ML17329A316

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Amends 158 & 142 to Licenses DPR-58 & DPR-74,respectively, Revising Tech Specs to Delete Refs to Bit
ML17329A316
Person / Time
Site: Cook  
Issue date: 11/20/1991
From: Marsh L
Office of Nuclear Reactor Regulation
To:
Shared Package
ML17329A317 List:
References
NUDOCS 9112110185
Download: ML17329A316 (53)
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Text

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II Ih 4y*~4 UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA MICHIGAN POMER COMPANY DOCKET NO. 50-315 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No. DPR-58 1.

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

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated March 26, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-58 is hereby amended to read as follows:

9iiZiiOi85 9iiiZO PDR ADOCN, 050003i5; P

PDR

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2 Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No. 158, are hereby incorporated in the license.

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

3.

This license amendment is effective prior to restart from the next refueling outage but no later than August 31, 1992.

FOR THE NUCLEAR REGULATORY COMMISSION QJN L. B. Marsh, Director l

Project Directorate III-I Division of Reactor Projects III/IY/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date.of Issuance:

November 2O I99]

ATTACHHENT TO LICENSE AMENDMENT NO.

15 FACILITY OPERATING LICENSE NO.

DPR-58 DOCKET NO. 50-315 Revise Appendix A Technical specifications by removing the pages identified below and inserting the attached pages.

The revised pages are identified by

'mendment number and contain marginal lines indicating the area of change.

REMOVE VII XIII 3/4 3-10 3/4 3-11 3/4 3-27 3/4 3-28 3/4 3-29 3/4 3-30 3/4 5'-9 3/4 5-10 5-5 B 2-5 B 3/4 3-1 B 3/4 3-1a B 3/4 3-1b B 3/4 3-1c B 3/4 3-2 B 3/4 3-3 B 3/4 3-4 B 3/4 3-5 B 3/4 5-2 B 3/4 5-3 INSERT VII XIII 3/4 3-10 3/4 3-11 3/4 3-27 3/4 3-28 3/4 3-29 3/4 3-30 3/4 5-9 3/4 5-10 5-5 B 2-5 B 3/4 3-1 B 3/4 3-2 B 3/4 3-3 B 3/4 3-4 B 3/4 3-5 B 3/4 3-6 B 3/4 3-7 B 3/4 5-2 B 3/4 5-3

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3 4.5 3/4.5.1 3/4.5.2 3/4.5.3 3/4.5.4 EMERGENCY CORE COOLING SYSTEMS ECCS ACCUMULATORS

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3/4 5

1 3/4 5-3 3/4 5-7 ECCS SUBSYSTEMS

- Tavg greater than or equal to 350 F ECCS SUBSYSTEMS

- Tavg less than 350 F.

~

BORON INJECTION SYSTEM 3/4.5.5 Intentionally Left Blank..

Intentionally Left Blank..

REFUELING WATER STORAGE TANK 3/4 5-9 3/4 5-10 3/4 5=11 3 4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity Containment Leakage Containment Air Locks.

'Internal Pressure Air Temperature Containment Structural Integrity.

Containment Ventilation Systems..

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4 3/4 3/4 3/4 3/4 3/4 3/4 6-1 6-2 6-4 6-6 6-7 6-9 6-9a 3/4. 6. 3 3/4.6.4 Containment Spray System..

5. ay Additive System CONTAINMENT ISOLATION VALVES.......

COMBUSTIBLE GAS CONTROL Hydrogen Analyzers.

Electric Hydrogen Recombiners W

3/4 6-10 3/4 6-12 3/4 6-14 3/4 6-23 3/4 6-24 COOK'NUCLEAR PLANT - UNIT 1 VII AMENDMENT NO. 77 N

2N,158

X5QEX I

I PPLIC B L

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B 3/4 0 1

3 4 ~ 1 REACTIVITY CONTROL S S

S 3/4'.1 'BORATZON CONTROL. ~ ~ ~ ~

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3/4.1.2 BORATION SYSTEMS.

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3/4.1,3 MOVABLE CONTROL ASSEMBLIES. ~.. ~. ~..

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B 3/4 1-1 B 3/4 1-2 B 3/4 1-3

.2 POWE DIST BUTION L S

3/4.2.1 3/4 2.2 3/4.2.4 3/4.2.5 3/4.2.6 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTORS o o ~. ~

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DNB PARAMETERS. ~..

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ALLOWABLE POWER LEVELS..

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B 3/4 2-4 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 AXIAL FLUX DZFFERENCE....

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B 3/4 2-1 3 4.3 INSTRUMENTATION 3/4.3.1 AND 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURE INSTRUMENTATZON... ~...

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B 3/4 3-1 3/4'.3 MONITORING INSTRUMENTATZON....

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, B 3/4 3-2 3

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REACTOR COOLANT SYSTE 3/4.4.1 REACTOR COOLANT LOOPS'

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3/4.4.2 and 3/4.4.3 SAFETY VALVES......

3/4 ~ 4 ~ 4 PRESSURI ZER a

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3/4.4

~ 5 STEAM GENERATOR TUBE INTEGRITY~

3/4 ~ 4(6 REACTOR COOLANT SYSTEM LEAKAGE.

3/4 ' '

CHEMISTRYo ~ ~

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3/4 ~ 4 ~ 8 SPECI FIC ACTIVITY~ ~ ~ ~ 0 ~ ~ ~ ~ ~ ~ ~ ~ ~

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B 3/4 4-1 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2a B 3/4 4-3 B 3/4 4-4 B 3/4 4-5 COOK NUCLEAR PLANT - UNIT 1 XZZ AMENDMENT NO+

S&g +&8;],5Q

TABLE 3.3-2 I

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT

RESPONSE

TIME

1. Manual Reactor, Trip 2.

Power Range, Neutron Flux (High and Low Setpoint)

NOT APPLICABLE Less than or equal to 0,5 seconds*

3.

Po~er Range, Neutron Flux, High Positive Rate 4.

Power Range, Neutron Flux, High Negative Rate NOT APPLICABLE Less than or equal to 0.5 seconds*

5. Intermediate
Range, Neutron Flux 6.

Source

Range, Neutron Flux
7. Overtemperature delta T
8. Overpower delta T

NOT APPLICABLE NOT APPLICABLE Less than or equal to 6.0 seconds*

r Less than or equal to 6.0 seconds*

9. Pressurizer Pressure--Low Less than or.equal to 1.0 seconds 10.Pressurizer Pressure--High Less than or equal to 1.0 seconds 11.Pressurizer Water Level--High Less than or equal to 2.0 seconds
  • Neutron detectors are exempt from response time testing.

Response

time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

COOK NUCLEAR PLANT - UNIT 1 3/4 3-10 AMENDMENT NO. W, 2N, XN 158

a TABLE 3.3-2 Continued REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT

RESPONSE

TIME 12.Loss of Flow - Single Loop (Above P-8)

Less than or equal to 1.0 seconds 13.Loss of Flow '- Two Loops (Above P-7 and below P-8) 14.Steam Generator Water Level--Low-Low 15.Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level Less than or equal to 1.0 seconds Less than or equal to 1.5 seconds NOT APPLICABLE 16.Undervoltage-Reactor Coolant Pumps Less than or equal to 1.2 seconds 17.Underfrequency-Reactor Coolant Pumps Less than or equal to 0.6 seconds 18.Turbine Trip A. Low Fluid Oil Pressure B. Turbine, Stop Valve 19.Safety Injection Input from ESF 20.Reactor. Coolant Pump Breaker Position Trip NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE COOK NUCLEAR PLANT - UNIT 1 3/4 3-11 AMENDMENT NO.

2/H, 158

TABLE 3.3-5 I

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS

1. Manual 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 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable 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 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable d.

Steam Line Isolation 2.

Containment Pressure-Hi h

Not Applicable a.

Safety Injection (ECCS) b, Reactor Trip (from SI) c.

Feedwater Isolation d.

Containment Isolation-Phase "A"

e.

Containment Purge and Exhaust Isolation f.

Auxiliary Feedwater Pumps g.

Essential Service Water System Less than or equal to 27.0QQ/27.0++

Less than or equal to 3.0 Less than or equal to 8.0 Less than or equal to 18.WI/28 OIJa Not Applicable Not Applicable Less than or equal to 13 Wl/48 O81I COOK NUCLEAR PLANT - UNIT 1 3/4 3-27 mmNDMENT gyes, 158

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS 3.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) b.

Reactor Trip (from SI) c.

Feedwater Isolation d.

Containment Isolation-Phase "A"

e.

Containment Purge and Exhaust Isolation f.

Auxiliary Feedwater Pumps g.

Essential Service Water System

'ess than or equal 27.089/27.0++

Less than or equal Less than or equal Less than or equal Not Applicable Not Applicable Less

$han or equal 48.0

/13.0P to to 3.0 to 8.0 to 18.04 to 4.

Differential Pressure Between Steam Lines-Hi h a ~

Safety Injection (ECCS) b.

C ~

d.

Reactor Trip (from SI)

Feedwater Isolation Containment Isolation-Phase "A"

e.

Containment Purge and Exhaust Isolation f.

Auxiliary Feedwater Pumps g.

Essential Service Water System Less than or equal 27.089/37.08 Less than or equal Less than or equal Less than or equal

18. Of/28. OPS Not Applicable

'ot Applicable Less than or equal 13.0f/48.0Pi to to 3.0 to 8.0 to to 5.

Steam Flow in Two Steam Lines - Hi h Coincident with Tav --Low-Low a ~

b.

C ~

d.

e.f.

go 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 Less than or equal 29.088/39.08 Less than or equal Less than or equal Less than or equal 20.0P/30.0ÃP Not Applica':.'

Not Applicable Less than or equal 15.0P/50.0ff Less than or equal to to 5.0 to 10.0 to to to 13.0 COOK NUCLEAR PLANT - UNIT 1 3/4 3-28 AMENDMENT NO. N

Ze7, 158

~

P TABLE 3.3-5 Continued I

ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS 6.

Steam Flow in Two Steam Lines-Hi h Coincident With Steam Line Pressure-Low a.

Safety Injection (ECCS) b.

Reactor Trip (from SI) c.

Feedwater Isolation d.

Containment Isolation-Phase "A"

e.

Containment Purge and Exhaust Isolation f.

Auxiliary Feedwater Pumps g.

Essential Service Water System h.

Steam Line Isolation 7.

Containment Pressure--Hi h-Hi h Less than or equal 27,0QQ/37.0Q Less than or equal Less than or equal Less than or equal 18.0¹/28.0¹¹ Not Applicable Not Applicable Less than or equal 14.0¹/48,0¹¹ Less than or equal to to 3.0 to 8.0 to to to 11.0 a.

b.

C.

d.

Containment Spray Containment Isolation-Phase "B"

Steam Line Isolation Containment Air Recirculation Fan Less than or equal to 45.0 Not Applicable Less than or equal to 10.0 Less than or equal to,660.0 8.

Steam Generator Water Level--Hi h-Hi h a.

Turbine Trip b.

Feedwater Isolation Less than or equal to 2.5 Less than or equal to 11.0 9.

Steam Generator Water Level--Low-Low

'a.

Motor Driven Auxiliary Feedwater Pumps b.

Turbine Driven Auxiliary Feedwater Pumps 10.

4160 volt Emer enc Bus Loss of Volta e Less than or equal to 60.0 Less than or equal to 60.0 a.

Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0 11.

Loss of Main Feedwater Pum s

a.

Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0

12. Reactor Coolant Pum Bus Undervolta e

a.

Turbine Driven Auxiliary Feedwater Pumps Less than or equal to 60.0 COOK NUCLEAR PLANT - UNIT 1 3/4 3-29 AMENDMENT NO. N, XN, lh7 158

TABLE 3.3-5 Continued TABLE NOTATION

¹ Diesel generator starting and sequence loading delays not included.

Offsite power available.

Response

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

¹¹ Diesel generator starting and sequence loading delays included.

Response

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

++ Diesel generator starting and sequence loading delays included.

Response

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

Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves

open, then VCT valves close) is NOT included.

Diesel generator starting and sequence loading delays included.

Response

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

Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves

open, then VCT valves close) is included.

QQ Diesel generator starting and sequence loading delays NOT included.

Offsite power available.

Response

time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging

pumps, Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves
open, then VCT valves close) is included.

COOK NUCLEAR PLANT - UNIT 1 3/4 3-30 AMENDMENT NO.

jib, 158

THIS PAGE INTENTIONALLYLEFT BLANK.

COOK NUCLEAR PLANT - UNIT 1 3/4 5-9 AMENDMENT NO.

pe, 158

0 I

I 1

THIS PAGE INTENTIONALLYLEFT BLANK.

COOK NUCLEAR PLANT - UNIT 1 3/4 5-10 AMENDMENT NO. $7, 158

DESIGN FEATURES a:

In accordance with the code requirements specified in Section 4.1.6 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 2485 psig, and c.

For a temperature of 650 F, except for the pressurizer which is 0

680 F.

VOLUME 5.4.2 5.5 The total contained volume of the reactor coolant system is 12,612

+ 100 cubic feet at a nominal T of 70 F.

avg EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.2 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, with one exception.

This exception is the CVCS boron makeup system and the BIT.

5.6 FUEL STORAGE CRITICALITY -

SPENT FUEL 5.6.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

A k equivalent to less than 0.95 when flooded with eff unborated water, b.

A nominal 10.5 inch center-to-center distance between fuel assemblies placed in the storage racks.

c.l.

A separate region within the spent fuel storage racks (defined as Region 1) shall be established for storage of Westinghouse fuel'ith nominal enrichment above 3.95 weight percent U-235 and with burnup less than 5,550 MWD/MTU.

In-Region 1, fuel shall be stored in a three-out-of-four cell configuration with one symmetric cell location of each 2 x 2 cell array vacant.

2.

The boundary between the Region 1 mentioned above and the rest, of the spent fuel storage racks (defined as Region 2) shall be such that the three-out-of-four storage requirement shall be carried into Region 2 by, at least, one row as shown in Figure 5.6-1.

COOK NUCLEAR PLANT - UNIT 1 5-5 AMENDMENT NO.

7S, XW, ZW, >5

LIMITING S TY SYSTEM SETTINGS BASES Over ower Delta T The Overpower delta T reactor trip provides assurance of fuel integrity, e.g.,

no melting, under all possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a backup to the High Neutron Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

The reference average temperature (T") is set equal to the full power indicated Tavg to ensure fuel integrity during overpower conditions for the range of full power average temperatures assumed in the safety analysis.

The overpower delta T reactor trip provides protection or back-up protection for at power steamline break events.

Credit was taken for operation of this

.trip in the steam line break mass/energy releases outside containment analysis.

In addition, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the reactor protection system.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.

The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The High Pressure trip provides protection for a Loss of External Load event.

The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves.

The pressurizer high water level trip precludes water relief for the Uncontrolled RCCA Withdrawal at Power event.

COOK NUCLEAR PLANT - UNIT l B 2-5 AMENDMENT NO.

ARAN ZRH, 158

3 4.3 INST NTATION BASES 3 4.3.1 and 3 4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES ESF INSTRUMENTATION The OPERABILITY of the protective and ESF instrumentation systems and interlocks ensure that 1) the associated ESF action and/or reactor trip will be initiated when the parameter monitored by each channel or combination thereof exceeds its setpoint,

2) the specified coincidence logic is maintained,
3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and 4) sufficient system'unctional capability is available for protective and ESF purposes from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable,

Response

time may. be demonstrated by any series of sequenti'al, over-lapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

ESF response times specified in Table 3,3-5 which include sequential operation of.

he RWST and VCT valves (Notes Q and QQ) are based on values assumed in the non-LOCA safety analyses.

These analyses take credit for injection of borated water from the RWST.

In5ection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves.

When sequential operation of the RWST and VCT valves is not included in the response times (Note ++), the values specified are based on the LOCA analyses, The LOCA analyses take credit for in]ection flow regardless of the source.

Verification of the response times specified in Table 3.3-5 will assure that the assumption used for VCT and RWST valves are valid.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-1 AMENDMENT NO, N t ZA ~

1 58

3 4.3.3 MONITORING INSTRUMENTATION 3 4.3.3.1 RADIATION MONITORING INSTRUMENTATION Noble gas effluent monitors provide information, during and following an accident, which is considered helpful to the operator in assessing the plant condition. It is desired that these monitors be OPERABLE at all time' during plant operation, but they are not required for safe shutdown of the plant.

In addition, a minimum of two in containment radiation-level monitors with a maximum range of 10 R/hr for photon only should be OPERABLE at all times except for cold shutdown and refueling outages.

In case of failure of the monitor, appropriate actions should be taken to restore its operational capability as soon as possible.

Table 3.3-6 is based on the-following Alarm/Trip Setpoints and Measurement Ranges for each instrument listed.

For the unit vent noble gas monitors, it should be noted that there is an automatic switchover from the low/mid-range channels to the high-range channel when the upper limits of the low-and mid-range channel measurement ranges are reached.

In this case there is no flow to the low-and mid-range channels from the unit vent sample line.

This is considered to represent proper operation of the monitor.

Therefore, if automatic switchover to the high-range should occur, and the low-and mid-range detectors are capable of functioning when flow is re-established, the low-and mid-range channels should not be declared inoperable and the ACTION statement in the Technical Specification does not apply.

This is also true while purging the low-and mid-range chambers following a large activity excursion prior to resumption of low-level monitoring and establishment of a new background.

INSTRUMENT ALARM/TRIP SETPOINT

1) Area Monitor-The monitor trip setpoint Upper Containment is based on 10 CFR 20 (VRS 1101/1201) limits.

A homogeneous mixture of the containment atmosphere is assumed.

The s<<tpoint value is defined as i<<e monitor reading when the purge is operating at the maximum flow rate.

MEASUREMENT RANGE*

10 R/hr to 10R/hr.

  • This is the minimum required sensitivity of the instrument.

Indicated values on these instruments above or below these minimum sensitivity ranges are acceptable and indicate existing conditions not instrument inoperability.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-2 AMENDMENT NO. Ni XN:158)

INSTRUMENTATION BASES Radiation Monitorin Instrumentation Continued INSTRUMENT

2) Area Monitor Containment High Range (VRA 1310/

1410)

ALARM/TRIP SETPOINT The monitor setpoint was selected to reflect the guidance provided in Generic Letter 83-37 for NUREG-0737 Technical Specifications MEASUREMENT RANGE*

1R/hr to 1 x 10 R/hr 7

Photons.

3) Process Monitor Particulate (ERS 1301/1401)

The monitor trip setpoint is based on 10 CFR 20 limits.

The set'point was determined using the Noble gas setpoint and historical monitor data of the ratio of particulates to Noble gases.

1.5 x 10 uCi to 7.5 Uci

4) Process Monitor Noble Gas (ERS 1305/1405)

-7 The monitor trip setpoint 5.8 x 10 2

's based on 10 CFR 20 2.7 x 10 limits.

A homogeneous mixture of the containment atmosphere is assumed.

The setpoint value is defined as the monitor reading when the purge is operating at the maximum flow rate.

uCi/cc to uCi/cc 5)

Steam Generator PORV (MRA 1601)

(MRA 1602)

(MRA 1701)

'MRA 1702)

Not Applicable.**

O.luCi/cc to 1.0xlO 2 uCi/cc.,

  • This is the minimum sensitivity of the instrument for normal operation, to follow the course of an accident, and/or take protective actions.

Values of the instrument above or below this minimum sensitivity range are acceptable.

    • These monitors are used to provide data to assist in post-accident off-site dose assessment.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-3 AMENDMENT NO. H XÃ,158 t

INSTRUMENTATION BASES Radiation Monitorin Instrumentation Coneinued INSTRUMENT

6) Noble Gas Unit Vent Monitors a)

Low Range (VRS 1505) b) Mid Range (VRS 1507) c) High Range (VRS 1509)

AIlQR/TRIP SETPOINT See Bases Section 3/4.3.3.10 Not Applicable~

Not Applicable~

MEASUKBKNT RANGE*

5.8x10 2uCi/cc to 2.7x10 BuCi/cc 1.3xl02 uCi/cc to 7.5x10 yCi/CC 2.9x104 uCi/cc eo 1.6x10 uCi/cc See Bases Section 3/4.3.3.10

8) Steam Jet Air Ejector Vent Noble Gas Monitor
7) Gland Steam Condenser Vent Noble Gas Monitor a)

Low Range (SRA 1805) 5.8x10 2uCi/cc to 2 'x10 uCi/cc a)

Low Range (SRA 1905) b) Mid Range (SRA 1907) c) High Range (SRA 1909)

See Bases Section 3/4.3.3.10 Not applicable.~

Not Applicable.~

5.8xlO uCi/cc to 2.7x10 uCi/

cc.

1.3xlO uCi/cc to 7.5x10 uCi/

-3 2

cc.

2.9x10'Ci/cc to 1.6x10 uCi/

,-2 4

cc

~

9)

Spene Fuel Storage (RRC-330)

The monitor seepoint is selected to alarm and trip consistent with 10 CFR 70.24(a)

(2 lx10 mR/hr to lx10 mR/hr

  • This is minimum sensitivity of the instrument for normal operaeion, to follow the course of an accident, and/or take protective actions.

Values of the instrument above or below this minimum sensitivity range are acceptable.

~ These monitors are used to provide daea eo assist in pose-accident off-site dose assessment.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-4 AMENDMENT NO. N XN 158

4 I

I

INSTRUMENTPP9:ON BASES Radiation Monitorin Instrumentation Continued The Radi.ation Monitoring -Instrumentation Surveillance Requirements per Table 4.3-3 are based on the following interpretation:

1)

The CHANNEL FUNCTIONAL TEST is successfully accomplished by the injection of a simulated signal into the channel, as close to the detector as practical, to verify the channel's alarm and/or trip function only.

2)

The CHANNEL CALIBRATION as defined in T/S Section 1.9 permi.ts the "known values" generated from radioactive calibration sources to be supplemented with "known values" represented by simulated signals for that subset of "known values" required for calibration and not practical to generate using the radioactive calibration sources.

3 4 '.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable in'core detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core.

3 4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the sei.smic instrumentation ensures that sufficient capabili.ty is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.

This capability is required to permit comparison of the measured response to that used in the design basis for the facility.

3 4.3.3.4 METEOROLOGICAL INSTRUMENTATION The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or tccidental release of radioactive materials to the atmospheres This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.

For the meteorological instrumentation, the required channel check consists of a qualitative assessment of channel behavior during operation by observation.

For the 10 m wind speed and wind direction instruments the channel check also includes, when possible, a comparison of channel indications.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-5 AHENDHENT No. jlfI,777,7/7 t

158

INSTRUMENTATION BASES 3 4.3.3.5 REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the remote shutdown instrumentation ensures that sufficient capability is available to p'ermit shutdown and maintenance of HOT STANDBY of the facility from locations outside of the control room.

This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.

3 4.3.3.5.1 APPENDIX R REMOTE SHUTDOWN INSTRUMENTATION The OPERABILITY of the Appendix R remote shutdown instrumentation ensures that sufficient instrumentation is available to permit shutdown of the facility to COLD SHUTDOWN conditions at the local shutdown indication (LSI) panel.

In the event of a fire, normal power to the LSI panels may be lost.

As a result, capability to repair the LSI panels from Unit 2 has been provided.

If the alternate power supply is not available, fire watches will be established in those fire areas where loss of normal power to the LSI panels could occur in the event of fire.

This will consist of either establishing continuous fire watches or verifying OPERABILITY of fire detectors per Specification 4.3.3.7 and establishing hourly fire watches.

The details of how these fire watches are to be implemented are included in a plant procedure.

3 4.3.3.7 FIRE DETECTION INSTRUMENTATION SYSTEMS DETECTORS OPERABILITY of the fire detection systems/detectors ensures that adequate detection capability is available for the prompt detection of fires.

This capability is required in order to detect and locate fires in their early stages.

Prompt detection of the fires will reduce the potential for damage to safety related systems or components in the areas of the specified systems and is an integral element in the 'overall facility fire protection program.

In the event that a portion of the fire detection systems is inoperable, the ACTION statements provided maintain the facility's fire protection program and allows for continued operation of the facility until the inoperable system(s)/detector(s) are restored to OPERABILITY.

However, it is not our intent to rely upon the compensatory. action for an extended period of time and action will be taken to restore the minimum number of detectors to OPERABLE status within a reasonable period.

3 4.3.3.8 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensure's that sufficient information is available on selected plant parameters to monitor and assess

" these variables during and following an accident.

The containment water level and containment sump level transmitters will be modified or replaced and OPERABLE by the end of the refueling outage to begin in February 1989.

+Amendment 112 (Effective before startup following refueling outage currently scheduled in 2/89).

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-6 AMENDMENT NO. 7)

(58

I I

~

~

INSTRUMENTZFCON BASES 3

4'.3 '

RADIOACTIVE LI UID EFFLUENT INSTRUMENTATION 3/4.3.3.9 The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases, The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11,3 of the Final Safety Analysis Report for the Donald CD Cook Nuclear Plant.

3 4.3.3.10 RADIOACTIVE GASEOUS EFFLUENT INSTRUMENTATION 3/4.3.3.10 The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases.

The alarm/trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to ensure that the alarm/trip will occur prior to exceeding the limits of 10 CFR Part 20'his instrumentation also includes provisions for monitoring 'the concentrations of potentially explosive gas mixtures in the waste gas holdup system.

The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria specified in Section 11,3 of the Final Safety Analysis Report for the Donald C.

Cook Nuclear Plant.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 3-7 AMENDMENT NO. gi),$58

EMERGENCY C8kE COOLING SYSTEM BASES With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumps, except the required OPERABLE charging, pump, to be inoperable below 170 F provides assurance that a mass addition pressure transient can be 0

relieved by the operation of a single PORV.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained.

Surveillance requirements for throttle'alve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 5-2 AMENDMENT NO.

gg, 158

EMERGENCY CORE COOLING 'YSTEMS BASES 3 4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that sufficient negative reactivity is in)ected into the core to counteract any positive increase in reactivity caused by RCS system cooldown, and ensures that a sufficient supply of borated water is available for in]ection by the ECCS in the event of a LOCA.

Reactor coolant system cooldown can be caused by inadvertent depressurization, a loss of coolant accident or a steam line rupture.

The limits on RWST minimum volume and boron concentration ensure that

1) sufficient water is available within,containment to permit recirculation cooling flow to the core, and 2) the reactor will remain subcritical in the

'cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.6 and 9.5 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

/

The ECCS analyses to determine F

limits in Specifications 3.2.2 and 3.2.6 assumed a

RWST water temperature of 70 F.

This temperature value of the RWST water determines that of the spray water initially'delivered to the containment following LOCA. It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance with the provisions of 10 CFR 50.46 and Appendix K to 10 CFR 50.

The value of the minimum RWST temperature in Technical Specification 3.5.5 has been conservatively changed to 80 F to increase the consistency between Units 1

0 and 2.

The lower RWST temperature results in lower containment pressure from containment spray and safeguards flow assumed to exit the break.

Lower containment pressure results in increased flow resistance of steam exiting the core thereby slowing'reflood and increasing PCT.

COOK NUCLEAR PLANT - UNIT 1 B 3/4 5-3 ENDMENT NO. py, ygp, 158

I AS flyboy

~c~

0 Cy O

.ri Op V/

+**y4 UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA MICHIGAN POMER COMPANY DOCKET NO. 50-316 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 142 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 Harch 26, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),

and the Commission's rules and regulations set forth,.in 10 CFR Chapter I; B.

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

'C.

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

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 2.

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.

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.

142, are hereby incorporated in the license.

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

3.

This license amendment is effective prior to restart from the next refueling outage but no later than August 31, 1992.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications 0

Date of Issuance:

November 2O 1991 Is)M~ 4 L. B. Marsh, Director Project Directorate III-1 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

t t

ATTACHMENT TO LICENSE AMENDMENT NO. 142 FACILITY OPERATING LICENSE 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 marginal lines indicating the area of change.

REMOVE VII XII 3/4 3-9 3/4 3-26 3/4 3-27 3/4 3-28 3/4 3-29 3/4 5-9 3/4 5-10 8 2-5 8 3/4 3-la 8 3/4 3-1b 8 3/4 3-lc 8 3/4 3-1d 8 3/4 5-2 8 3/4 5-3 INSERT VII XII 3/4 3-9 3/4 3-26 3/4 3-27 3/4 3-28 3/4 3-29 3/4 5-9 3/4 5-,10 8 2-5 8 3/4 3-la 8 3/4 3-1b 8 3/4 3-lc 8 3/4 3-1d 8 3/4 5-2 8 3/4 5-3

~

~

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE UIREMENTS SECTION PAGE 3 4.5 EMERGENCY CORE COOLING SYSTEMS ECCS 3/4.5.1 3/4.5.2 3/4.5.3 ECCS SUBSYSTEMS - Tavg greater than or equal ECCS SUBSYSTEMS - Tavg less than 350 F.....

to 350 F.

CCUMULATORS

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

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A 3/4 5-1 3/4 5-3 3/4 5-7 Containment Leakage

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Containment Air Locks

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Internal Pressure.

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Air Temperature Containment Structural Integrity Containment Ventilation System.

3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System Spray Additive System.......,....,

3/4.5.4 BORON INJECTION SYSTEM Intentionally Left Blank.............................

~

Intentionally Left Blank.

3/4.5.5 REFUELING WATER STORAGE TANK 3 4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Containment Integrity 3/4 5-9 3/4 5-10 3/4 5-11 3/4 6-1 3/4 6-2 3/4 6-4 3/4 6-6 3/4 6-7 3/4 6-9 3/4 6-9a 3/4 6-10 3/4 6-11 3/4.6.3 3/4.6.4 CONTAINMENT ISOLATION VALVES COMBUSTIBLE GAS CONTROL Hydrogen Analyzer Electric Hydrogen Recombiners

- W..

3/4 6-13 3/4 6-33 3/4 6-34 COOK NUCLEAR PLANT - UNIT 2 VII AMENDMENT NO

~~> ~~~, f42

26KR ZKGQH C

V 0

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o B 3/4 0 1 3/4.1.1 3/4.1.2 3/4.1.3 4

2 3/4.2.1 3/4.2.2 3/4.2.4 3/4.2.5 3/4.2.6 4

3 BORATION CONTROL

~

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BORATION SYSTEMS MOVABLE CONTROL ASSEMBLIES OWE D

S BUT ON T

AXIAL FLUX DIFFERENCE and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY HOT CHANNEL FACTOR QUADRANT POWER TILT RATIO DNB PARAMETERS e

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ALLOWABLE POWER LEVEL INS UME TIO

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B 3/4 1-1 B 3/4 1-2 B 3/4 1-4 B 3/4 2-1 B 3/4 2-4 B 3/4 2-5 B 3/4 2-5 B 3/4 2-5 3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED FEATURE INSTRUMENTATION SAFETY 3/4.3.3 3/4'.4 3 4 3/4.4.1 3/4.4.2 MONITORING INSTRUMENTATION...........

TURBINE OVERSPEED PROTECTION REACTOR COOIANT SYSTEM REACTOR COOIANT LOOPS and 3/4.4.3 SAFETY VALVES 3.4.4.4 PRESSURIZER 3/4.4.5 STEAM GENERATOR TUBE INTEGRITY.

3/4.4.6 REACTOR COOIANT SYSTEM ~GE 3.4.4.7 CHEMISTRY

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3/4.4.8 SPECIFIC ACTIVITY..................................,

B 3/4 3-1 B 3/4 3-la B 3/4 3-4 B 3/4 4-1 B 3/4 4-2 B 3/4 4-2 B 3/4 4-2a B 3/4 4-3 B 3/4 4-4 B 3/4 4-5 COOK NUCLEAR PIANT - UNIT 2 XII AMENDMENT NO. M, ~,

142

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT

RESPONSE

TIME 1.

Manual Reactor Trip

- NOT APPLICABLE 2.

Power Range, Neutron Flux (High and Low Setpoint)

Less'than or equal to 0.5 seconds*.

3.

Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE 4.

Power Range, Neutron Flux High Negative Rate r

5.

Intermediate

Range, Neutron Flux 6.

Source

Range, Neutron Flux 7.

Overtemperature Delta T Less than or equal to 0.5 seconds*

NOT APPLICABLE NOT APPLICABLE Less than or equal to 6.0 seconds*

8.

Overpower Delta T Less than or equal to 6.0 seconds*

9.

Pressurizer Pressure--Low Less than or equal to 2.0 seconds

10. Pressurizer Pressure--High Less than or equal to 2.0 seconds ll. Pressurizer Water Level--High Less than or equal to 2.0 seconds
  • Neutron detectors are exempt from response, time testing.

Response

time of the neutron flux signal portion of the channel shall be measured from de ector output or input of first electronic component in channel.

r COOK NUCLEAR PIANT - UNIT 2 3/4 3-9 AMENDMENT NO. yg, gag, 142

I

TABLE. 3. 3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS

1. Manual 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 Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable b.

Containment Spray Containment Isolation-Phase "B"

Containment Purge and Exhaust Isolation Containment Air Recirculation Fan Not Applicable Not Applicable Not Applicable Not Applicable c.

Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation Not Applicable Not Applicable d.

Steam Line Isolation 2.

Containment Pressure-Hi h

Not Applicable a.

b.

c ~

d.

e.f.

g 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 Less than or equal to 27.0QQ/27.0++

Less than or equal to 3.0 Less than or equal to 8.0 Not Applicable Not Applicable Not Applicable Not Applicable COOK NUCLEAR PLANT - UNIT 2 3/4 3-26 AMENDMENT No. rye, pe 142

0 k

C L

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS 3.

Pressurizer Pressure.-Low a.

Safety Injection (ECCS) c

~

d.

e

~

Reactor Trip (from SI)

Feedwater Isolation Containment Isolation - Phase "A"

Containment Purge and Exhaust Isolation Motor Driven Auxiliary Feedwater Pumps Essential Service Water System Less than or equal 27

~ OIQ/27.0++

Less than or equal Less than or equal Less than or equal

'to to 3.0 to 8.0 to 18.0¹ Not Applicable Less than or equal to 60.0 Less than or equal to 48.0++/13.0¹ 4.

Differential Pressure Between Steam Lines - Hi h a.

Safety Injection (ECCS) b.

Reactor Trip (from SI) c.

Feedwater Isolation d.

Containment Isolation - Phase "A"

Less than or equal 27.0QQ/37.0Q Less than or equal Less than or equal Less than or equal 18.0¹/28.0¹¹ to to 3 '

to 8 '

to ga Containment Purge and Exhaust Isolation Motor Driven Auxiliary Feedwater Pumps Essential Service Water System Not Applicable Less than or equal to 60.0 Less than or equal to 13.0¹/48.0¹¹ 5.

Steam Flow in Two Steam Lines - Hi h Coincident with Tav --Low-Low a.

b.

c

~

d.

f.

g

~

h.

Safety Injection (ECCS)

Reactor Trip (from SI)

Feedwater Isol>>" ton 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 Less than or equal to 13.0 I

COOK NUCLEAR PLANT - UNIT 2 3/4 3-27 AMENDMENT NO. g$

Ng X7$

142,

I

TABLE 3.3-5 Continued ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

RESPONSE

TIME IN SECONDS 6.

Steam Line Pressure--Low a.

Safety Injection (ECCS) e.

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

Steam Line Isolation 7.

Containment Pressure--Hi h-Hi h b.

Reactor Trip (from SI) c.

Feedwater Isolation

d. 'ontainment Isolation-Phase "A"

Less than or equal 27.0QQ/37.0Q Less than or equal Less than or equal Less than or equal 18.0¹/28.0¹¹ Not Applicable Less than or equal Less than or equal 14.0¹/48.0¹¹ Less than or equal to to 3.0 to 8.0 to to 60 '

to to 11.0 a

~

b.

C.

d.

Containment Spray Containment Isolation-Phase "B"

Steam Line Isolation Containment Air Recirculation Fan Less than or equal to 45.0 Not Applicable Less than or equal to 10.0 Less than or equal to 600.0 8.

Steam Generator Water Level--Hi h-Hi h a.

Turbine Trip b.

Feedwater Isolation Less than or equal to 2.5 Less than or equal to 11.0 9.

Steam Generator Water Level--Low-Low a.

Motor Driven Auxiliary Feedwater Pumps b.

Turbine Driven Auxiliary Feedwater Pumps Less than or equal to 60.0 Less than or equal to 60.0 10.

4160 volt Emer enc Bus Loss c

Volta e a.

Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60 '

11.

Loss of Main Feedwater Pum s

a.

Motor Driven Auxiliary Feedwater Pumps Less than or equal to 60.0

12. Reactor Coolant Pum Bus Undervolta e

a.

Turbine Driven Auxiliary Feedwater Pumps COOK NUCLEAR PIANT - UNIT 2 Less than or equal to 60.0 134 3/4 3-28 AMENDMENT NO. jig, gP7

@gal 142

TABLE 3.3-5 Continued TABLE NOTATION

¹ Diesel generator starting and sequence loading delays not included.

Offsite power available.

Response

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

¹¹ Diesel generator starting and sequence loading delays included.

Response

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

++ Diesel generator starting and sequence loading delays included.

Response

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

Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves

open, then VCT valves close) is NOT included.

6 Diesel generator starting and sequence loading delays included.

Response

time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps.. Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves

open, then VCT valves close) is included.

Qe Diesel generator starting and sequence loading delays NOT included.

Offsite power available.

Response

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

Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves

open, then VCT valves close) is included.

COOK NUCLEAR PLANT - UNIT 2 3/4 3-29 AMENDMENT NO.

142

~ ~

THIS PAGE INTENTIONALLYLEFT BLP2K.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-9 AMENDMENT NO.

graf,

].4Z

THIS PAGE INTENTIONALLYLEFT BLANK.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-10 AMENDMENT NO

~~,

14'.

LIMITING SAFETY SYSTEM SETTINGS BASES Over ower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g.,

no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

The reference average temperature (T) is set equal to the full power indicated Tavg to ensure fuel integrity during overpower conditions for the range of full power average temperatures assumed in the safety analysis.

The overpower delta T

reactor trip provides protection or back-up protection for at-power steam line break events.

Credit was,taken for operation of this trip in the steam line break mass/energy releases outside containment analysis.

In addition, its functional capability at the specified trip setting is required by this specification to enhance'the overall reliability of the Reactor Protection System.

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.

The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressuie for these valves (2485 psig).

The High Pressure trip provides protection for a Loss of External Load event.

The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves.

The pressurizer high water level trip precludes water relief for the uncontrolled control rod assembly bank withdrawal at-power event.

COOK NUCLEAR PLANT - UNIT 2 B 2-5 AMENDMENT NO.

'o t '

j

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INSTRUMENTATION Continued BASES ESF response times specified in Table 3.3-5 which include sequential operation of the RWST and VCT valves (Notes Q and QQ) are based on values assumed in the non-LOCA safety analyses.

These analyses take credit for injection of borated water from the RWST.

Injection of borated water is assumed not to occur until the VCT charging pump suction valves are closed following opening of the RWST charging pump suction valves.

When sequential operation of the RWST and VCT valves is not included in the response times (Note +t.), the values specified are based on the LOCA analyses.

The LOCA

.analyses take credit for injection flow regardless

-of the source.

Verification of the response ti'me's'specified in Table 3.3-5 will assure that the assumption used for VCT and RWST valves are valid.

3 4.3.3 MONITORING INSTRUMENTATION 3 4.3.3.1 RADIATION MONITORING INSTRUMENTATION Noble gas effluent monitors provide information, during and following an accident, which is considered helpful to the operator in assessing the plant condition. It is desired that these monitors be OPERABLE at all times during plant operation, but they are not required for safe shutdown of the plant.

In addition, a minimum of two in containment radiation-level monitors with a maximum range of 10 R/hr for photon only should be OPERABLE at all times except for cold shutdown and refueling outages.

In case of failure of the monitor, appropriate actions should be taken to restore its operational capability as soon as possible.

Table 3.3-6 is based on the following Alarm/Trip,Setpoints and'easurement Ranges for each instrument listed.

For the unit vent noble gas monitors, it should be noted that there is an automatic switchover from the low/mid-range channels to the high-range channel when the upper limits of the low-'and mid-range channel measurement ranges are reached.

In this case there is no flow to the low-and mid-range channels from the unit vent sample line.

This is considered to represent proper operation of this monitor.

Therefore, if automatic switchover to the high-range should occur, and the low-and mid-range detectors are capable of functioning when flow is re-established, the low-and mid-range channels should not be declared inoperable and the ACTION statement in the Technical Specification does not apply.

This is also true while purging the low-and mid-range chambers following a large activity excursion prior to resumption of low-level monitoring and establishment of a new background.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 3-la AMENDMENT NO.SP, AXED,],g

0 t ~

I f

vi

INSTRUMENTATION C)

BASES Radiation Monitorin Instrumentation Continued INSTRUMENT

1) Area Monitor-Upper Containment (VRS 2101/2201)
2) Area Monitor-Containment High Range (VRA 2310/

2410)

3) Process Monitor Particulate (ERS 2301/2401)
4) Process Monitor Noble Gas (ERS 2305/2405)

ALARM/TRIP SETPOINT The monitor trip setpoint is based on 10 CFR 20 limits.

A homogenous mixture of the containment atmosphere is assumed.

The setpoint value is defined as the monitor reading when the purge is operating at the maximum flow rate.

The monitor setpoint was selected to reflect the guidance provided in Generic Letter 83-37 for NUREG-0737 Technical Specifications.

The monitor trip setpoint is based on 10 CFR 20 limits.

The setpoint was detemined using the Noble gas setpoint and historical monitor data of the ratio of particulate to Noble gases.

The monitor trip setpoint; is based on 10 CFR 20 limits.

A homogenous mixture of the containment atmosphere is assumed.

The setpoint value is defined as the monitor reading when the purge is operating at the maximum flow rate.

MEASUREMENT RANGE*

10 R/hr to 10R/hr.

1R/hr to 1 x 10 R/hr 7

Photons.

1.5x10 uCi to 7.5 uCi.

5.8x10 uCi/cc to 2.7xlO uCi/cc

5) Steam Generator PORV Not Applicable.**

(MRA 2601)

(MRA 2602)

(MRA 2701)

(MRA 2702)

O.luCi/cc to 1,0x10 2 uCi/cc.

  • This is the minimum required sensitivity of the instrument.

Indicated values on these instruments above or below these minimum sensitivity ranges are acceptable and indicate existing conditions not instrument inoperability.

    • These monitors are used to provide data to assist in post-accident off-site dose assessment.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 3-lb AMENDMENT NO. 142

~

~

INSTRUMENTATION BASES Radiation Monitorin Instrumentation Continued INSTRUMENT ALARM/TRIP SETPOINT MEASUREMENT RANGE*

6) Noble Gas Unit Vent Monitors')

Low Range (VRS 2505) b) Mid Range (VRS 2507) c) High Range (VRS 2509)

See Bases Section 3/4.3.3.10 Not Applicable**

Not Applicable**

5.8x10 uCi/cc to 2.7x10 uCi/cc.

1.3x102 uCi/cc to

7. 5x10 yCi/CC 2.9xl04 uCi/cc to 1.6x10 uCi/cc See Bases Section 3/4.3.3.10
7) Gland Steam Condenser Vent Noble Gas Monitor a)

Low Range (SRA 2805) 5.8x10 2uCi/cc to 2.7x10 uCi/cc.

8)

Steam Jet Air Ejector Vent Noble Gas Monitor a)

Low Range (SRA 2905) b) Mid Range (SRA 2907) c) High Range (SRA 2909)

9) Spent Fuel Storage (RRC-330)

See Bases Section 3/4.3.3.10 Not applicable.**

Not Applicable.**

The monitor setpoint is

,selected to alarm and trip consistent with 10 CFR 70.24(a)

(2) 5.8x10 uCi/cc to 2,7x10 uCu cc.

1.3xlO uCi/cc to 7.5x10 uCi/

-3 2

cc.

2.9x10 uCi/cc to 1.6x10 uCi/

-2 4

cc.

lx10 mR/hr to lx10 mR/hr

-1 4

  • This is minimum sensitivity of the instrument for normal operation, to follow the course of an accident, and/or take protective actions, Values of the instrument above or below this minimum sensitivity range are acceptable.
    • These monitors are used to provide data to assist in post-accident off-site dose assessment.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 3-lc AMENDMENT NO. gp

isa, 14$

INSTRUMENTATION BASES The Radiation Monitoring Instrumentation Surveillance Requirements per Table 4.3-3 are based on the following interpretation:

1)

The CHANNEL FUNCTIONAL TEST is successfully accomplished by the in)ection of a simulated signal into the channel, as close to the detector as practical, to verify the channel's alarm and/or trip function only.

2)

The CHANNEL CALIBRATION as defined in T/S Section 1.9 permits the "known values" generated from radioactive calibration sources to be supplemented with "known values" represented by simulated signals for that subset of "known values" required for calibration and not practical to generate using the radioactive calibration sources.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 3-1d AMENDMENT NO. f/), ]qp

EMERGENCY CQQE COOLING SYSTEMS BASES 0

Mich the RCS tern'perature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the limited core cooling requirements.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety injection pumps, except the required OPERABLE charging pump, to be inoperable below 152 F provides assurance that a mass addition pressure transient can be 0

relieved by the operation of a single PORV.

The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that at a minimum, the assumptions used in the safety analyses are met and that.subsystem OPERABILITY is maintained.

Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow 'resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses.

COOK NUCLEAR PLY' UNIT 2 B 3/4 5-2 AMENDMENT NO N, 142

S S

B EMERGENCY COOLING SYSTEMS BASES 3 4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the RWST as part of the ECCS ensures that sufficient negative reactivity is in]ected into the core to counteract any positive increase in reactivity caused by RCS system

cooldown, and ensures that a sufficient supply of borated water is available for in)ection by the ECCS in the event of a LOCA.

Reactor coolant system cooldown can be caused by inadvertent depressurization,.

a LOCA or steam line rupture.

The limits of RWST minimum volume and boron concentration ensure that 1) sufficient water is available within containment to permit recirculation cooling flow to the

core, and 2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive control assembly.

These assumptions are consistent with the LOCA analyses.

The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a

pH value of between 7.6 and 9.5 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The ECCS analyses to determine F

limits in Specifications 3.2.2 and 3.2.6 assumed a

RWST water temperature of 80 F.

This temperature value of the 0

RWST water determines that of the spray water initially delivered to the containment following LOCA. It is one of the factors which determines the containment back-pressure in the ECCS analyses, performed in accordance with the provisions of 10 CFR 50 '6 and Appendix K to 10 CFR 50.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 5-3 AMENDMENT NO

)P7, 142

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