ML17333A813

Amends 214 & 199 to Licenses DPR-58 & DPR-74,respectively, Revising TS to Allow Operation of Unit 1 at Steam Generator Tube Plugging Levels Up to 30%

ML17333A813
Person / Time
Site:	Cook
Issue date:	03/13/1997
From:	John Hickman NRC (Affiliation Not Assigned)
To:
Shared Package
ML17333A814	List: ML17333A813 ML17333A815
References
NUDOCS 9703190231
Download: ML17333A813 (57)

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON> D.C. 20555-0001 NDIANA HIC IGAN POWER CO PANY IKWJ N DON C.

COOK NUC A

ANT IT NO.

1 HENDMENT TO FACILITY OPERATI G

ICENSE Amendment No.

214 License No.

DPR-58 l.

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

A.

B.

C.

D.

The application for amendment by Indiana Michigan Power Company (the licensee) dated Hay 26,

1995, and supplemented September 26,

1995, August 2,

1996, and February 6,

1997, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; The issuance of this. amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

9703f9023f 9703f3 PDR ADQCK 050003f5 PDR

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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-58 is hereby amended to read as follows:

echnical S eci ications 3.

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 214

, are hereby incorporated in the license.

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

This license amendment is effective as of the date of issuance, with full implementation to occur within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 13, 1997 ohn B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

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T CHMENT TO LICENSE A NDMENT NO. 214 TO AC ITY OPERATING LICENSE NO.

OPR-58 OCK NO. 50-3 5

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 2-2 2-5 2-7 2-8 2-9 B 2-,1(a)

B 2-4 B 2-5 3/4 1-1 3/4 1-15 3/4 1-16 3/4 2-14 3/4 3-17 3/4 3-21 3/4 3-23a 3/4 3-24 3/4 3-26 3/4 3-31

, 3/4 3-33 3/4 4-4 3/4 4-5 3/4 5-11 5-5 B, 3/4 1-1 B 3/4 4-1 B 3/4 5-3 B 3/4 6-2 INSERT 2-2'-5 2-7 2-8 2-9 B 2-1(a)

B 2-4 B 2-5 3/4 1-1 3/4 1-15 3/4 1-16 3/4 2-14 3/4 3-17 3/4 3-21 3/4 3-23a 3/4 3-24 3/4 3-26 3/4 3-31 3/4 3-33 3/4 4-4 3/4 4-5 3/4 5-11 5-5 B 3/4 1-1 B 3/4 4-1 B 3/4 5-3 B 3/4 6-2.

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2.0 'AFETYLIMITSAND LIMI'riNGSAFETY SYSTEM SETI'INGS TABL'E2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS FUNCTIONALUNIT 1.

Manual Reactor Trip 2.

Power Range, Neutron Flux 3.

Power Range, Neutron Flux, High Positive Rate 4.

Power Range, Neutron Flux, High Negative Rate TRIP SETPOINT Not Applicable Low Setpoint - less than or equal to 25% of RATED THERMAL POWER High Setpoint - less than or equal to 109% of RATED THERMAL POWER Less than or equal to 5% of RATED THERMALPOWER with a time constant greater than or equal to 2 seconds Less than or equal to 5% of RATED THERMALPOWER with a time constant greater than or equal to 2 seconds ALLOWABLEVALUES Not Applicable Low Setpoint - less than or equal to 26% of RATED THERMAL POWER High Setpoint - less than or equal to 110% of RATED THERMAL POWER Less than or equal to 5.5% of RATED THERMALPOWER with a time constant greater than or equal to 2 seconds Less than or equal to 5.5% of RATED THERMALPOWER with a time constant greater than or equal to 2 seconds 6.

Intermediate Range, Neutron Flux Source Range, Neutron Flux Less than or equal to 25% of RATED THERMALPOWER Less than or equal to 10'ounts per second Less than or equal to 30% of RATED THERMALPOWER Less than or equal to 1.3 x 10'ounts per second 7.

0vertemperature Delta T 8.

Overpower Delta T See Note 1

See Note 2 See Note 3 See Note 4 9.

Pressurizer Pressure-Low Greater than or equal to 1875 psig Greater than or equal to 1865 psig 10.

Pressurizer Pressure High 11.

Pressurizer Water Level-

- High 12.

Loss of Flow Less than or equal to 2385 psig Less than or equal to 92% of instrument span Greater than or equal to 90% of design flow per loop~

Less than or equal to 2395 psig Less than or equal to 93% of instrument span Greater than or equal to 89.1% of design flow per loop*

~Design fiow is I/4 Reactor Coolant System total fiow rate from Table 3.2-1.

COOK NUCLEAR PLANT-UNIT1 Page 2-5 AMENDMENTOa, uS, ~ ~~4

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TABLE2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS OTATION 1+Tts Note 1: Overtemperature hT 6 dT, tK,-K

'T-T') + K (P-P') -f, (II)]

1+TP where:

hT, Indicated hT at RATED THERMALPOWER p/

1+Tts 1+TP Tp Ta Average temperature, 'F indicated T~ at RATED THERMALPOWER (( 576.3'F)

Pressurizer pressure, psig Indicated RCS nominal operating pressure (2235 psig or 2085 psig)

The function generated by the lead-lag controller for T~ dynamic compensation Time constants utilized in the lead-lag controller for T,~

Tt = 22 secs.

T> ~ 4 secs.

Laplace transform operator

TABLE2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS NOTATIONS Continued Operation with 4 Loops K, = 1.17 Q = 0.0230 Ks = 0.00110 and f,(d,l) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i)

For q, - q, between -37 percent and +3 percent, f,(hi)=.0 (where q, and g are percent RATED THERMALPOWER in'he top and bottom halves of the core respectively, and q, + q, is total THERMALPOWER in percent of RATED THERMALPOWER).

(ii)

For each percent that the magnitude of(q, - qg exceeds -37 percent, the hT trip setpoint shall be automatically reduced by 0.33 percent of its value at RATED THERMALPOWER.

'I (iii)

For each percent that the magnitude of (q, - qg exceeds +3 percent, the 4T trip setpoint shall be automatically reduced by 2.34 percent of its value at RATED THERMALPOWER.

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TABLE2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATIONTRIP SETPOINTS NOTATION Continued v>S Note 2:

Overpower hT 6 hT,

[K~ - Ks T - K< (T - T') -fz (hl)]

'"3 where:

Indicated hT at RATED THERMALPOWER Average temperature, 'F Indicated T,~ at RATED THERMALPOWER (< 563.0'F) 1.083 Kq 0.0177/'F for increasing average temperature and 0 for decreasing average temperature 0.0015 for T ) T"; Ks=0 for T 6 T" wqS 1+~>S

~9 f, (dl)

The function generated by the rate lag controller for T,~ dynamic compensation Time constant utilized in the rate lag controller forT,vs = 10 secs.

Laplace transform operator Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 3.4 percent hT span Note 4: The channel's maximum trip point shall not exceed its computed trip point by more than 2.5 percent hT span.

BASES 2.0 SAFEIY LIMITSANDLIMI'HNGSAI'KITSYSTEM SETIINGS 2.1 SAFETY LIMITS BASES 4 Loop Operation Westinghouse Fuel (15x15 OFA)

(WRB-1 Conelation)

Correlation Limit Design Limit DNBR Safety Analysis LimitDNBR Typical Cell 1.17 1.23 L40 Thimble Cell" 1.17 1.22 1.42 The cutves of Figure 2.1-1 show the loci ofpoints ofTHERMALPOWER, Reactor Coolant System pressure and average temperature forwhich the minimum DNBR is no less than the applicable design DNBRlimit,or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

represents typical fuel rod represents fuel rods near guide tube COOK NUCLEAR PLANT-UNIT1 Page B 2-l(a)

AMENDMENTS,uO, m ~~4

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BASES 2.0 SAR'TY LIMITSANDLIMITINGSAH'TYSYSTEM SET'I'INGS SAFETY LIMITS BASES The Power Range Negative Rate Trip provides protection forcontrol rod drop accidents. Athigh power, a rod drop accident could cause local flux peahng which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate Tripwillprevent this from occurring by tripping the reactor. No credit is taken for operation of the Power Range Negative Rate Trip for those control rod drop accidents for which the DNBRs willbe greater than the applicable design limitDNBR value for each fuel type.

Intermediate and Source Ran e Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup.

These trips provide redundant protection to the low setpoint trip ofthe Power Range, Neutron Flux channels.

The source Range Channels willinitiate a reactor trip at about 10+'ounts per second, unless manually blocked when P-6 becomes active. The Intermediate Range Channels willinitiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMALPOWER unless manually blocked when P-10 becomes active.

No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtem rature Delta T The Overtemperature delta T trip provides core protection to prevent DNB for all combinations ofpressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for piping delays fmm the core to the loop temperature detectors.

With normal axial power distribution, this reactor trip limit is always below the core safety limit as shown in Figure 2.1-1. Ifaxial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations. in Table 2.2-1.

COOK NUCLEAR PLANT-UNIT1 Page B 2Q AMENDMENTm, m 214

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BASES 2.0 SAFETY LIMITSANDLIMITINGSAFETY SYSTEM SETTINGS t

LIMITINGSAFETY SYSTEM SETTINGS BASES Ove wer 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 ofwater with

'emperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

The overpower delta T reactor trip provides protection or back-up protection forat power steamline break events. Credit was taken for operation of this trip in the steamline break masslenergy 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-UNITl,

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3/4"LIMA'INGCONDITIONS FOR OPZRATION ANDSURVEILLANCERE(}UIREMENTS 3/4.1 REACIIVITYCONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN-TAVG GREATER THAN200'F LIMITINGCONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGINshall be greater than or equal to 1.3% Delta k/k.

APPLICABILITY:

ACTION:

MODES 1, 2, 3, and 4.

With the SHUTDOWN MARGINless than 1.3% Delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm of a solution containing greater than or equal to 20,000 ppm boron or equivalent until the required SHUTDOWN MARGINis restored.

SURVEILLANCERE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGINshall be determined to be greater than or equal to 1.3% Delta k/k:

Within one hour after detection of an inoperable control rods(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGINshall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b.

When in MODE 1 or MODE 2 with Keffgreater than or equal to 1.0, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limits of Specification 3.1.3.5.

C.

d.

When in MODE 2 with Keff less than 1.0, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticalityby verifying that the predicted critical control rod position is within the limits of SpeciQcation 3.1.3.5.

Prior to initialoperation above 5% RATEDTHERMALPOWER af'ter each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limitof Specification 3.1.3.5.

See Special Test Exception 3.10.1.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 1-1 AMENDMENT~, uO, m ~~4

0 3/4 LIMH'INGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMENTS 3/4.1 REACTIVH'YCONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITINGCONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:

A boric acid storage system and associated heat tracing with:

l.

A minimum usable borated water volume of 4300 gallons, 2.

Between 20,000 and 22,500 ppm of boron, and 3.

A minimum solution temperature of 145'F.

The refueling water storage tank with:

l.

2.

3.

A minimum usable borated water volume of 90,000 gallons, A minimum boron concentration of 2400 ppm, and A minimum solution temperature of70'F.,

APPLICABILITY:

ACTION:

MODES 5 and 6.

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONSor positive reactivity changes'ntil at least one borated water source is restored to OPERABLE status.

SURVEILLANCERE UIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:

ao At least once per 7 days by:

l.

2.

3.

Verifying the boron concentration of the water, Verifying the water level volume of the tank, and Verifyingthe boric acid storage tank solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water.

'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.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 1-15 AMENDMENTm, ~,21 4

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/4. LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.1 REACIIVH'YCONTROL SYSTEMS BORATED WATER SOURCES - OPERATIONS LIMITINGCONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE:

a.

b.

APPL'ICABILITYt

~CT ION:

A boric acid storage system and associated heat tracing with:

1.

A minimum usable borated water volume of 5650 gallons,'.

Between 20,000 and 22,500 ppm of boron, and 3.

A minimum solution temperature of 145'F.

The refueling water storage tank with:

l.

A minimum contained volume of 350,000 gallons of water, 2.

Between 2400 and 2600 ppm of boron, and 3.

A minimum solution temperature of 70'F.

MODES 1, 2, 3 and 4.

Withthe boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGINequivalent to at least 1%~ at 200'F; restore the boric. acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBYwithin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCERE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

COOK NUCLEAR PLANT-UNIT1 Page 3/4 1-16 AMENDMENT40,ua 214

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'I 3/4' LIMITINGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMENTS 3/4.2 POWER DISTRIBUXIONLIMITS TABLE3.2-1 DNB PARAMETERS ARAMETER Reactor Coolant System Tavg Pressurizer Pressure Reactor Coolant System Total Flow Rate LIMITS 4 Loops in Operation at RATED THERMALPOWER c 579.3 F R 2050 psig M 341,100 gpm" Indicated average of at least three OPERABLE instrument loops.

"Limitnot applicable during either,a THERMALPOWER ramp increase in excess of 5 percent RATED THERMALPOWER per minute or a THERMALPOWER step increase in excess of 10 percent RATED THERMALPOWER.

"1ndicated value.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 2-14 AMENDMENTSa, uO, m, ~214

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3/4 LIMITINGCONDITIONS FOR OPERATION ANDSURyglLLANCEREQUIIKMENTS 3/4.3 INSTRUMENTATION TABLE3 3-3 Continued ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATION FUNCTIONALUNIT f.

Steam Line Pressure-Low MINMUM TOTALNO. OF CHANNELS CHANNELS APPLICABLE CHANNELS TO TRIP OPERABLE MODBS ECTION Four Loops Operating, Three Loops Operating 1 pressure/loop 1 pressure/

operating loop 2 pressures any loops 1"'ressure in any operating loop 1 pressure 1,2,3" any 3 loops 1 pressure in 3"

any 2 operating loops 14 15 COOK NUCLEAR PLANT-UNIT1 Page 3/4 3-17 AMENDMENTa, uO, m

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3/4'mrrme CONDITIONS FOR OPERATION AND SURVEILLANCEazgUIREMI:mS, 3/4.3 INSTRUMENTATION TABLE3.3-3 Continued ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATION FUNCTIONALUNIT COINCIDENTWITH MINIMUM TOTALNO. OF CHANNELS CHANNELS APPLICABLE CHANNELS TO TRiP OPERABLE MODES ACTiON T<< Low-Low Four Loops Operating 1 Troop 2 T<<any loops I T<<any3 1,2,3" loops 14 Three Loops Operating 1 T<</operating loop 1"'<< in 1 T, in any 3"

any operating two operating loop loops 15 e.

Steam Line Pressure-Low Four Loops Operating 1 pressure/loop 2 pressures any loops 1 pressure 1; 2, 3" any 3 loops 14 Three Loops Operating

5. TURBINE TRIP &

FEEDWATER ISOLATION 1 pressure/

operating loop 1"'ressure ln any operating loop 1 pressure in 3'ny 2

operating loops 15 a.

Steam Generator 3/loop Water Level-High-High 2/loop in any 2/loop in 1,2,3 operating each loop operating loop 14'OOK NUCLEAR PLANT-UNIT1 Page 3/4 3-21 AMENDMENTm, uO, ~,214

3/4

'/4.3 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS INSTRUMENTATION ENGINEERED SAFETY FEATURES INTERLOCKS DESIGNATION P-11 P-12 CONDITION AND SETPOINT With 2 of 3 pressurizer pressure channels greater than or equal to 1915 psig.

With 2 of 4 T,~ channels less than or equal to Setpoint.

Setpoint greater than or equal to 541'F FUNCTION P-11 prevents or defeats manual block of safety injection actuation on low pressurizer pressure.

P-12 allows the manual block of safety injection actuation on low steam line pressure causes steam line isolation on high steam flow. Affects steam dump blocks.

With 3 of 4 T,~ channels above the reset point, prevents or defeats the manual block of safety injection actuation on low steam line pressure.'OOK NUCLEAR PLANT-UNIT1 Page 3/4 3-23a AMENDMENT~,214

3/4 LIMI'HNGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMINTS 3/4.3 INSTRUMENTATION TABLE3.3<

ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATIONTRIP SETPOINTS FUNCTIONALUNIT 1.

SAFETY INJECTION, TURBINETRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN FEEDWATER PUMPS TRIP SETPOINT ALLOWABLEVALUES a.

Manual Initiation b.

Automatic Actuation Logic Not Applicable See Functional Unit 9 Not Applicable c.

Containment Pressure-High Less than or equal to 1.1 psig Less than or equal to 1.2 psig d.

Pressurizer Pressure-Low Greater than or equal to 1815 psig Greater than or equal to 1805 psig e.

Differential Pressure Between Steam Lines-High f.

Steam Line Pressure-Low Less than or equal to 100 psi Greater than or equal to 500 psig steam line pressure Less than or equal to 112 psi Greater than or equal to 480 psig steam line pressure COOK NUCLEAR PLANT-UNIT1 Page 3/4 3-24 AMENDMENT49, 4K, 4A, 2] 4

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3/4 'IMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.3 INSTRUMENTATION TABLE3.3C Continued ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATIONTRIP SETPOINTS FUNCTIONALUNIT 2.

Containment Radio-activity-High Train A (VRS-1101, ERS-1301, ERS-1305) 3.

Containment Radio-activity-High Train B (VRS-1201, ERS-1401, ERS-1405)

4. STEAM LINEISOLATION TRIP SETPOINT See Table 3.3-6 See Table 3.3-6 LLOWABLEVALUES Not Applicable Not Applicable a.

Manual

b. Automatic Actuation Logic Not Applicable See Functional Unit 9 Not Applicable c.

Containment Pressure-High-High d.

Steam Flow in Two Steam Lines-High Coincident with T,~-Low-Low Less than or equal to 2.9 psig Less than or equal to 1.42 x 10'bs/hr from 0% load to 20% load.

Linear from 1.42 x 10'bs/hr at 20% load to 3.88 x 10~ lbs/hr at 100% load.

Less than or equal to 3 psig Less than or equal to 1.56 x 10~ lbs/hr fmm 0% load to 20% load.

Linear from 1.56 x 10'bs/hr at 20% load to 3.93 10'bs/hr at 100%

load.

e. Steam Line Pressure-Low

5. TURBINETRIP AND FEEDWATER ISOLATION a.

Steam Generator Water Level-High-High T~ greater than or equal to 541 F Greater than or equal to 500 psig steam line pressure Less than or equal to 67% of narrow-range instrument span each steam generator T~ greater than or equal to 539'F Greater than or equal to 480 psig steam line pressure Less than or equal to 68% of narrow-range instrument span each steam generator COOK NUCLEAR PLANT-UNIT1 Page 3/4 3-26 AMENDMENT94, m, ~,21 4

3/4'IMrI'INGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMENTS 3/4.3 INSTRUMENTATION TABLE4.3-2 ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATION SURVEILLANCERE UIREMENTS FUNCTIONALUNIT 1.

SAFETY INJECTION, TURBINETRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARYFEEDWATER PUMPS a.

Manual Initiation TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE CHECK CALIBRATION TEST See Functional Unit 9 b.

Automatic Actuation Logic N.A.

c.

Containment Pressure-High d.

Pressurizer Pressure-Low e.

Differential Pressure Between Steam Lines-High f.

Steam Line Prcssure-Low S

N.A.

M(2)

M(3)

N.A.

N.A.

I N.A.

N.A.

I, 2, 3, 4 1, 2, 3 1, 2, 3 "1,2, 3 1, 2, 3 2.

CONTAINMENTSPRAY a.

Manual Initiation b.

Automatic Actuation Logic N.A.

c.

Containment Pressure-High-High N.A.

R See Functional Unit 9 M(2)

M(3)

N.A.

N.A.

1,2,3,4 I, 2, 3 COOK NUCLEAR PLANT-UNIT1 Page 3/4 3-31 AMENDMENT400, 420,444,

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3/4' LIMITINGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMENTS 3/4.3 INSTRUMENTATION TABLE4.3-2 Continued

'ENGINEERED SAFETY FEATURE ACTUATIONSYSTEM INSTRUMENTATION SURVEILLANCERE UIREMENTS FUNCTIONALUNIT 4.

STEAM LINEISOLATION TRIP ACTUATING MODES IN CHANNEL DEVICE WHICH CHANNEL CHANNEL FUNCTIONAL OPERATIONAL SURVEILLANCE CBBCK CALlBBATION TEST a.

Manual b.

Automatic Actuation Logic N.A.

c.

Containment Pressure-High-High N.A.

See Functional Unit 9 M(2)

M(3)

N.A.

N.A.

1,2,3, 1, 2, 3 d.

Steam Flow in Two Steam Lines-High Coincident with T~-Low-Low e.

Steam Line Pressure-Low 5.

TURBINETRIP AND FEEDWATER ISOLATION a.

Steam Generator Water Level-High-High 6.

MOTOR DRIVEN AUXILIARYFEEDWATER PUMPS S

R N.A.

N.A.

N.A.

1, 2, 3 1, 2, 3 1,2,3 a.

Steam Generator Water Level-Low-Low

~

b.

4 kv Bus Loss of Voltage c.

Safety Injection d.

Loss of Main Feed Pumps N.A.

N.A.

N.A.

N.A.

M M(2)

R N.A.

N.A.

N.A.

N.A.

1, 2, 3 1, 2, 3 1, 2, 3 1,2 COOK NUCLEARPLANT-UNITI Page 3/4 3-33 AMENDMENTgg, ~, ~, ~, ~,214

3/4 "

LIMITINGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANTSYSTEM SAFETY VALVES-SHUTDOWN LIMITINGCONDITION FOR OPERATION 3.4.2 A minimum of one pressurizer code safety valve shall be OPERABLE with a liftsetting of2485 PSIG +

3%.

APPLICABILITY:

MODES 4 and 5.

ACTION:

With no pressurizer code safety valve OPERABLE:

ao Immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

b.

Immediately render all Safety Injection pumps and all but one charging pump inoperable by removing the applicable motor circuit breakers from the electric power circuit within one hour.

SURVEILLANCERE UIREMENTS 4.4.2 The pressurizer code safety valve shall be demonstrated OPERABLE per Surveillance Requirement 4 4.3.

'The liftsetting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

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 3.1.2.8.b.2 (MODE 4) or 3.1.2.7.b.2 (MODE 5).

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-4 AMENDMENT$$, 4K> ~" 4

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LIMHINGCONDITIONS FOR OPERATION ANDSURVEILLANCEREQUIREMENTS 3/4.4 REACTOR COOLANTSYSTEM SAFETY VALVES-OPERATING LIMITINGCONDITION FOR OPERATION 3.4.3 Allpressurizer code safety valves shall be OPERABLE with a liftsetting of 2485 PSIG + 3%.

APPLICABILITY:

ACTION:

MODES 1, 2 and 3.

With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCERE UIREMENTS 4.4.3 No additional surveillance requirements other than those required by Specification 4.0.5.

"Re liftsetting pressure shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

COOK NUCLEAR PLANT-UNIT1 Page 3/4 4-5 NDMENTuO, m,214

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3/4 LMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

REFUELING WATER STORAGE TANK LIMITINGCONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained volume of 350,000 gallons of borated water.

b.

Between 2400 and 2600 ppm of boron, and c.

A minimum water temperature of 70'F.

MODES 1, 2, 3 and 4.

ACTION:

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBYwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCERE UIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

a.

b.

At least once per 7 days by:

1.

Verifying the contained borated water level in the tank, and 2.

Verifying the boron concentration of the water.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature.

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DESIGN FEATURF>

5.4 REACTOR COOLANT SYSTEM Continued a.

In accordance with the code requirements specified in Section 4.1.6 ofthe 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 680'F.

VOLUME 5A.2 The total contained volume of the reactor coolant system is approximately 12,466 cubic feet at 0% steam generator tube plugging and 11,551 cubic feet at 30% steam generator tube plugging at a nominal T,~ of 70'F.

5.5 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:

b.

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

Anominal 8.97 inch center-to~uter distance between fuel assemblies placed in the storage racks.

C.

The fuel assemblies willbe classified as acceptable for Region 1, Region 2, or Region 3 storage based upon their assembly average burnup versus initial nominal enrichment.

Cells acceptable for Region 1, Region 2, and Region 3 assembly storage are indicated in Figures 5.6-1 and 5.6-2. Assemblies that are acceptable for storage in Region 1, Region 2, and Region 3 must meet the design criteria that define the regions as follows:

COOK NUCLEAR PLANT-UNIT1 Page 5-5 AMENDMENT~,m,~

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3/4 BASES 3/4.1 REACTIVITYCONTROL SYSTEMS

/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN enmes that 1) the reactor can be made subcritical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor willbe maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T~.

The most restrictive condition occurs at EOL, with T~ at no load operating temperature, and is associated with a postulated stcam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGINof 1.3% Delta k/k is initially required to control the reactivity transient and automatic ESF is assumed to be available.

With T,~ less than 200'F, the

'eactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% Delta k/k SHUTDOWN MARGINprovides adequate protection for this event.

The SHUTDOWNMARGINrequirements arc based upon the limitingconditions described above and are consistent with FSAR safety analysis assumptions.

3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 2000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes willbc gradual during boron concentration reductions in the Reactor Coolant System.

A flow rate of at least 2000 GPM willcirculate an equivalent Reactor Coolant System volume df 12,612 plus or minus 100 cubic feet in approximately 30 minutes.

The reactivity change rate associated with boron reductions willtherefore be within the capability for operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT T

The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid through each fuel cycle. The surveillance requirement for measurement ofthe MTC at the beginning, and near the end ofeach fuel cycle is adequate to confirm the MTCvalue since this coefficient changes slowly due principally to the reduction in RCS boron COOK NUCLEAR PLANT-UNIT1 Page B 3/4 1-1 AMENDMENTm, uO, ~, 214

0

3/4 'ASES 3/4A REACTOR COOLANTSYSTEM 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops in operation, aad maintain DNBR above the safety analysis limit during all normal operations and anticipated transients.

A loss of flow in two loops will cause a reactor trip ifoperating above P-7 (11 percent ofRATED THERMALPOWER) while a loss of flowin one loop willcause a reactor trip ifoperating above P4 (31 percent of RATED THERMALPOWER).

In MODE.3, a single reactor coolant loop provides sufficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

Three loops are required to be OPERABLE and to operate ifthe control rods are capable of withdrawal and the reactor trip breakers are closed.

The requirement assures adequate DNBR margin in the event of an uncontrolled rod withdrawal in this mode.

In MODES 4 aad 5, a single reactor coolant loop or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE. Thus, ifthe reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump below P-7 with one or more RCS cold legs less thaa or equal to 152'F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS willbc protected against ovcrpressure transients and willnot exceed the limitsofAppendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restricting starting of the RCPs to when the secoadary water temperature ofeach steam generator is less than 50'F above each of the RCS cold leg temperatures.

COOK NUCLEAR PLANT-UNIT1 Page B 3/44-1 'MENDMENTSS, uO, ~,

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3/4 BASES 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITYof the RWST as part of the ECCS ensures that sufficient negative reactivity is injected 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 injection 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 fiow to the core, and 2) the reactor willremain subcritical in the cold condition following a LOCA assuming mixing of the RWST, RCS, ECCS water, and other sources of water that may eventually reside in the sump, with all control rods assumed to be out. These assumptions are consistent with the LOCA analyses.

The contained water volume limitincludes an allowance for water not usable because oftank 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 of70'F.

This temperature value of the RWST water determines that of the spray water initiallydelivered to the containment followingLOCA. It is one ofthe 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.

COOK NUCLEAR PLANT-UNIT1 Page B 3/4 5-3 AMENDMENTu, uO, ~,214

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3/4'ASES 3/4.6 CONTAINMENTSYSTEMS 3/4.6.1.4 INTERNALPRESSURE The limitations on containment internal pressure ensure that

1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 8 psig and 2) the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.

The maximum peak pressure resulting from a LOCA event is calculated to be 11.49 psig, which includes 0.3 psig for initial positive containment pressure.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions and 2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specified for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initiallycontained air mass during a LOCA. The contained air mass increases with decreasing temperature.

The lower temperature limitof 60'F willlimitthe peak pressure to 11.49 psig which is less than the containment design pressure of 12 psig. The upper temperature limitinfluences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

3/4.6.1.6 CONTAINMENTVESSEL STRUCTURAL INTEGRITY This limitation ensures that the structural integrity of the containment steel vessel willbe maintained comparable to the original design standards for the lifeof the facility. Structural integrity is required to ensure that (1) the steel liner remains leak tight and (2) the concrete surraunding the steel liner remains capable ofproviding external missile protection forthe steel liner and radiation shielding in the event ofa LOCA. Avisual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

COOK NUCLEAR PLANT-UNlT1 Page B 3/4 6-2 AMENDMENTm 2~4

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NDIANA HICH GAN POWER COHPANY DOCKET NO. 50-316 ONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 HENDHENT TO F

C LITY OPERAT G LICENSE Amendment No. 199 License No.

DPR-74 The Nuclear Regulatory Commission (the Commission) has found that:

A.

B.

C.

D.

The application for amendment by Indiana Michigan Power Company (the licensee) dated Hay 26,

1995, and supplemented September 26,

1995, August 2,

1996, and February 6,

1997, 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; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such actjvities will be conducted in compliance with the Commission's regulations; The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

~ ~

e 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:

echnical S ec'fications 3.

The Technical Specifications contained in Appendices A and B,

as revised through Amendment No. 199, are hereby incorporated in the license.

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

This license amendment is effective as of the date of issuance, with full implementation to occur within 45 days.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 13, 1997 John B. Hickman, Project Manager Project Directorate III-3 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

TTACHM NT TO LICENSE AMENDMENT NO.

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 vertical lines indicating the area of change.

REMOVE 3/4 1-1 3/4 1-15 3/4 1-16 3/4 5-11 B 3/4 1-1 8 3/4 5-3 B 3/4 6-2 INSERT 3/4 1-1 3/4 1-15 3/4 1-16 3/4 5-11 B 3/4 1-1' 3/4 5-3 B 3/4 6-2,

0 3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.1 REACTIVITYCONTROL SYSTEMS 3/4.1.1 BORATION CONTROL SHUTDOWN MARGIN T y GREATER THAN 200 F LIMITINGCONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGINshall be greater than or equal to 1.3% Delta k/k.

A APPLICABILITY:

MODES I, 2, 3, and 4.

ACTION:

With the SHUTDOWN MARGINless than 1.3% Delta k/k, immediately initiate and continue boration at greater than or equal to 10 gpm ofa solution containing greater than or equal to 20,000 ppm boron or equivalent until the required SHUTDOWN MARGINis restored.

SURVEILLANCERE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGINshall be determined to be greater than or equal to 1.3% Delta k/k:

a0 b.

Within one hour after detection of an inoperable coritrol rod(s) and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereaf'ter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGINshall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

When in MODE 1 or MODE 2,with~ greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by verifying that control bank withdrawal is within the limitsof Specification 3.1.3.6.

C.

When in MODE 2 with~ less than 1.0, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> prior to achieving reactor criticalityby verifying that the predicted critical contml rod position is within the limits of Specification 3.1.3.6.

Prior to initialoperation above 5% RATEDTHERMALPOWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limitof Specification 3.1.3.6.

See Special Test Exception 3. 10.1.

COOK NUCLEAR PLANT-UNIT2 Page 3/4 1-1 AMENDMENTg2, 4', m,199

3/4 'IMH'INGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.1 REACTIVI'IYCONTROL SYSTEMS BORATED WATER SOURCES - SHUTDOWN LIMITINGCONDITION FOR OPERATION 3.1.2.7 As a minimum, one of the following borated water sources shall be OPERABLE:

a.

boric acid storage system and associated heat tracting with:

1.

A minimum usable borated water volume of 4300 gallons, 2.

Between 20,000 and 22,500 ppm of boron, and 3.

A minimum solution temperature of 145'F.

b.: The refueling water storage tank with:

1.

A minimum usable borated water volume of 90,000 gallons, 2.

A minimum boron concentration of 2400 ppm, and 3.

A minimum solution temperature of 70'F.

APPLICABILITY:

MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONSor positive reactivity changes until at least one borated water source is restored to OPERABLE status.

SURVEILLANCERE UIREMENTS 4.1.2.7 The above required borated water source shall be demonstrated OPERABLE:

At least once per 7 days by:

1.

Verifying the boron concentration of the water, 2.

Verifying the contained borated water volume, and 3.

Verifyingthe boric acid storage tank solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water.

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

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3/4 LIMI'HNGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.1 REACIIVITYCONTROL SYSTEMS BORATED WATER SOURCES - OPERATING LIMITINGCONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE:

b.

APPLICABILITY:

ACI'ION:

A boric acid storage system and associated heat tracing with:

1.

A minimum contained borated water volume of 5650 gallons, 2.

Between 20,000 and 22,500 ppm of boron, and 3.

A minimum solution temperature of 145'F.

The refueling water storage tank with:

l.

A minimum contained borated water volume of 350,000 gallons of water, 2.

Between 2400 and 2600 ppm of boron, and 3.

A minimum solution temperature of 70'F.

MODES 1, 2, 3 and 4.

Withthe boric acid storage system inoperable;restore the storage system to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at least 1% Delta k/k at 200'F; restore the boric acid storage system to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b.

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within one hour or be in at least HOT STANDBYwithh the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCERE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

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3/4 LIMITINGCONDITIONS FOR OPERATION AND SURVEILLANCEREQUIREMENTS 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS)

REFUELING WATER STORAGE TANK LIMITINGCONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained volume of 350,000 gallons of borated water, b.

Between 2400 and 2600 ppm of boron, and c.

A minimum water temperature of 70'F.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the refueling water storage tank inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBYwithin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCERE UIREMENTS 4.5.5 The RWST shall be demonstrated OPERABLE:

At least once per 7 days by:

1.

Verifying the contained borated water level in the tank, and 2.

Verifying the boron concentration of the water.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature.

COOK NUCLEAR PLANT-UNIT2 Page 3/4 5-11 AMENDMENT~, ~, 1 99

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3/4 'ASES 3/4.1 REACIXVITYCONTROL SYSTEMS 3/4.1.1 BORATION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made subcritical fiem all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor willbe maintained sufficiently subcritical to preclude inadvertent criticalityin the shutdown condition.

SHUTDOWN MARGIN requirements vary throughout core life as a function of fuel depletion, RCS boron concentration, and RCS T~. The most restrictive condition occurs at EOL, with T, at no load operating temperature, and is associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGINof 1.3% Delta k/k is initially required to control the reactivity transient and automatic ESF is assumed to be available.

4 With T,~ less than 200'F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% Delta k/k SHUTDOWN MARGINprovides adequate protection for this event.

The SHUTDOWNMARGINrequirements are based upon the limitingconditions described above and are consistent with FSAR safety analysis assumptions.

3/4.1.1.3 BORON DILUTION A minimum flow rate of at least 2000 GPM provides adequate mixing, prevents stratification and ensures that reactivity changes willbe gradual during boron concentration reductions in the Reactor Coolant System.

A flow rate of at least 2000 GPM will citculate an equivalent Reactor Coolant System volume of 12,612 cubic feet in approximately 30 minutes. The reactivity change rate associated withboron reductions willtherefore be within the capability for operator recognition and control.

COOK NUCLEAR PLANT-UNIT2 Page B 3/4 1-1 AMENDMENT8R, 408, 434

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3/4 BASES

~

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITYof the RWST as part of the ECCS ensures that sufficient negative reactivity is injected 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 injection by the ECCS in the event ofa 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 recuculation cooling flow to the core, and 2) the reactor will remain subcritical in the cold condition following a LOCA assuming mixing of the RWST,. RCS, ECCS water, and other sources of water that may eventually reside in the sump,.with all control rods assumed to be out.

These assumptions are consistent with the LOCA analyses.

The contained water volume limitincludes an allowance for water not usable because oftank discharge line location or other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value ofbetween 7.6 and 9.5 for the solution recirculated within containment after a LOCA. This pH band mininuzes the evolution of iodine and niinimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

8 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 temperanire value of the RWST water determines that of the spray water initiallydelivered to the containment followingLOCA. It is one ofthe 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.

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3/4 BASES 3/4.6 CONTAINMENTSYSTEMS 3/4.6.1.4 INTERNALPRESSURE 4

The limitations on containment internal pressure ensure that 1) the containment structure is prevented from exceeding its design negative pressure differential with respect to the outside atmosphere of 8 psig and 2) the containment peak pressure does not exceed the design pressure of 12 psig during LOCA conditions.

The maximum peak pressure resulting from a LOCA event is calculated to be 11.49 psig, which includes 0.3 psig for initial positive containment pressure.

3/4.6.1.5 AIR TEMPERATURE The limitations on containment average air temperature ensure that 1) the containment air mass is limited to an initial mass sufficiently low to prevent exceeding the design pressure during LOCA conditions and 2) the ambient air temperature does not exceed that temperature allowable for the continuous duty rating specifi'ed for equipment and instrumentation located within containment.

The containment pressure transient is sensitive to the initiallycontained air mass during a LOCA. The contained air mass increases with decreasing temperature.

The lower temperature limitof 60'F willlimit the peak pressure'o 11.49 psig which is less than the containment design pressure of 12 psig. The upper temperature limitinfluences the peak accident temperature slightly during a LOCA; however, this limit is based primarily upon equipment protection and anticipated operating conditions. Both the upper and lower temperature limits are consistent with the parameters used in the accident analyses.

/4.6.1.6 CONTAINMENTVESSEL STRUCTURAL INTEGRITY This limitationensures that the structural integrity ofthe containment willbe maintained comparable to the original design standards for the lifeof the facility. Structural integrity is required to ensure that (1) the steel liner remains leak tight and (2) the concrete surrounding the steel liner remains capable ofproviding external missile protection for the steel liner and radiation shielding in the event of a LOCA: A visual inspection in conjunction with Type A leakage tests is sufficient to demonstrate this capability.

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