Text

Amend 159 to License DPR-74,revising TS to Allow Tests Designated as 18-month Surveillances to Be Delayed Until Next Refueling Outage Scheduled to Begin 940806
	ML17331B201
Person / Time
Site:	Cook
Issue date:	01/26/1994
From:	Blough A; Office of Nuclear Reactor Regulation
To:	;
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ML17331B203	List: ML17331B201;
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	NUDOCS 9402010377
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n k))*y4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20555-0001 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-316 DONALD C.

COOK NUC EAR LANT UN T NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 159 License No.

DPR-74 1.

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

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated April 16,

1993, as supplemented September 28 and December 3,

1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There i,s reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of

'the Commission's regulations and all applicable requirements have been satisfied.

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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:

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

as revised through Amendment No.'159

, 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 issuance.

Attachment:

Changes to the Technical Specifications Date of Issuance:

vanuarv 25, 1994 FOR T LEAR REGULATORY COMMISSION gr6..@,.

<o<

I A. Randolph Blough, Acting Director Project Directorate III-I Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

ATTACHMENT TO LICENSE AMENDMENT NO. 159 TO FACILITY OPERATING 4.ICENSE NO.

DPR-74 DOCKET NO. 50-316 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by amendment number and contain vertical lines indicating, the area of change.

REMOVE 3/4 1-22 3/4 3-11 3/4 3-14 3/4 3-31 3/4 3-32 3/4 3-44d 3/4 3-47 3/4 4-14 3/4 4-33 3/4 5-5 3/4 5-8 3/4 6-47 3/4 7-6 3/4 7-13 3/4 7-20 3/4 7-31 3/4 8-4 3/4 8-9 INSERT 3/4 1-22 3/4 3-11 3/4 3-14 3/4 3-31 3/4 3-32 3/4 3-44d 3/4 3-47 3/4 4-14 3/4 4-33 3/4 5-5 3/4 5-8 3/4 6-47 3/4 7-6 3/4 7-13 3/4 7-20 3/4 7-31 3/4 8-4 3/4 8-9

EACTIV CO OL OSI 0

ND C f

MIT NG COND ON 0

T N

3.1.3.3 At least one rod position indicator channel (excluding demand position indication) shall be OPERABLE for each shutdown or control rod not fully inserted.

~C~IO With less than the above required position indicator channel(s)

OPERABLE, immediately open the reactor trip system breakers.

URVEILLANCE RE UI 4.1.3.3 Each of the above required rod position indicator channel(s) shall be determined to be OPERABLE by performance of a CHANNEL FUNCTIONAL TEST at least once per 18 months.f

+With the reactor trip system breakers in the closed position.

/See Special Test Exception 3.10.5.

f'The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PIANT - UNIT 2 3/4 1-22 AMENDMENT NO. 40,

%L, ~,

159

nOO I

ABLE 4 3-FUNCTIONAL UNIT 2.

Power Range, Neutron Flux CHANNEL CHECK N.A.

N.A.

CHANNEL CALIBRATION N.A.

N.A.

D(2,8),M(3,8) and Q(6,8)

CHANNEL FUNCTIONAL TEST S/U(l)(10)

S/U(l)(10)

M and S/U(1)

EACTOR TR SYSTEM INSTRUMENTATION SURVEILLANCE RE REMENTS MODES IN WHICH SURVEILLANCE RE UIRED 3*, 4*, 5*

]

2 3*, 4*, 5*

1, 2 and*

IA C

I 3.

Power Range, Neutron Flux, High Positive Rate 4 ~

Power Range, Neutron Flux, High Negative Rate 5.

Intermediate

Range, Neutron Flux 6.

Source

Range, Neutron Flux N.A.

N.A.

R(6)

R(6,8)

R(6,14)

M S/U(1)

M(14) and S/U(l) 1 2

1, 2

1, 2and*

2(7), 3(7),

4 and 5

7.

Overtemperature AT 8.

Overpower AT 9.

Pressurizer Pressure--Low

10. Pressurizer Pressure--High

11. Pressurizer Water Level--High

12. Loss of Flow - Single Loop R(9)f R(9)f Rf Rf Rt

'R(8) 1 2

1 2

1, 2

f The provisions of Technical Specification 4.0.8 are applicable.

S ENTAT 0 3 4 3

ENGI ERED FETY T 0 SYS MIT G CONDI ON FOR 0 3.3.2.1 The Engineered Safety Feature Actuation System (ESFAS) instrumenta-tion channels and interlocks shown in Table 3.3-3 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3-4 and with RESPONSE TIMES as shown in Table 3.3-5.

~C'~0 a.

With an ESFAS instrumentation channel trip setpoint less conserva-tive than the value shown in the Allowable Values column of Table 3.3-4, declare the channel inoperable and apply the applicable ACTION requirement of Table 3.3-3 until the channel is restored to OPERABLE status with the trip setpoint adjusted consistent with the Trip Setpoint value.

b.

With an ESFAS instrumentation channel inoperable, take the ACTION shown in Table 3.3-3.

URVEILLANCE RE UI S

4.3.2.1.1 Each ESFAS instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, QOQINEL CALIBRATION, CHANNEL FUNCTIONAL TEST and TRIP ACTUATING DEVICE OPERATIONAL TEST operations for the MODES and at the frequencies shown in Table 4.3-2. f 4.3.2.1.2 The logic for the interlocks shall be demonstrated OPERABLE during the automatic actuation logic test.

The total interlock function shall be demonstrated OPERABLE at least once per 18 months during CHANNEL CALIBRATION testing of each channel affected by interlock operation.f 4.3.2.1.'3 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be demonstrated to be within the limit at least once per 18 months.

Each test shall include at least one logic train such that both logic trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the "Total No. of Channels" Column of Table 3.3-3.f f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PIANT - UNIT 2 3/4 3-14 AMENDMENT NO. W, W, 434, ~,

~,

159

BLE 4 3-2 Co ued ENG NE D

S F F

TURED ACTUATION SYSTEM(

S ENT ON URVEILLANCE RE UIREM CT ON

c. Purge and Exhaust Isolation TRIP ACTUATING MODES IN AQNNEL DEVICE WHICH CHANNEL QiANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE ZUH-XEI1 MLIRE

1) Manual

2) Containment Radio-S activity-High 4.

STEAM LINE ISOLATION See Functional Unit 9 M

N.A.

1, 2, 3, 4

a. Manual

b. Automatic Actuation Logic

c. Containment Press-ure--High-High d.

Steam Flow in Two Steam Lines--

High Coincident with T~z- -Low-Low e.

Steam Line Pressure--

Low N.A.

M(3)

N.A.

See Functional Unit 9-N.A.

M(2)

N.A.

1, 2, 3

1 2

3 1, 2, 3

5. TURBINE TRIP AND FEEDWATER ZSOIATION a.

Steam Generator Water Level-'-High-High 6.

MOTOR DRIVEN AUXILIARY FEEDWATZR PUMPS a.

Steam. Generator Water Level--Low-Low

b. 4'V Bus Loss of Voltage

c. Safety Injection

d. Loss of Main Feed Pumps N.A.

N.A.

M" M(2)

Rf N.A.

N.A.

1,2,3 1, 2, 3

1, 2, 3

1, 2

f The provisions of Technical Specification 4.0.8 are'applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-31

1 tl

TABLE 4 3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 7.

TURBINE DRIVEN AUXILIARY FEEDWATER PUMP TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE CHECK CALIBRATION TEST TEST ml a.

Steam Generator Water Level--Low-low

b. Reactor Coolant Pump Bus Undervoltage 8.

LOSS OF POWER N.A.

N.A, N.A.

1, 2,

3 1, 2, 3

a.

4 kv Bus Loss of Voltage

b. 4 kv Bus Degraded Voltage 9.

MANUAL Rf Rf M

N.A.

a. Safety Injection (ECCS)

Feedwater Isolation Reactor Trip (SI)

Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation Auxiliary Feedwater Pumps Essential Service Water System b.

Containment Spray Containment Isolation-Phase "B"

Containment Purge and Exhaust Isolation Containment Air Recirculation Fan c.

Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation N.A.

N.A.

N,A.

N.A.

Rf Rf Rf 1

2 3

4 I

1,2,3,4 d.

Steam Line Isolation N.A.

N.A.

M(lf).

Rf

>.,a,3 f The provisions of Technical Specifi'cation 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-32 AMENDMENT NO. 88> &7, 434, 159

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XERXBQSHK P

END X REHO E

S TDO ONITORING S

SUVE C

EURE ENS

~CT~O CHANNEL cog'cK CHANNEL C

0 1.

Steam Generators 1

and 4 Level LSI Cabinet 1 and LSI Cabinet 4 2.

Steam Generators 2

and 3 Level 3.

Steam Generators 1

and 4 Pressure LSI Cabinet 2 and LSI Cabinet 4 LSI Cabinet 4 and LSI Cabinet 5

4.

Steam Generators 2

and 3 Pressure LSI Cabinet 4 and LSI Cabinet 6

5.

Reactor Coolant Loop 4 Temperature (Cold) 6.

Reactor Coolant Loop 4 Temperature (Hot)

LSI Cabinet 4 and LSI Cabinet 5

I LSI Cabinet 4 and LSI Cabinet 5

Rg Rf 7.

Reactor Coolant Loop 2 Temperature (Cold)

LSI Cabinet 4 and LSI Cabinet 6

8.

Reactor Coolant Loop 2 Temperature (Hot)

LSI Cabinet 4 and LSI Cabinet 6

9.

Pressurizer Level 10.

Reactor Coolant System Pressure 11.

Charging Cross-Flow Between Units 12.

Source Range Neutron Detector (N-23)

LSI Cabinet 3

gorridor Elev.

587'SI Cabinet 4 n/a n/a Charging Cross-Flow between Units is an instrument common to both Unit 1 and 2.

This surveillance will only be conducted on an interval consistent with Unit: 1 refueling.

The provisions of Technical Specification 4.0.8 are applicable.

l

A OO INSTRUMENT TABLE 4 3-10 POST-ACCIDENT MONITORING NSTRUMEN ATION SURVEILLANCE RE UIREMENTS CHANNEL CHECK CHANNEL CALIBRATION 1.

Containment Pressure 2.

Reactor Coolant Outlet Temperature T<oT (Wide Range) 3.

Reactor Coolant Inlet Temperature

- Tcoz> (Wide Range) 4.

Reactor Coolant Pressure

- Wide Range 5.

Pressurizer Water Level 6.

Steam Line Pressure 7.

Steam Generator Water Level - Narrow Range 8.

RWST Water Level 9.

~

Boric Acid Tank Solution Level

10. Auxiliary Feedwater Flow Rate ll. Reactor Coolant System Subcooling Margin Monitor 12.

PORV Position Indicator - Limit Switches 13.

PORV Block Valve Position Indicator - Limit Switches

14. Safety Valve Position Indicator - Acoustic Monitor

15. Incore Thermocouples (Core Exit Thermocouples)(4)

16. Reactor Coolant Inventory Tracking System (Reactor Vessel Level Indication)

17. Containment Sump Level+

18. Containment Water Level*

M M

M M(2)

M M

R Rt Rt R

R R

Rt R

R R

R(1) t R(3) t R

R Q

O (1)

Partial range channel calibration for sensor to be performed below P-12 in MODE 3.

(2)

With one train of Reactor Vessel Level Indication inoperable, Subcooling Margin Indication and Core Exit Thermocouples may be used to perform a CHANNEL CHECK to verify the remaining Reactor Vessel Indication train OPERABLE.

(3)

Completion of channel calibration for sensors to be performed below P-12 in MODE 3.

(4)

The core exit thermocouples will not be installed until the 1988 refueling outage; therefore, surveillances will not be required until that time.

See license amendment dated April 10, 1987.

The requirements for these instruments will become effective after the level transmitters are modified or replaced and become operational.

The schedule for modification or replacement of the transmitters is described in the Bases.

The provisions of Technical Specification 4.0.8 are applicable.

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REACTOR COOLANT SYSTEM 3 4 4 6

REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.6.1 The following Reactor Coolant System leakage detection systems shall be OPERABLE:

a.

One of the containment atmosphere particulate radioactivity monitoring channels (ERS-2301 or ERS-2401),

b.

The containment sump level and flow monitoring system, and c.

Either the containment humidity monitor or one of the containment atmosphere gaseous radioactivity monitoring channels (ERS-2305 or ERS-2405).

APPLICABILITY:

MODES 1, 2, 3 and 4

ACTION:

With only two of the above required leakage detection systems

OPERABLE, operation may continue for up to.30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required gaseous and/or particulate radioactivity monitoring channels are inoperable; otherwise, 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 UIREMENTS 4.4.F 1 The leakage detection systems shall be demonstrated OPERABLE by:

a

~

Containment atmosphere particulate and gaseous (if being used) monitoring system-performance of CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL FUNCTIONAL TEST at the frequencies specified in Table 4.3-3, b.

Containment sump level and flow monitoring system-performance of CHANNEL CALIBRATION at least once per 18 months,$

c.

Containment humidity monitor (if being used)

- performance of CHANNEL CALIBRATION at least once per 18 months, f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 4-14 AMENDMENT NO. W, ~,

159

'l J

4I I

lg

REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION Continued 2.

With two or more block valves inoperable, Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either (1) restore a total of at least two block valves to OPERABLE status, or (2) close the block valves and remove power from the block valves, or (3) close the associated PORVs and remove power from their associated solenoid valves; and apply the portions of ACTION a.2 or a.3 above for inoperable PORVs, relating to OPERATIONAL MODE, as appropriate.

c.

With PORVs and block valises not in the same line inoperable,*

within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either (1) restore the valves to OPERABLE status or (2) close and de-energize the other valve in each line.

Apply the portions of ACTION a.2 or a.3 above, relating to OPERATIONAL MODE, as appropriate for two or three lines unavailable.

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.4.11.1 Each of the three PORVs shall be demonstrated OPERABLE:

a.

At least once per 31 days by performance of a CHANNEL FUNCTIONAL TEST, excluding valve operation, and b.

At least once per 18 months by performance of a CHANNEL CALIBRATION.f 4.4.11.2 Each of the three block valves shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel.

The block valve(s) do not have to be tested when ACTION 3.4.11.a or 3.4.1l.c is applied.

4.4.11.3 The emergency power supply for the PORVs and block valves shall be demonstrated OPERABLE at least once per 18 months by operating the valves through a complete cycle of full travel while the emergency buses are energized by the onsite diesel generators and onsite plant batteries.

This testing can be performed in conjunction with the requirements of Specifications 4.8.1.1.2.e and 4.8.2.3.2.d.f t

PORVs isolated to limit RCS leakage through their seats and the block valves shut to isolate this leakage are not considered inoperable.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 4-33 AMENDMENT NO.

+kg 109

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued d ~

At least once per 18 months by:

1.

Verifying automatic i.solation and interlock action of the RHR system from the Reactor Coolant System when the Reactor Coolant System pressure is above 600 psig.

2.

A visual inspection of the containment sump and verifying that the subsystem sucti.on inlets are not restricted by debris and that the sump components (trash racks,

screens, etc.)

show no evidence of structural distress or corrosion.t e.

At least once per 18 months, during shutdown, by:t 1.

Verifying that each automatic valve in the flow path actuates to its correct position on a Safety Injection test signal.

2.

Verifying that each of the followi.ng pumps start auto-matically upon receipt of a safety injection test signal:

a)

Centrifugal charging pump b)

Safety injection pump c)

Residual heat removal pump By verifying that each of the following pumps develops the indicated discharge pressure on recirculation flow when tested pursuant to Specification 4.0.5:

1.

Centrifugal charging pump Greater than or equal to 2405 psig 2.

Saf ety Injection pump Greater than or equal to 1409 psig I

3.

Residual heat removal pump Greater than or equal to 190 psig By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:

1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS sub-systems are required to be OPERABLE.

The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT UNIT 2 3/4 5-5 AMENDMENT NO. 43+a ~t ~s

1

i ~

RGENCY COR C OL NG S

SURVEILLANC U REME 4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable Surveillance Requirements of 4.5.2.f 4.5.3.2 All charging pumps and safety in)ection pumps, except the above required OPERABLE charging pump, shall be demonstrated inoperable, by verifying that the motor circuit breakers have been removed from their electrical power supply circuits, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months whenever the temperature of one or more of the RCS cold legs is less than or equal to 152'F as determined at least once per hour when any RCS cold leg temperature is between 152'F and 200'F.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-8 AMENDMENT NO. SQ, 00, 434.

159

CO A NM IV DER ARRIE IMI ING COND T 0 R OPE T

N 3.6.5.9 The divider barrier seal shall be OPERABLE.

+C~N:

With the divider barrier seal inoperable, restore the seal to OPERABLE status prior to increasing the Reactor Coolant System temperature above 200'F.

S VE L CE RE U R 4.6.5.9 The divider barrier seal shall be determined OPERABLE at least once per 18 months during shutdown by:f a.

Removing two divider barrier seal test coupons and verifying that the physical properties of the test coupons are within the acceptable range of values shown in Table 3.6-2.

b.

Visually inspecting at least 95 percent of the seal's entire length and:

1.

Verifying that the seal and seal mounting bolts are pro-perly installed, and 2.

Verifying that the seal material shows no visual evidence of deterioration due to holes,

ruptures, chemical attack,

abrasion, radiation damage, or changes in physical appearances.

f. The provisions of'echnical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 6-47 AMENDMENT NO. m, ~,

159

PLANT SYSTEMS SURVEILLANCE RE UIREMENTS 4.7.1.2 Each auxiliary feedwater pump shall be demonstrated OPERABLE when tested pursuant to Specification 4.0.5 by:

a.

Verifying that each motor driven pump develops an equivalent discharge pressure of greater than or equal to 1240 psig at 60'F in recirculation flow.

b.

Verifying that the steam turbine driven pump develops an equivalent discharge pressure of greater than or equal to 1180 psig at 60'F and at a flow of greater than or equal to 700 gpm when the secondary steam supply pressure is greater than 310 psig.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

c.

Verifying that each non-automatic valve in the flow path that is not locked,

sealed, or otherwise secured in position is in its correct position.

Verifying that each automatic valve in the flow path is in the fully open position whenever the auxiliary feedwater system is placed in automatic control or when above 10X RATED THERMAL POWER.

This requirement is not applicable for those portions of the auxiliary feedwater system being used intermittently to maintain steam generator level.

e.

Verifying at least once per 18 months during shutdown that each automatic valve in the flow path actuates to its correct position upon receipt of the appropriate engineered safety features actuation test signal required by Specification 3/4.3.2.t Verifying at least once per 18 months during shutdown that each auxiliary feedwater pump starts as designed automatically upon receipt of the appropriate engineered safety features actuation test signal required by Specification 3/4.3.2.t g-Verifying at least once per 18 months during shutdown that the unit cross-tie valves can cycle full travel.

Following cycling, the valves will be verified to be in their closed positions.

The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-6 AMENDMENT NO.~, ~, +Se,15'

\\ ~ *

~

4 P

3 4 7 4 ESSENTIAL SERVICE WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.4.1 a.

'At least two independent essential service water loops shall be OPERABLE.

b.

At least one essential service water flowpath associated with support of Unit 1 shutdown functions shall be available.

APPLICABILITY: Specification 3.7.4.1.a.

MODES 1, 2, 3,

and 4.

Specification 3.7.4.1.b.

- At all times when Unit 1 is in MODES 1, 2, 3, or 4.

ACTION:

When Specification 3.7.4.1.a is applicable:

With only one essential service water loop OPERABLE, restore at least two loops 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 .

When Specification 3.7.4.1.b is applicable:

With no essential service water flow path available in support of Unit 1 shutdown functions, return at least one flow path to available status within 7 days or provide equivalent shutdown capability in Unit 1 and return the equipment to service within the next 60 days, or have Unit 1 in HOT STANDBY within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and HOT SHUTDOWN within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The requirements of Specification 3.0.4 are not applicable.

SURVEILLANCE RE UIREMENTS 4.7.4.1 At least two essential service water loops shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) servicing safety related equipment that is not locked,

sealed, or otherwise secured in position, is in its correct position.

b.

At least once per 18 months during shutdown, by verifying that each automatic valve servicing safety related equipment actuates to its correct position on a Safety Injection test signal.f t The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-13 AMENDMENT NO. O7, 109

SYT S

4 7

7 SNUBBE S

ING COND ION FO 0

TI N 3.7.7.1 All safety-related snubbers shall be OPERABLE'ystems required OPERABLE in those MODES).

~C~IO With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber(s) to OPERABLE status and perform an engineering evaluation per Specification 4.7.7.1.c on the supported component or declare the supported system inoperable and follow the appropriate ACTION statement for that system.

SURVEILLANCE RE UIREHENTS 4 '.7.1 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the requirements of Specification 4.0.5.

a.

Visua Ins ect'o Snubbers are categorized as inaccessible or accessible during reactor operation.

Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table 3.7-9.

The visual inspection interval for each type of snubber shall be determined based upon the criteria provided in Table 3.7-9 and the first inspection interval determined using this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Amendment No.

b.

Visual Ins ection Acce tance Criteria Visual inspections shall verify (1) that there are no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are secure, and (3) in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not frozen up.

Snubbers which appear inoperable as a result

,of visual inspections shall be classified as unacceptable and may be reclassified as acceptable for the purpose of establishing the next visual inspection interval, providing that (1) the cause of the rejection is clearly established and remedied for that The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-20 AMENDMENT NO. 402,48k, 456-,

159

PLANT SYSTEMS SURVEILLANCE RE U EMENTS Continued 4.7.9.2 Each of the above required water spray and/or sprinkler systems shall be demonstrated to be OPERABLE't least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel as provided by Technical Specification 4.7.9.1.l.e.

At least once per 18 months:t 1.

By performing a

system functional test which includes simulated automatic actuation of the system, and:

a) Verifying that the automatic valves in the flow path actuate to their correct positions on a test signal, and*

b) Cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

By visual inspection of deluge and preaction system piping (this is not required'or systems supervised by air) to verify their integrity.

3.

By visual inspection of each open head deluge nozzle to verify that there is no blockage.

At least once per 3 years by performing an air flow test through the piping of each open head deluge system and verifying each open head deluge nozzle is unobstructed.

t The provisions of Technical Specification 4.0.8 are applicable.

The fire protection water flow surveillance testing may be suspended until the completion of the fire protection water storage tank and fire pump installations (May 31, 1993).

The surveillance testing suspended as a result of this amendment will be initiated at its normal frequency within four months of the new fire protection water storage tanks and fire pumps being declared

OPERABLE, with the exception of unit outage required testing which would be completed before the end of the next scheduled unit outage.

COOK NUCLEAR PLANT - UNIT 2 3/4 7-31 AMENDMENT NO. ~ ~,~,

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C 4 1

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S S

SURV I C

RE UIR S

Con i ued a)

A kinematic viscosity of greater than or equal to 1.9 centistokes but less than or equal to 4.1 centistokes at 40'C (alternatively, Saybolt viscosity, SUS at 100'F of greater than or equal to 32.6 but less than or equal to 40.1), if gravity was not determined by comparison with supplier's certification.

b)

A flash point equal to or greater than 125'F.

2)

By verifying, in accordance with the test specified in ASTM D1298-80 and prior to adding the new fuel to the storage tanks, that the sample has either an API gravity of greater than or equal to 30 degrees but less than or equal to 40 degrees at 60'F or an absolute specific gravity at 60/60'F of greater than or equal to 0.82 but less than or equal to 0.88, or an API gravity of within 0.3 degrees at 60'F when compared to the supplier's certificate or a specific gravity of within 0.0016 at 60/60'F when compared to the supplier's certificate.

3)

By verifying, in accordance with the test specified in ASTM D4176-82 and prior to adding new fuel to the storage

tanks, that the sample has a clear and bright appearance with proper color.

4)

By verifying within 31 days of obtaining the sample that the other properties specified in Table 1 of ASTM D975-81 are within the appropriate limits when tested in accordance with ASTM D975-81 except that the analysis for sulfur may be performed in accordance with ASTM D2622-82.

d.

At least once per 31 days by obtaining a sample of fuel oil from the storage tanks in accordance with ASTM D2276-83, and verifying that total particulate contamination is less than 10 mg/liter when tested in accordance with ASTM D2276-83, Method A+.

e.

At least once per 18 months, during shutdown, by:f 1.

Subjecting the diesel engine to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby

service,

~e actions to be taken should any of the properties be found outside of the specified limits are defined in the Bases.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 8-4 AMENDMENT NO. 444,

)59

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'0 LECTR CA POWER

~HU'~DO I ITING CO TIO OR 0 E

0 3.8.1.2 As a minimum, the following A.C. electrical power sources shell be OPERABLE:

a.

One circuit between the offsite transmission network and the onsite Class lE distribution system, and b.

One diesel generator with:

1.

A day fuel tank containing a minimum of 70 gallons of fuel, 2.

A fuel storage system containing a minimum indicated volume of 46,000 gallons of fuel, and 3.

A fuel transfer pump.

P C B MODES 5 and 6.

ROTION:

With less than the above minimum required A.C. electrical power sources OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes* until the minimum required A.C. electrical power sources are restored to OPERABLE status.

SURVEILLANCE E UI E

4.8.1.2 The above required A.CD electrical power sources shall be demonstrated OPERABLE by the performance of each of the Surveillance Requirements of 4.8.1.1.1 end 4.8.1.1.2 except for requirement 4.8.1.1.2.e.5.$

For purposes of this specification, addition of water from the RWST does not constitute a positive reactivity addition provided the boron concentration in the RWST is greater than the minimum required by Specification 3.1.2.7.b.2.

f The provisions of Technical Specification 4.0.8 are applicable.

COOK NUCLEAR PLANT - UNIT 2 3/4 8-9 AMENDMENT NO. 4488 ~,

159

ML17331B201

Text

Attachment:

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