.

D. C.

COOK - UNIT 1

6-4 Amendment No; PP 63

ADMINISTRATIVE CONTROLS COMPOSITION 6.5.2.2 The NSDRC shall be composed of the:

Chairman:

Member:

~

Member:

~ Member:

Alternate:

ALTERNATES Assistant Vice President, Nuclear Engineering Vice Chairman, Engineering and Construction President and Chief Operating Officer of IQlECo Executive Vice President, Construction and New York Engineering Vice President, Mechanical Engineering Vice President, Electrical Engineering Vice President, Engineering Administration Assistant Vice President, Design Division Assistant Vice President, Environmental Engineering Division Plant Manager, D.

C.

Cook Plant

Manager, Nuclear Safety and Licensing Section Assistant Chief Mechanical Engineer Assistant Plant Manager, D.

C.

Cook Plant Executive Assistant to the President of IMECo Assistant Division Manager, Nuclear Engineering 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 members 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 FRE UENCY 6.5.2.5 The NSDRC shall meet at least once per calendar quarter during the initial year of facility operation following fuel loading and at least once per six months thereafter.

UORUM 6.5.2.6 A quorum of NSDRC shall consist of the Chairman or his designated alternate and more than half the NSDRC membership including alternates or at least 5 members including alternates whichever is greater.

No more than a

minority of the quorum shall have line responsibility for operation of the fac i 1 ity.

D.

C.

COOK " UNIT 1

6-9 Amendment No.

63

ADMINISTRATIVE CONTROLS REVIEW 6.5.2.7 The NSDRC shall review:

a

~

b.

c

~

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.

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.

'd.

Proposed changes in Technical Specifications or licenses.

e.

9 ~

Violations of applicable statutes,

codes, regulations,

orders, Technical Specifications, license requirements, or of intet nal 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 Comoission.

h.

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

systems, or components.

Reports and meetings minutes of the PNSRC.

D.

C.

COOK - UNIT 1

6-10 AMENDMENT NO.

63

ADMINISTRATIVE CONTROLS 6.9 REPORTING RE UIREMENTS (Continued) e.

Seismic event analysis, Specification 4 ~ 3.3.3.2.

f.

Sealed Source leakage in excess of limits, Specification

4. 7. 7. 1. 3.

g.

Fire Detection Instrumentation, Specification 3.3.3.7 h.

Fire Suppression

Systems, Specifications

3. 7. 9. 1, 3. 7. 9. 2, 3.7.9.3 and 3.7.9.4.

6. 10 RECORD RETENTION

~ 6.10.1 The following records shall be retained for at least five years:

a.

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

b.

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

c.

All REPORTABLE OCCURRENCES submitted to the Commission.

d.

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.

g.

Records of radioactive shipments.

h.

Records of sealed source leak tests and results.

i.

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

6. 10.2 The following records shall be retained for the duration of the Facility Operating License:

a ~

b.

Records and drawing changes reflecting unit design modifi-cations 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.

D.

C.

COOK - UNIT 1

6-19 Amendment No. 63

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.

f.

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

g.

h.

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

~

'ecords of equality Assurance activities required by the gA Manual.

i.

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.

k.

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

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 High Radiation Area in which the intensity of radiation r

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

D.

C.

COOK - UNIT 1 6-20 Amendment No.

63

~pg REgy (4

0 e

O 0

n +**++

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

COOK NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

45 License No.

DPR-74

).

The B.

C.

D.

E.

Nuclear Regulatory Commission (the Commission) has found that:

0 The applications for amendment by Indiana and Michigan Electric Company (the licensee) dated December 22, 1978, February 22, 1980 and May 26, 1981 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; The facility will operate in conformity with the applications, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

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

DPR-74 is hereby amended to read as follows:

(2)

Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 45, are hereby incorporated in the license.

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

3.

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

F R THE NUCL R REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

October 4, ]9Q2 ven V rg, Chief Operating Reac o s Branch No.

1 Division of Li e sing

ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO. 45 TO FACILITY OPERATING LICENSE NO. DPR-74 DOCKET NO. 50-316 Revise Appendix A as follows:

Remove Pa es 1 1 1*

1V 1-1 1-1 a 3/4 l-l*

3/4 1-2 3/4 2-15*

3/4 2-16 3/4 3-3 3/4 3-4*

3/4 3-8 3/4 3-25 3/4 3-27 3/4 3-39 3/4 3-40 3/4 3-41 3/4 3-42*

3/4 3-48 3/4 3-53 3/4 3-54*

3/4 4-1 3/4 4-2*

3/4 4-25 3/4 4-26*

3/4 5-5*

3/4 5-6 3/4 6-11*

3/4, 6-12 3/4 6-43*

3/4 6-44 3/4 9-9 3/4 9-10*

8 3/4 2-3*

8 3/4 2-4 8 3/4 3-2 8 3/4 4-1 6-3 6-4 6-9 6-10 6-19 6-20

Included for convenience Insert Pa es 111*

1V 1-1 3/4 1-1*

3/4 1-2 3/4 2-15*

3/4 2-16 3/4 3-3 3/4 3-4*

3/4 3-8 3/4 3-25 3/4 3-27 3/4 3-38a 3/4 3-38b 3/4 3-38c 3/4 3-39 3/4 3-40 3/4 3-41 3/4 3-42*

3/4 3-48 3/4 3-53 3/4 3-54*

3/4 4-1 3/4 4-2*

3/4 4-25 3/4 4-26*

3/4 5-5*

3/4 5-6 3/4 6-11*

3/4 6-12 3/4 6-43*

'3/4 6-44

3/4 9-9 3/4 9-10*

8 3/4 2-3*

8 3/4 2-4 8 3/4 3-2 8 3/4 4-1 6-3 6-4 6-9 6-10 6-19 6-20

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION 3/4.0 APPLICABILITY............................

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTROL Shutdown Margin - T

) 200'F avg Shutdown Margin - T

( 200'F.......

avg-Boron Dilution..........

Moderator Temperature Coefficient...

Minimum Temperature for Criticality..

~Pa e

~

3/4 0

1 3/4 1-1 3/4 1-3 3/4 1-4 3/4 145 3/4 1-7 3/4.1.2 BORATION SYSTEMS Flow Paths - Shutdown..............

Flow Paths

- Operating..

Charging Pump - Shutdown................

Charging Pumps -'perating.

Boric Acid Transfer Pumps - Shutdown.

Boric Acid Transfer Pumps - Operating.

Borated Water Sources

- Shutdown........

Borated Water Sources

- Operating....

3/4.1.3 MOVABLE CONTROL ASSEMBLIES Group Heighto ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

Position Indicator Channels

- Operating.

Position Indicator Channels

- Shutdown..

~

Control Rod Insertion Limits.........

~

~ ~

~

R p

~

od Drop Tlmet ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

Shutdown Rod Insertion Limit............................

3/4 1-8 3/4 1-9 3/4 1-11 3/4 1-12 3/4 1-13 3/4 1-14 3/4 1-15 3/4 1-16 3/4 1-18 3/4 1-21 3/4 1-22 3/4 1-23

'3/4 1-24 3/4 1-25 D.

C.

COOK - UNIT 2 AMENDMENT NO. 45

INDEX LIMITING CONDITIONS'FOR OPERATION 5 SURVEILLANCE RE UIREMENTS Axial Power Distribution Monitoring Fire Detection Instrumentation.

3/4. 3. 4 TURBINE OVERSPEED PROTECTION.

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.

1 REACTOR COOLANT LOOPS System.............

3/4. 2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axial Flux Difference...

3/4.2.2 Heat Flux Hot Channel Factor...

3/4.2.3 RCS Flow Rate and R..

3/4.2.4 quadrant Power Tilt Ratio....................

3/4.2.5 DNB Parameters

'3/4.2.6 Axial Power Distribution.

3/4.3 INSTRUMENTATION 3/4.3.

1 REACTOR TRIP SYSTEM INSTRUMENTATION 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION.

3/4. 3. 3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.....

Movable Incore Detectors......

Seismic Instrumentation..

Meteorological Instrumentation...

Remote Shutdown Instrumentation.

Post-Accident Instrumentation..........................

3/4 2-1 3/4 2-5 3/4 2-9 3/4 2-13 3/4 2-15 3/4 2"17 3/4 3-1 3/4 3-14 3/4 3-34 3/4 3-38 3/4 3-38a 3/4 3"39 3/4 3-42 3/4 3"45 3/4 3"48 3/4 3-50 3/4 3"53 3/4.4.2 3/4.4.3 3/4.4.4 3/4. 4. 5 Normal Operation.............,.

SAFETY VALVES - SHUTDOWN...............................

SAFETY VALVES - OPERATING..........

PRESSURIZER............................

STEAM GENERATORS.......................................

3/4 4-1 3/4 4"4 3/4 4-5 3/4 4-6 3/4 4-7 D.

C.

COOK - UNIT 2 IV Amendment No.

45

1. 0 DEFINITIONS DEFINED TERMS 1.1 The DEFINED TERMS of this section appear in capitalized type and are applic-able throughout these Technical Specifications.

THERMAL POWER

1. 2 THERMAL POWER shall be the total reactor core heat transfer rate to the reac-tor coolant.

RATED THERMAL POWER

l. 3

. RATED-THERMAL POWER shall be a total reactor core heat transfer r'ate to the reactor coolant of 3391 MWt.

OPERATIONAL MODE I

1.4 An OPERATIONAL MODE shall correspond to any one inclusive combination of core reactivity condition, power level and average reactor coolant temperature specified in Table 1. 1.

ACTION 1.5 ACTION shall be those additional requirements specified as corollary state-ments to each principle specification and shall be part of the specifications.

OPERABLE - OPERABILITY 1.6 A system, subsystem,

train, component or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function(s).

Implicit in this definition shall be the assumption that all necessary attendant instru-

'entation,

controls, normal and emergency electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the

system, subsystem,

train, component or device to perform its function(s) aye also capable of performing their related support function(s).

D.

C.

COOK - UNIT 2

'MENDMENT NO. 45

l)

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REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UIREMENTS Continued d.

e.

Prior to initial operation above 5X RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification

3. 1.3.6.

When in MODES 3 or 4, at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

l.

Reactor coolant system boron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4. l. l. 1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within + 1X h,k/k 'at least once per 31 Effective Full Power Days (EFPD).

This comparison shall consider at least those factors stated in Specification 4. l. 1. l. l.e, above.

The predicted reactivity values shall be adjusted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading.

D. C.. COOK - UNIT 2 3/4 1-2 Amendment No. 45

OWER DISTRIBUTION LIMITS NB PARAMETERS IMITING 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 Tay b.

Pressurizer Pressure.

APPLICABILITY; MODE 1

ACTION:

With any of the above parameters exceeding its limit, restore the param eter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5". of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURV E I LLANCE R E 0 IR EMENTS 4.2.5 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 .

D.

C.

COOK - UNIT 2 3/4 2-15 AMENDMENT NQ.

45

TABLE 3.2-1 DNB PARAMETERS PARAMETER Reactor Coolant System Tavg Pressurizer Pressure LIMITS 4 Loo s In 0 eration

< 578 F

> 2220 psia" 3 Loo s In 0 eration

< 570 F

> 2220 psia" Lim)t not app

~ca e during either a THERMAL POWER ramp in excess of 5X RATED THERMAL POWER per minute or a THERMAL POWER step in excess of lOX RATED THERMAL POWER.

TABLE 3.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION FUNCTIONAL UNIT TOTAL NO.

OF CHANNELS MINIMUM CHANNELS CHANNELS TO TRIP OPERABLE APPLICABLE MODES ACTION 9.

Pressurizer Pressure-Low 10.

Pressurizer Pressure High 11.

Pressurizer Water LevelHigh 12.

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

Loss of Flow - Two Loops (Above P-7 and below P-8) 14.

Steam Generator Water Level--Low-Low 3/loop 3/1 oop 3/1 oop 2/loop in any oper-ating loop 2/loop in two oper-ating loops 2/loop in any oper-ating loop 2/loop in each oper-ating loop 2/1 oop each oper-ating loop 2/loop each oper-ating loop 1,

2 1,

2 1 j 2 e¹ 6¹ 7¹ 7¹ 7¹ 15.

Steam/Feedwater Flow Mismatch and Low Steam Generator. Water 2/loop-level and 2/loop-flow mismatch in same loop 1/loop-level coincident with 1/1 oop-f1 ow mismatch in same loop 1/loop-level 1,

2 and 2/loop-flow mismatch or 2/loop-level and 1/1 oop-f1 ow mismatch

TABLE 3.3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 16.

Undervoltage-Reactor Coolant Pumps 17.

Underfrequency-Reactor Coolant Pumps TOTAL NO.

OF CHANNELS 4-1/bus 4-1/bus CHANNELS TO TRIP MINIMUM CHANNELS APPLICABLE OPERABLE MODES ACTION 18.

Turbine Trip A.

Low Fluid Oil Pressure 3

B.

Turbine Stop Valve Closure 4

19.

Safety Injection Input from ESF 1,

2 6

20.

Reactor Coolant Pump Breaker Position Trip A.

Above P-8 B.

Above P-7 21.

Reactor Trip Breakers 22.

Automati'c Trip Logic

'1/breaker 1/breaker 1/breaker 1

1/breaker 1

per oper-ating loop 1,

2 and

1, 2 and

10ll

TABLE 3. 3-1 Continued DESIGNATION CONDITION AND SETPOINT FUNCTION P-7 With 2 of 4 Power Range Neutron Flux Channels

> illof RATED THERMAL POWER or 1 of 2 Turbine impulse chamber pressure channels

> 66 psia.

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

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

P"8 With 2 of 4 Power Range Neutron Flux channels

> 31K of RATED THERMAL POWER.

P-8 prevents or defeats the automatic block of reactor trip on low coolant flow in a single loop.

i'-10 With 3 of 4 Power range neutron flux channels ( 9X of RATED THERMAL POWER.

P-10 prevents or defeats the manual block of:

Power range low setpoint reactor trip, Intermediate range reactor trip, and intermediate range'od stops.

Provides input to P-7.

D.

C.

COOK - UNIT 2 3/4 3-8 Amendment No.

45

TABLE 3. 3-4 Continued)

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 4.

STEAM LINE ISOLATION a.

Manual b.

Automatic Actuation Logic c.

Containment Pressure High-High d.

Steam Flow in Two Steam lines--

High Coincident with T

--Low-Low avg Not Applicable Not Applicable

< 2.9 psig

< A function defined as follows: A hp co~respond-ing to 1.47 x 10 lbs/hr steam flow between OX and 20K load and then a hp increasing linearly to a hp corresponding to 110K of full steam flow at full 1 oad.

T

> 541 F

Not Applicable Not Applicable

< 3.0 psig

< A function defined as follows:

A gp corresponding to 1.62 x 10 lbs/hr steam flow between OX and 20K load and then a hp increasing linearly to a hp corresponding to 111.5X of full steam flow at full load.

T

> 540'F e.

Steam Line Pressure--Low 5.

TURBINE TRIP AND FEED WATER ISOLATION a.

Steam Generator Water level High-High

> 600 psig steam line pressure

< 67X-of narrow range 7nstrument span each steam generator

> 585'sig steam line pressure

< 68K of narrow range instrument span each steam generator

TABLE 3.3-5 Continued)

ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS e.

g.

Containment Purge and Exhaust Isolation Motor Driven Auxiliary Feedwater Pumps Essential Service Water System 3.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) b.

Reactor Trip (from SI) c.

Feedwater Isolation d.

Containment Isolation-Phase "A"

c 24.0"/12.0¹

< 2.0

< %.0

< 18.0¹ Not Applicable

< 60.0

< 48. 0*/13. 0¹

'4.

Differential Pressure Between Steam Lines-Hi h

a ~

b.

C.

d.

e.

g.

Safety Injection (ECCS)

Reactor Trip (from SI)

Feedwater Isolation Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation Motor Driven Auxiliary Feedwater Pumps Essential Service Water System 12.0¹/24.0¹¹

< 2.0

< 8.0

<. 18. 0¹/28. 0¹¹ Not Applicable

< 60.0

< 13.0¹/48.0¹¹ 5.

Steam Flow in Two Steam Lines - Hi h Coincident w1th Tav Low Low a 0 b.

C.

d.

e.

g.

h.

Safety Injection (ECCS)

Reactor Trip (from SI)

Feedwater Isolation Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System Steam Line Isolation Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable D.

C.

COOK " UNIT 2 3/4 3-27 Amendment No.

45

INSTRUMENTATION SEISMIC INSTRUMENTATION" LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic monitoring instrumentation shown in Table 3.3-7 shall be OPERABLE.

APPLICABILITY: At all times.

ACTION:

a.

b.

C.

Mith the number of OPERABLE seismic monitoring instruments less than required by Table 3.3-7, restore the inoperable instru-ment(s) to OPERABLE status within 30 days.

With one or more seismic monitoring instruments inoperable for more than 30 days, prepare and submit a Special Report to the

.Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans 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 REQUIREMENTS

4. 3. 3. 3. 1 Each of the above seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST operations at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of the above seismic monitoring instruments actuated during a seismic event shall be restored to OPERABLE status and a

CHANNEL CALIBRATION performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the seismic event.

Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.

A Special Report shall be pre-pared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, frequency spectrum and resultant effect upon facility features important to safety.

Shared System with D.

C.

Cook Unit 1.

D.

C.

COOK - UNIT 2 3/4 3-38a Amendment No. 4g

TABLE 3.3-7 SEISMIC MONITORING INSTRUMENTATION INSTRUMENTS AND SENSOR LOCATIONS 1.

STRONG MOTION TRIAXIAL ACCELEROGRAPHS a.

Reactor Pit FLoor b.

Top of Crane Wall c.

Free Field

,2.

'PEAK RECORDING ACCELEROGRAPHS a.

Containment. Spring Line b.

Diesel Generator Room Floor c.

Spent Fuel Pool MEASUREMENT RANGE O-l g

0-1 g

O-l g

0-2 g 0-2 g 0-2 g MINIMUM INSTRUMENTS OPERABLE D.

C.

COOK - UNIT 2 3/4 3-38b Amendment No.

45

TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE RE UIREMENTS INSTRUMENT CHANNEL 1.

STRONG MOTION TRIAXIALACCELEROGRAPHS a.

Reactor Pit Floor 1.

Time History Recorder 2.

Seismic Trigger b.

Top of Crane Mall 1.

Time History Recorder c.

Free Field 1.

Time History Recorder 2.

Seismic Trigger 2.

PEAK RECORDING ACCELEROGRAPHS a.

Containment Spring Line b.

Diesel Generator Room Floor c.

Spent Fuel Pool CHANNEL CHECK M

NA NA CHANNEL CALIBRATION CHANNEL FUNCTIONAL TEST M

NA M

NA NA NA

0 INSTRUMENTATION METEOROLOGICAL INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3 '

The meteorological monitoring instrumentation channels shown in Table 3.3-8 shall be OPERABLE.

APPLICABILITY: At al 1 times.

ACTION:

a.

b.

C.

With the number of OPERABLE meteorological monitoring channels less than required by Table 3.3-8, suspend all release of gaseous radioactive material from the radwaste gas decay tanks until the inoperable channel(s) is restored to OPERABLE status.

With one or more required meteorological monitoring channels

.inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9. 2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel(s) to OPERABLE status.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.3.3.4 Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-5.

Shared system with D.

C.

COOK - UNIT l.

D.

C.

COOK - UNIT 2 3/4 3-39 Amendment No. '45

TABLE 3.3-8 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT 1.

WIND SPEED LOCATION INSTRUMENT MINIMUM ACCURACY MINIMUM OPERABLE a.