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Amends 148 & 134 to Licenses DPR-58 & DPR-74,respectively, Modifying Tech Specs to Achieve Consistency Between Both Units,Proposing Administrative Changes & Corrections & Allowing Westinghouse 17 X 17 Vantage 5 Fuel for Unit 2
	ML17328A426
Person / Time
Site:	Cook
Issue date:	08/27/1990
From:	Pierson R; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML17328A427	List: ML17328A426; ML17328A428;
References
	NUDOCS 9009060149
	Download: ML17328A426 (91)
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA MICHIGAN POWER COMPANY DOCKET NO. 50-315 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 148 License No.

DPR-58 1.

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

The applications for amendment by Indiana Michigan Power Company (the licensee) dated February 6, 1990 (as supplemented May 29, 1990) and May ll, 1990, comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

C.

D.

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 E.

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

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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:

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 148

, are hereby incorporated in the license.

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

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

August 27, 1990 Changes to the Technical Specifications Robert Pierson, Director Project Directorate III-1 Division of Reactor Projects - III, IV, V 8.Special Projects Office of Nuclear Reactor Regulation Date of Issuance:

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O. C. 20555 INDIANA MICHIGAN POWER COMPANY DOCKET NO.

50-316 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. f34 License No.

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

A.

The application for amendment by Indiana Michigan Power Company (the licensee) dated February 6, 1990, and as supplemented May 29, 1990 and July 23, 1990, 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 wi 11 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 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.

2.

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

DPR-74 is hereby amended to read as follows:

Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 134

, are hereby incorporated in the license.

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

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION Robert Pierson, Director Project Directorate III-1 Division of Reactor Projects - III, IV, V & Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 27, 1990

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO.

TO 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 amendment number and contain marginal lines indicating the area of change.

REMOVE 3/4 1-1 3/4 1-2 3/4 1-3 3/4 1-3a 3/4 1-3b 3/4 2-2 3/4 3-11 3/4 5-5 B3/4 1-1 INSERT 3/4 1-1 3/4 1-2 3/4 1-3 3/4 1-3a 3/4 1-3b 3/4 2-2 3/4 3-11 3/4 5-5 B3/4 1"1

3 4.1 REACTIVITY CONTROL SYSTEMS 3 4.1.1 BORATION CONTROL SHUTDOWN MARGIN -

TAVG GREATER THAN 200 F

LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% Delta k/k.

APPLICABILITY:

MODES 1, 2*, 3, and 4.

ACTION:

With the SHUTDOWN MARGIN less than 1.6% 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 MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% Delta k/k:

& ~

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

If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall 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 Keff greater than or equal to 1.0, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is within the limits of Specification 3.1.3.5.

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

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.5.

See Special Test Exception 3.10.1

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.COOK NUCLEAR PIANT - UNIT 1 3/4 1-1 AMENDMENT N0.7A~VH.

148

REACTIVITY CONTROL SYSTEHS SURVEILLANCE RE UIREMENTS Continued e.

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

1.

Reactor coolant system boron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within plus nr minus 1% Delta k/k at least once per 31 Effective Full Power Days (EFPD),

This comparison shall consider at least those factors stated in Specification 4.1.l.l.l.e, above.

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

COOK NUCLEAR PIANT - UNIT 1 3/4 1-2 mmmm NO.728 148

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REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN -

TAVG LESS THAN OR E UAL TO 200 F

LIMITING CONDITION FOR OPERATION 3.1.1.2 The SHUTDOWN MARGIN shall be greater than or equal to 1.0% Delta k/k.

APPLICABILITY:

MODE 5.

ACTION:

I With the SHUTDOWN MARGIN less than 1.0% 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 MARGIN is restored.

SURVEILIANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% Delta k/k:

l a.

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

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

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1.

Reactor coolant system boron concentration, 2.

Control rod position, 3.

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

COOK NUCLEAR PIANT - UNIT 1 3/4 1-3 AMENDMENT NO

728, 148

This page intentionally left blank.

COOK NUCLEAR PIANT - UNIT l 3/4 1-3a AMENDMENT NO. "2~~ 148

This page intentionally left blank.

COOK NUCLEAR PLANT - UNIT 1 3/4 l-3b mENDXENT NO.728~

148

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Cont'nued b.

TH~ POWER shall not be increased above 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER unless the indicated AFD is within the target band and ACTION 2.a) 1), above has been satisfied.

C.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD has not been outside of the target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE'E UIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

l.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and b.

2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status if the AFD has been outside of the target band for any period of time in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of operation.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

COOK NUCLEAR PEANT - UNIT 1 3/4 2-2 AMENDMENT NO 6L.728 148

S

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued d.

At least once per 18 months by:

1.

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

2.

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

screens, etc.)

show no evidence of structural distress or abnormal corrosion.

e.

At least once per 18 months, during shutdown, by:

1.

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

2.

Verifying that each of the following pumps start automatically upon receipt of a safety injection signal:

a)

Centrifugal charging pump b)

Safety injection pump c)

Residual heat removal pump f.

By verifying that each of the following pumps develops the indicated discharge pressure on recirculation flow when tested pursuant to Specification 4.0,5 at least once per 31 days on a STAGGERED TEST BASIS.

1.

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

Safety Injection pump Greater than or equal to 1409 psig 3.

Residual heat removal pump Greater than or equal to 190 psig g.

By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:

1.

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

COOK NUCLEAR PIANT - UNIT 1 3/4 5-5 AMENDMENT NO. A)7,)Pg Q$4, 148

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

RESPONSE

TIME 12.

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

< 1.0 seconds 13.

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

Steam Generator Water Level--Low-Low

< 1.0 seconds

< 1.5 seconds 15.

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

Undervoltage-Reactor Coolant Pumps 17.

Underfrequency-Reactor Coolant Pumps 18.

Turbine Trip A.

,Low Fluid Oil Pressure B.

Turbine Stop Valve 19.

Safety In]ection Input from ESF NOT APPLICABLE

< 1.2 seconds

< 0.6 seconds NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE 20.

Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE COOK NUCLEAR PLANT - UNIT 1 3/4 3-11 AMENDMENT NO. 148

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3 4.1 REACTIVITY CONTROL SYSTEMS BASES 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 from all operating conditions,

2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be 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 Poled operating avR ave 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 MARGIN of 1.6% Delta k/k is initially required to control the reactivity transient and automatic ESF is assumed to be available.

With Tavg less than 200 F, the reactivity transients resulting from a postulated 0

steam line break cooldown are minimal and a 1% Delta k/k SHUTDOWN MARGIN provides adequate protection for this event.

The SHUTDOWN MARGIN requirements are based upon the limiting conditions 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 will be gradual during boron concentration reductions in the Reactor Coolant System.

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

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

3 4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC 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 of the MTC at the beginning, and near the end of each fuel cycle is adequate to confirm the MTC value since this coefficient changes slowly due principally to the reduction in RCS boron COOK NUCLEAR PLANT - UNIT 1 B 3/4 1-1 AMENDMENT NO. 74,788,148

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ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT N0.134 TO FACILITY OPERATING LICENSE NO.

DPR-74 DOCKET NO. 50-316 Revise Appendix A Technical Specifications by removing the page~ 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 2"2 2-5 through 2-9 3/4 1-1 through 3/4 1-3b 3/4 1-16 3/4 1-23 3/4 2-2 3/4 2-15 through 3/4 2-19 3/4 3-9 through 3/4 3-10 3/4 3-20 through 3/4 3-21 3/4 3-23 through 3/4 3-26 3/4 3-28 3/4 3-30 through 3/4 3-33 3/4 4-6 3/4 5"1 3/4 5-5 th~ough 3/4 5-6 8 2-1 8 2-4 through 8 2-5 8 2-7 8 3/4 1-1 8 3/4 1-3 8 3/4 2-4 and 8 3/4 2-5 8 3/4 7-1 INSERT 2-2 2-5 through 2-9 3/4 l-l through 3/4 1-3b 3/4 1"16 3/4 1-23 3/4 2-2 3/4 2"15 through 3/4 2-19 3/4 3-9 through 3/4 3"10 3/4 3-20 through 3/4 3-21 3/4 3-23 through 3/4 3-26 3/4 3-28 3/4 3-30 through 3/4 3"33 3/4 4-6 3/4 5-1 3/4 5-5 through 3/4 5-6 8 2"1 8 2-4 through 8 2-5 8 3/4 2"7 8 3/4 1-1 8 3/4 1"3 8 3/4 1"4 8 3/4 2-4 through 8 3/4 2-6 8 3/4 7-1

DESIGiV FLOW - 91,600 GPM/LOOP DESCRIPTION OF SAFETY LIMITS Pressure

~(a'a>

Power

~fr ac)

Tavg

(

F)

Power

~(fcac Tavg

~F Power

~fcac)

TcLVg

~F Power

~fzac)

Tavg

(

F) 0.00 615.4 0.98 583.8 1.02 580.9 1.2 55S.'000 0.00 631.S 0.86 605. 8

0. 96 597. 5

1. 2 568. 5 2100 0.00 639.1 0.82 614.0 0.96 601.6 1.2 573.1 660 2250 2400 0.00 0.00 649.2 659.0 0.72 0'. 62 628.6 0.98 642.0 1.1 605.2 599.0 1.2 580.4 588.1 650 640-630 620 2400 PSIA 2250 PSIA 2100 PSIA 610 Pl Ch 600 590 2000 PSIA 1775 PSIA 580 570 560

-'50 0.2 Oa4 0.6 0.8 FRACTION OF RATED THERMAL POWER Figure 2. 1-1 Reactor Core Safety Limits Four Loops in Operation COOK NUCLEAR PLANT - UNIT 2 2-2 AMENDMENT NO. H2>787i

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TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT

1. Manual Reactor Trip Not Applicable 2.

Power Range, Neutron Low Setpoint

- Less than Flux or equal to 25% of RATED THERMAL POWER ALLOWABLE VALUES Not Applicable Low Setpoint

- Less than or equal to 26% of RATED THERMAL POWER High Setpoint

- Less than or equal to 109% of RATED THERMAL POWER High Setpoint

- Less than or equal to 110% of RATED THERMAL POWER 3.

Power Range, Neutron Less than or equal to 5% of Flux, High Positive RATED THERMAL POWER with a Rate time constant greater than or equal to 2 seconds 4.

Power Range, Neutron Less than or equal to 5% of Flux, High Negative RATED THERMAL POWER with a Rate time constant greater than or equal to 2 seconds

5. Intermediate

Range, Less than or equal to 25%

Neutron Flux of RATED THERMAL POWER Less than or equal to 5.5%

of RATED THERMAL POWER with a time constant greater than or equal to 2 seconds I

Less than or equal to 5.5%

of RATED THERMAL POWER with a time constant greater than or equal to 2 seconds Less than or equal to 30%

of RATED THERMAL POWER 6.

Source

Range, Neutron Flux Less than or equal to 10 5 counts per second Le~s than or equal to 1.3 x 10 counts per second

.7. Overtemperature See Note 1

Delta T See Note 3

8. Overpower Delta T See Note 2

See Note 4

9. Pressurizer Pressure

-- Low 10.Pressurizer Pressure

-- High 11.Pressurizer Water Level -- High Greater than or equal to 1950 psig Less than or equal to 2385 psig Less than or equal to 92%

of instrument span Greater than or equal to 1940 psig Less than or equal to 2395 psig Less than or equal to 93%

of instrument span 12.Loss of Flow Design flow is 91 Greater than or equal to 90% of design flow per loop*

1

,600 gpm per loop.

Greater than or equal to 89.1% of design flow per loop*

COOK NUCLEAR PLANT - UNIT 2 2-5 AMENDMENT NO.ggp"~4

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TABLE 2.2-1 Continued

,REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 13.Steam Generator Water Level-Low-Low 14.Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level Greater than or equal to 218 of narrow range instrument span

- each steam generator Less than or equal to 1.47 x 10 lbs/hr of steam flow at RATED THERMAL POWER coincident with steam generator water level greater than or equal to 25% of narrow range instrument span

- each steam generator Greater than or equal to 19.2% of narrow range instrument span

- each steam generator Less6than or equal to 1.56 I

x 10 lbs/hr of steam flow at RATED THERMAL POWER coincident with steam generator water level greater than or equal to 24% of narrow range instrument span

- each steam generator 15.Undervoltage-Reactor Coolant Pumps Greater than or equal to 2905 volts - each bus Greater than or equal to I

2870 volts - each bus 16.Underfrequency-Reactor Coolant Pumps Greater than or equal to 57.5 Hz - each bus Greater than or equal to 57.4 Hz - each bus 17.Turbine Trip A. Low Trip System Pressure B. Turbine Stop Valve Closure Greater than or equal 58 psig Greater than or equal 1% open to to Greater than or equal to I

57 psig'reater than or equal to 1% open 18.Safety Infection Not Applicable Input from ESF Not Applicable 19.Reactor Coolant Pump Not Applicable Breaker Position Trip Not Applicable COOK NUCLEAR PIANT - UNIT 2 2-6 AMENDMENT NO, 82~

134'.

TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Note 1:

OvertemPerature 5T < AT

[Kl-K2[(1 + TIS)/(1 + v2S) ] (T T )+K3(P P ) fl(bl)]

Where:

bT0

- Indicated AT at RATED THERMAL POWER

- Average temperature, F

0

- Indicated,T at RATED THERMAL POWER less than or equal to 576.O 'F Pl 1+ v S

1+

r2S 1

1'

- Pressurizer

Pressure, psig 2235 psig (indicated RCS nominal operating pressure)

- The function generated by the lead-lag controller for T

dynamic compensation avg Time constants utilized in the lead-lag controller for T

vl - 28 secs, w2 4 secs.

avg'

- Laplace transform operator COOK NUCLEAR PLANT - UNIT 2 2-7 AMENDMENT NO. N [~4

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TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION Continued 4 Loo s in eration Kl - 1.09 K2 0.01331 K3 0.00058 and fl(AI) 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

-33 percent and +6 percent, f (2)-0 (where q

and qb are percent RATED THERMAL POWER in the top and bottom halves b

1 of the core respectively, and q

+ qb is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (q

- qb) exceeds

-33 percent, the hT trip setpoint shall be automatically reduced by 3.5 percent of its value at RATED THERMAL POWER.

(iii)

For each percent that the magnitude of (q

- qb) exceeds

+6 percent, the hT trip setpoint shall be automatically t

reduced by 1.0 percent of its value at RATED THERMAL POWER.

COOK NUCLEAR PLANT - UNIT 2 2-8 AMENDMENT NO.H2. 134

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TABLE 2.2-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATIONS Continued Note 2:

Overpower bT < hT

[K4-K5[T3S/(1+T3S)]T-K6[T-T"]-f2(~I)]

Where:

AT0 K4

- Indicated hT at rated pover 0

- Average temperature, F

Indicated T

'at RATED THERMAL POWER less, than or avg equal to 576.0 F

1.08 0.02/

F for increasing average temperature and 0 for 0

decreasing average temperature K6

- 0.00197 for T greater than T"; K6 0 for T less than or equal to T" Q3S/ ( 1+F3 S )- The function generated by the rate 1ag contro 1 1er for T

dynamic compensation avg

- Time constant utilized in the rate lag controller for T

<3 10 secs.

avg 3

S

- Laplace transform operator

~ f2(gl)

- 0.0 Note 3:

The channel's maximum trip point shall not exceed its computed trip point by more than 1.3 percent AT span.

I Note 4:

The channel's maximum trip point shall not exceed its computed trip point by more than 3.0 percent AT span.

COOK NUCLEAR PLANT - UNIT 2 2-9 AMENDMENT NO. 88, 134

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3 4.1 REACTIVITY CONTROL SYSTEMS 3 4.1.1 BORATION CONTROL SHUTDOWN MARGIN - T GREATER THAN 200 F

LIMITING CONDITION FOR'PERATION 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.6% Delta k/k.

t APPLICABILITY:

MODES 1, 2*, 3, and 4.

ACTION'ith the SHUTDOWN MARGIN less than 1.6% 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 MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1

~ 1.1.1 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.6% Delta k/k:

a.

Within one hour after detection of an inoperable control rod(s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod(s) is inoperable. If the inoperable control rod is immovable or untrippable, the above required SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or untrippable control rod(s).

b.

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

When in MODE 2 with K f less than 1.0, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to I

achieving reactor criffcality by verifying that the predicted critical control rod position is within the limits of Specification 3.1.3.6.

d.

Prior to initial operation above 5% RATED THERMAL POWER after each fuel loading, by consideration of the factors of e below, with the control banks at the maximum insertion limit of Specification 3.1.3.6.

See Special Test Exception 3.10.1 COOK NUCLEAR PLANT - UNIT 2 3/4 l-l AMENDMENT NO. 82 ~88

~34 1

REACTIVITY CONTROL SYSTEMS SURVEILLANCE RE UI1UMENTS Continued e.

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

1.

Reactor coolant system boron concentration, 2.

Control rod position, 3.

Reactor co'olant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5.

Xenon concentration, and 6.

Samarium concentration.

4.1.1.1.2 The overall core reactivity balance shall be compared to predicted values to demonstrate agreement within plus or minus 1% Delta k/k at least I

once per 31 Effective Full Power Days (EFPD).

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

The predicted reactivity values shall be ad]usted (normalized) to correspond to the actual core conditions prior to exceeding a fuel burnup of 60 Effective Full Power Days after each fuel loading, COOK NUCLEAR PLANT - UNIT 2 3/4 1-2 AMENDMENT NO. HR,KSS P4

REACTIVITY CONTROL SYSTEMS SHUTDOWN MARGIN - T LESS THAN OR E UAL TO 200 F

LIMITING CONDITION FOR OPERATION 3.1. 1.2 The SHUTDOWN HARGTN shall be greater than or equal to 1.0e Delta k/k.

L APPLICABILITY: MODE 5.

ACTION:

With the SHUTDOWN MARGIN less than 1.0% 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 MARGIN is restored.

SURVEILLANCE RE UIREMENTS 4.1.1.2 The SHUTDOWN MARGIN shall be determined to be greater than or equal to 1.0% Delta k/k:

a.

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

If the inoperable control rod is immovable or

'ntrippable, the SHUTDOWN MARGIN shall be verified acceptable with an increased allowance for the withdrawn worth of the immovable or, untrippable control rod(s).

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration of the following factors:

1.

Reactor coolant system boron concentration, 2.

Control rod position, 3,

Reactor coolant system average temperature, 4.

Fuel burnup based on gross thermal energy generation, 5 ~

Xenon concentration, and 6.

Samarium concentration.

COOK NUCLEAR PIANT - UNIT 2 3/4 1-3 AMENDMENT NO. 8~S "+2 "~~3

'134

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COOK NUCLEAR PLANT - UNIT 2 3/4 1-3a mENDZENT NO. 82~

134

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COOK NUCLEAR PLANT - UNIT 2 3/4 1-3b

REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCES

- OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.8 Each of the following borated water sources shall be OPERABLE:

a.

A boric acid storage system and associated heat tracing with:

1.

A minimum contained borated water volume of 7715 gallons, 2.

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

A minimum solution temperature of 145 F.

0 b.

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 80 F.

0 APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

a.

With the boric acid storage system inoperable, restore the storage system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and 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 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

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

SURVEILLANCE RE UIREMENTS 4.1.2.8 Each borated water source shall be demonstrated OPERABLE:

COOK NUCLEAR PLANT - UNIT 2 3/4 1-16 AMENDMENT NO. SPY 1 34

0

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REACTIVITY CONTROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position (specified in the COLR) shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a.

T greater than or equal to 541 F, and 0

avg F

b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 AND 2 ACTION:

Pith the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

SURVEILLANCE RE UIREMENTS 4.1.3.4 through a.

The rod drop time of full length rods shall be demonstrated measurement prior to entering MODE 2:

For all rods following each removal of the reactor vessel

head, b.

For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those specific rods, and C ~

At least once per 18 months,'OOK NUCLEAR PLANT - UNIT 2 3/4 1-23 AMENDMENT NO. 82, N7,122, 134

POWER DISTRIBUTION LIMITS ACTION:

(Centdnned) b.

THERMAL POWER shall not be increased above 90% or 0.9 x APL (whichever is less) of RATED THERMAL POWER unless the indicated AFD is within the target band and ACTION 2.a) 1), above has been satisfied.

THERMAL POWER shall not be increased above 50% of RATED TH~

POWER unless the indicated AFD has not been outside of the target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

SURVEILLANCE RE UIREMENTS 4.2.1.1 The indicated AXIALFLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a.

b.

Monitoring the indicated AFD for each OPERABLE excore channel:

l.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after restoring the AFD Monitor Alarm to OPERABLE status if the AFD has been outside of the target band for any period of time in the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of operation.

Monitoring and logging the indicated AXIALFLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable, The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

COOK NUCLEAR PIANT - UNIT 2 3/4 2-2 AMENDMENT NO. Hg~ P4

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POWER DISTRIBUTION LIMITS DNB AND Tav OPERATING PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the following operational indicated limits:

a.

DNB

1. Reactor Coolant System T

2. Pressurizer Pressure avg

3. Reactor Coolant System Total Flow Rate Less than or equal to 578.7 W Greater than or equal to 2200 psig*/**

Greater than or equal to 366,400 gpm***

b.

Tavg

1. Reactor Coolant System Tavg APPLICABILITY:

MODE 1 Greater than or equal to 543.9 W ACTION'ith any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5%

of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

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

4.2.5.2 The indicators used to determine RCS total flow shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.5.3 The RCS total, flow rate shall be determined by a power balance around the steam generators at least once per 18 months.

4.2.5.4 The provisions of Specification 4.0.4 shall not apply to primary flow surveillances.

Indicated average of at least three OPERABLE instrument loops.

- Limit not applicable during either a THERMAL POWER ramp in excess of 5%

of RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RTP

- - Indicated value COOK NUCLEAR PLANT - UNIT 2 3/4 2-15 AMENDMENT NO.88,134

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COOK NUCLEAR PLANT - UNIT 2 3/4 2-16

~NENT NO. 82,787. 13<

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COOK NUCLEAR PLANT - UNIT 2 3/4 2-17 mENDmrr NO, 88,134

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COOK NUCLEAR PLANT - UNIT 2 3/4 2-18 AMENDMENT N0.82, KS7, 134

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POWER DISTRIBUTION LIMITS ALLOWABLE POWER LEVEL - APL LIMITING CONDITION FOR OPERATION 3.2.6 THERMAL POWER shall be less than or equal to ALLOWABLE POWER LEVEL (APL), given by the following relationshios:

F (Z)xV(Z)xF P

o CFQ is the F limit at RATED THERMAL POWER specified in the COLR for Westinghlusc or Exxon fuel.

o F (Z) is the measured hot channel factor, including a 3%

m~nufacturing tolerance uncertainty and a 54 mcasurcment uncertainty.

o V(Z) is the function specified in the COLR.

~

F 1.00 except when successive steady-state power distribution mRps indicate an increase in max over Z.of

~F Z with exposure.

K7Z)

Then either of the following penalties, F

, shall be taken:

P F

1.02 or, P

F 1.00 provided that Surveillance Requirement 4.2.6.2 is sRtisfied once per 7 Effective Full'ower Days until 2 successive maps indicate that the'ax over Z of

~F Z

is not increasing.

QKZ)

~

The above limit is not applicable in the following core regions.

1)

Lower core region Oi to 10i inclusive.

2)

Upper core region 90% to 100% inclusive.

COOK NUCLEAR PLANT - UNIT 2 3/4 2-19 AMENDMENT NO. SX,f87, $2g, 134

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TABLE 3,3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES FUNCTIONAL UNIT 1.

Manual Reactor Trip 2.

Power Range, Neutron Flux

RESPONSE

TIME NOT APPLICABLE Less than o1 equal to 0.5 seconds*

3.

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

Power Range, Neutron Flux High Negative Rate Less than or equal to 0.5 seconds*

5.'ntermediate

Range, Neutron Flux 6.

Source

Range, Neutron Flux 7.

Overtemperature Delta T NOT APPLICABLE NOT APPLICABLE Less than or equal to 6.0 seconds*

8.

Overpower Delta T 9.

Pressurizer Pressure--Low NOT APPLICABLE Less than or equal seconds 10 'ressurizer Pressure--High Less than or equal to 2.0 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 PIANT - UNIT 2 3/4 3-9 AMENDMENT NO@g, 7gf, f 3Q

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

RESPONSE

TIME 12.Loss of Flow - Single Loop (Above P-8) 13.Loss of Flow - Two Loops (Above P-7 and below P-8)

Less than or equal to 1.0 seconds Less than or equal to 1.0 secollds 14.Steam Generator Water Level--Low-Low Less than or equal to 2.0 seconds 15.Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level NOT APPLICABLE 16.Undervoltage-Reactor Coolant Pumps 17.Underfrequency-Reactor Coolant Pumps 18.Turbine Trip A. Low Fluid Oil Pressure B. Turbine Stop Valve 19.Safety In]ection Input from ESF Less than or equal to 1.5 seconds Less than or equal to 0.6 seconds NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE 20.Reactor Coolant Pump Breaker Position Trip NOT APPLICABLE COOK NUCLEAR PLANT - UNIT 2 3/4 3-10 AMENDMENT NO.

1 34

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.TABLE 3.3-3 Continued ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I

C M

FUNCTIONAL UNIT Three Loops Operating TOTAL NO.

OF CHANNELS 1 T /

operafing loop CHANNELS TO TRIP 1¹¹¹ T

in avg anygoper-ating loop MINIMUM CHANNELS OPERABLE 1 T in an/'wo operating loops APPLICABLE MODES 3¹¹ ACTION 15 e.

Steam Line Pressure-Low I

hJo Four Loops Operating Three Loops Operating 5.

TURBINE TRIP 6 FEEDWATER ISOLATION a.

Steam Generator Water Level--High-High 1 pressure/

loop 1 pressure/

operating loop 3/loop 2 press-ures any loops 1¹¹¹ pressure in any operating loop 2/loop in any oper-ating loop 1 press-ure any 3 loops 1 press-ure in any 2 operating loops 2/loop in each oper-ating loop 1,2,3¹¹ 3¹¹ 1,2 and 3

14*

15 14*

0

TABLE 3.3-3 Continued TABLE NOTATION

¹Trip function may be bypassed in this MODE belo~ P-ll.

¹¹Trip function may be bypassed in this MODE below P-12.

¹¹¹The channel(s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped mode.

¹¹¹¹Manually trip all bistables which would be automatically tripped in the event pressure in the associated active loop were less than the pressure in the inactive loop.

For example, if loop 1 is the inactive loop then the bistables which indicate low pressure in loops 2, 3, and 4 relative to loop 1 should be tripped.

The provisions of Specification 3.0.4 are not applicable.

COOK NUCLEAR PIANT - UNIT 2 3/4 3-21 AMENDMENT NO. 88, 134

TABLE 3.3-4 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINTS ALLOWABLE VALUES 1.

SAFETY INJECTION, TURBINE

TRIP, FEEDWATER ISOLATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS

a. Manual Initiation

b. Automatic Actuation Logic

c. Containment Pressure--

High

d. Pressurizer Pressure--

Low

e. Differential Pressure Between Steam Lines--

High Not Applicable Not Applicable Less than or equal to 1.1 psig Greater than or equal to 1900 psig Less than or equal to 100 psi Not Applicable Not Applicable Less than or equal to 1.2 psig Greater than or equal to 1890 psig Less than or equal to 112 psi

f. Steam Line Pressure--

Low Gaeater than or equal Greater than or equal to 600 psig steam line to 585 psig steam pressure line pressure COOK NUCLEAR PLANT - UNIT 2 3/4 3-23 AMENDMENT NO. E, IE,28.

134

TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINTS ALLOWABLE VALUES 2.

CONTAINMENT SPRAY

a. Manual Initiation

b. Automatic Actuation Logic Not Applicable Not Applicable Not Applicable Not Applicable

c. Containment Pressure--

High-High 3.

CONTAINMENT ISOIATION a.

Phase "A" Isolation Less than or equal to Less than or equal to

,2.9 psig 3.0 psig

1. Manual 2.

From Safety In)ection Automatic Actuation Logic b.

Phase "B" Isolation Not Applicable Not Applicable Not Applicable Not Applicable

1. Manual

2. Automatic Actuation Logic

3. Containment Pressure-High-High c.

Purge and Exhaust Isolation

1. Manual

2. Containment Radio-activity--High Train A (VRS 2101 i ERS 2301 i ERS - 2305)

3. Containment Radio-activity--High Train B (VRS-2201, ERS-2401, ERS-2405)

Not Applicable Not Applicable Less than or equal to 2.9 psig Not Applicable See Table 3.3-6 See Table 3.3-6 Not Applicable Not Applicable Less than or equal to 3.0 psig Not Applicable Not Applicable Not Applicable COOK NUCLEAR PLANT - UNIT 2 3/4 3-24 AMENDMENT N0.88, 134

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TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINTS ALLOWABLE VALUES 4.

STEAM LINE ISOLATION

a. Manual

b. Automatic Actuation Logic

c. Containment Pressure--

High-High Not Applicable Not Applicable Less than or equal to 2.9 psig Not Applicable Not Applicable Less than or equal to 3.0 psig d.

Steam Flow in Two Steam Lines--High Coincident with T

--Low-Low avg Less than or equal to a function defined as follows:

A Delta-p correspogding to 1.6 x 10 lbs/hr steam flow between 0%

and 20% load and then a Delta-p increasing linearly to a Delta-p correspogding to 4.5 x 10 lbs/hr at full load.

Less than or equal to a function defined as follows:

A Delta-p corresponfing to 1 ~ 75 x 10 lbs/hr steam flow between 0%

and 20% load and then a Delta-p increasing linearly to a Delta-p correspon)ing to 4.55 x 10 lbs/hr at at full load.

e.

Steam Line Pressure--Low 5.

TURBINE TRIP AND FEE5WATER ISOIATION T

greater than or equal to 541 F

Greater than or equal to 600 psig steam line pressure Tav greater than or equal to 539 F

Greater than or equal to 585 psig steam line pressure a.

Steam Generator Water Level--High-High Less than or equal to 67% of narrow range instrument span each steam generator Less than or equal to 68% of narrow range instrument span each steam generator COOK NUCLEAR PLANT - UNIT 2 3/4 3-25 AMENDMENT NO. 88 788, 134

TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES 6.

MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.

Steam Generator Water Level--Low-Low b.

4 kV Bus Loss of Voltage

c. Safety In)ection

d. Loss of Main Feedwater Pumps Greater than or equal to 21% of narrow range instrument span each steam generator 3196 volts with a 2 second delay Not Applicable Not Applicable Greater than or equal to 19.2% of narrow range instrument span each steam generator 3280 + 120 volts with a 2 + - 0.2 second delay Not Applicable Not Applicable 7.

TURBINE DRIVEN AUXILIATY FEEDWATER PUMPS a.

Steam Generator Water Level<<-Low-Low Greater than or equal to 21% of narrow range instrument span each steam generator Greater than or equal to 19.2% of narrow range instrument span each steam generator

b. Reactor Coolant Pump Bus Undervoltage Greater than or equal Greener chan or equal to 2750 Volts--each bus 2725 Volts--each bus 8.

LOSS OF POWER a.

4 kV Bus Loss of Voltage b.

4 kV Bus Degraded Voltage 3196 volts with a 2 second delay 3596 volts with a 2.0 minute time delay 3280 + 120 volts with a 2 + - 0.2 second delay 3638 + 60 volts with a 2.0 minute + -

6 second time delay COOK NUCLEAR PLANT - UNIT 2 3/4 3-25a AMENDMENT NO.

82~ 72Ar 134

TABLE 3.3<<5 ENGINEERED SAFETY FEATURES

RESPONSE

TIMES INITIATING SIGNAL AND FUNCTION

1. Manual

RESPONSE

TIME IN SECONDS Safety Injection (ECCS)

Feedwater Isolation Reactor Trip (SI)

Containment Isolation-Phase "A"

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

Containment Spray Containment Isolation-Phase "B"

Containment Purge and Exhaust Isolation c.

Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation d.

Steam Line Isolation 2.

Containment Pressure-Hi h

Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable 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 27.0*

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 134

.0 l

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

RESPONSE

TIME IN SECONDS 6.

Steam Line Pressure--Low a

~

b.

Ce d.

gh.

Safety In)ection (ECCS)

Reactor Trip (from SI)

Feedwater Isolation.

Containment Isolation-Phase "A"

Containment Purge and Exhaust Isolation Motor Driven Auxiliary Feedwater Pumps Essential Service Water System Steam Line Isolation Less than or equal Less than or equal Less than or equal Less than or equal to 12.0¹/24.0¹¹ to 2.0 to 8.0 to 18.0¹/28.0¹¹ Not Applicable Less than or equal to 60.0 Less than or equal to 14.0¹/48.0¹¹ Less than or equal to 8.0 7.

Containment Pressure--Hi h-Hi h a.

b.

C.

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 7.0 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.

b.

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

4160 volt Emer enc Bus Loss of Volta e a.

Motor Driven Auxiliary Feedwater Pumps

11. Loss of Main Feedwater Pum s Less than or equal to 60,0 a.

Motor Driven Auxiliary Feedwater Pumps

12. Reactor Coolant Pum Bus Undervolta e

Less than or equal to 60.0 a.

Turbine Driven Auxiliary Feedwater Pump Less than or equal to 60.0 COOK NUCLEAR PLANT - UNIT 2 3/4 3-28 AMENDMENT NO, )IIl,7/7, 134

TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIMMENTS FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION CHANNEL MODES IN WHICH FUNCTIONAL SURVEIL1ANCE 1.

SAFETY INJECTION, TURBINE TRIP, FEEDWATER ISOIATION, AND MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.

Manual Initiation

b. Automatic Actuation Logic

c. Containment Press-ure-High

d. Pressurizer Press-ure-Low

e. Differential Press-ure Between Steam Lines--High

f. Steam Line Pressure--

Low N.A.

N.A.

R R

M(1)

M(2)

M(3)

M M

1,2,3,4 1,2,3,4 1,2,3 1,2,3 1,2,3 1,2,3 2.

CONTAINMENT SPRAY a.

Manual Initiation

b. Automatic Actuation Logic

c. Containment Press-ure-High-High 3.

CONTAINMENT ISOLATION P

a.

Phase "A" Isolation

1) Manual

2) From Safety Injection Automatic Actuation Logic b.

Phase "B" Isolation

1) Manual

2) Automatic Actua-tion Logic

3) Containment Press-ure-High-High COOK NUCLEAR PLANT - UNIT 2 N.A.

N.A.

3/4 3-30 M(1)

M(2)

M(3)

M(1)

M(2)

M(1)

M(2)

M(3) 1,2,3,4 1,2,3,4 1,2,3 1,2,3,4 1,2,3,4 1,2,3,4 1,2,3,4 1,2,3 AMENDMENT NO. 84,134

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TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILIANCE RE UIREMENTS FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION CHANNEL MODES IN WHICH FUNCTIONAL SURVEILLANCE

c. Purge and Exhaust Isolation

1) Manual N.A.

2) Containment Radio -

S activity-High 4.

STEAM LINE ISOLATION N.A.

R M(1) 1,2,3,4 1,2,3,4

a. Manual

b. Automatic Actuation Logic

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

Steam Flow in Two Steam Lines--

High Coincident with Tavg--Low-Low e.

Steam Line Pressure--

Low 5.

TURBINE TRIP AND FEEDWATER ISOLATION a.

Steam Generator Water Level- -High-High 6.

MOTOR DRIVEN AUXILIARY FEEDWATER PUMPS a.

Steam Generator Water Level--Low-Low

b. 4 kV Bus Loss of Voltage

c. Safety In)ecti,on

d. Loss of Main Feed Pumps N.A.

N.A.

N.A, R

R R

N.A.

M(1)

M(2)

M(3)

M M(2)

R 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2,3 1,2 COOK NUCLEAR PLANT - UNIT 2 3/4 3-31 AMENDMENT N0.88 97 187 134

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I TABLE 4.3-2 Continued ENGINEERED SAFETY FEATURED ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION CHANNEL MODES IN WHICH FUNCTIONAL SURVEILLANCE 7.

TURBINE DRIVEN AUXILIARYFEEDWATER PUMP a.

Steam Generator Water Level--Low-low b.

Reactor Coolant Pump Bus Undervoltage 8.

LOSS OF POWER N.A.

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

4 kv Bus Loss of Voltage b.

4 kv Bus Degraded Voltage R

1,2,3,4 1,2,3,4 COOK NUCLEAR PLANT - UNIT 2 3/4 3-32 AMENDMENT NO. H2 >7 134 M

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COOK NUCLEAR PIANT - UNIT 2 3/4 3-32a AHENDMENT NO. ~"

"~4

h TABLE 4.3-2 Continued TABLE NOTATION Manual actuation switches shall be tested at least once per 18 months during shutdown.

All other circuitry associated with manual safe-guards actuation shall receive a

CHANNEL FUNCTIONAL TEST at least once per 31 days.

1 (2)

Each train or logic channel shall be tested at least every other 31 days.

(3)

The CHANNEL FUNCTIONAL TEST shall include exercising the transmitter by applying either a vacuum or pressure to the appropriate side of the transmitter.

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

82> 334

REACTOR COOIANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a water volume.less than or equal to 92% of span and at least 150 kW of pressurizer heaters.

APPLICABILITY:

MODES 1, 2, and 3.

ACTION:

With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer

heaters, either restore the inoperable emergency power supply within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With the pressurizer otherwise inoperable, be in at least HOT SHUTDOWN with the reactor trip breakers open within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE RE UIREMENTS 4.4.4.1 The pressurizer water volume shall be determined to be within its limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.4.2 The emergency power supply for the pressurizer heaters shall be demonstrated OPERABLE at least once per 18 months by transferring power from the normal to the emergency power supply and energizing the required capacity of heaters.

COOK NUCLEAR PLANT - UNIT 2 3/4 4-6 AMENDMENT NO. 8),

134

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3 4.5 EMERGENCY CORE COOLING SYSTEMS ECCS ACCUMUIATORS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system accumulator shall be OPERABLE with:

a.

The isolation valve open, b.

A contained borated water volume of between 921 and 971 cubic feet, c.

A boron concentration between 2400 ppm and 2600 ppm, and d.

A nitrogen cover-pressure of between 585 and 658 psig.

APPLICABILITY:

MODES 1, 2, and 3.*

ACTION:

a.

With one accumulator inoperable, except as a result of a closed isolation valve, restore the inoperable accumulator to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

With one accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILIANCE RE UIREMENTS 4.5.1 Each accumulator shall be demonstrated OPERABLE:

a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1.

Verifying the contained borated water volume and nitrogen cover-pressure in the tanks, and 2.

Verifying that each accumulator isolation valve is open.

Pressurizer Pressure above 1000 psig.

COOK NUCLEAR PLANT - UNIT 2 3/4 5-1 AMENDMENT NO. 9~ 9<

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EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued At least once per 18 months by:

1.

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

2.

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

screens, etc.)

show no evidence of structural distress or corrosion.

e.

At least once per 18 months, during shutdown, by:

1.

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

2.

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

a)

Centrifugal charging pump b)

Safety injection pump c)

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

1.

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

Safety Injection pump Greater than or equal to 1445 psig 3.

Residual heat removal pump Greater than or equal to 195 psig g.

By verifying the correct position of each mechanical stop for the following Emergency Core Cooling System throttle valves:

1.

Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months following completion of each valve stroking operation or maintenance on the valve when the ECCS sub-systems are required to be OPERABLE'OOK NUCLEAR PLANT - UNIT 2 3/4 5-5 AMENDMENT NO.)7

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE RE UIREMENTS Continued 2.

At least once per 18 months.

Boron Injection Throttle Valves Valve Number

1. 2-SI-141 Ll

2. 2-SI-141 L2 Safety Injection Throttle Valves Valve Number

1. 2-SI-121 N

2. 2-SI-121 S

3. 2-SI-141 L3

4. 2-SI-141 L4 h.

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

Boron Injection System Sin le Pum

Safety Injection System Sin le Pum **

Loop 1 Boron Injection Flow 117.5 gpm Loop 2 Boron Injection Flow 117.5 gpm Loop 3 Boron Injection Flow 117.5 gpm Loop 4 Boron Injection Flow 117.5 gpm Loop 1 and 4 Cold Leg Flow greater than or equal to 300 gpm Loop 2 and 3 Cold Leg Flow greater than or equal to 300 gpm

~Combined Loop 1,2,3 and 4 Cold Leg Flow (single pump) less than or equal to 640 gpm.

Total SIS (single pump) flow, including miniflow, shall not exceed 700 gpm.

The flow rate in each boron injection (BI) line should be adjusted to provide 117.5 gpm (nominal) flow into each loop.

Under these conditions there is zero mini-flow and 80 gpm plus or minus 5 gpm simulated RCP seal injection line flow.

The actual flow in each BI line may deviate from the nominal so long as:

a) thy difference between the highest and lowest flow is 25 gpm or less.

b) the total flow to the four branch lines does not exceed 470 gpm.

c) the minimum flow through the three most conservative (lowest flow) branch lines must not be less than 300 gpm, d) the charging pump discharge resistance (2.31*Pd/Qd"2) must not be less than 4.73E-3 ft/gpm*2 and must not be greater than 9.27E-3 ft/gpm"2, (Pd is the pump discharge pressure at runout; Qd is the total pump flow rate).

COOK NUCLEAR PIANT - UNIT 2 3/4 5-6 AMENDMENT NO. 64, 134

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the WRB-2 correlation and W-3 correlation for conditions outside the range of WRB-2.

The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, and is indicative of the margin to DNB.

The DNB design basis is as follows:

there must be at least a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-2 correlation for Vantage-5 fuel, and the W-3 correlation for ANF fuel and conditions which fall outside the range of applicability of the WRB-2).

The correlation DNBR limits are established based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for WRB-2 and 1.3 for the W-3).

In meeting the DNB design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are statistically combined with the DNBR correlation statistics such that there is at least a 95 percent probability with a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to a calculated design limit DNBR.

The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty.

This DNBR uncertainty, combined with the DNBR correlation statistics, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

For Cook Nuclear Plant Unit 2, the design DNBR values are 1.23 and 1.22 for Vantage-5 fuel typical and thimble cells, respectively, and 1.39 and 1.36 for typical and thimble cells for the ANF fuel.

In addition, margin has been maineained in both fuel types by performing safety analyses to a safety analysis limit DNBR.

The margin between the design and safety analysis limit DNBR is used to offset known DNBR penalties (i.e., transition core penalties, rod bow, etc.)

and provide DNBR'argin for operating and design flexibility.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, R'eactor Coolant System pressure, and average temperature below which the calculated DNBR is no less than the design DNBR limit value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

l COOK NUCLEAR PLANT - UNIT 2 B 2-1 AMENDMENT NO. gg, 134

LIMITING SAFETY SYSTEM SETTINGS BASES The Power Range Negative Rate trip provides protection to ensure that the calculated DNBR is maintained above the design DNBR value for multiple control rod drop accidents.

The analysis of a single control rod drop (or some multiple rod drops) accident indicates a return to full power may be initiated by the automatic control system in response to a continued full power turbine load demand or by the negative moderator temperature feedback.

1 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 of the Power Range, Neutron Flux chgnels.

The Source Range Channels will initiate a reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER 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 erature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, 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 from the core to the loop temperature detectors.

This reactor trip limit is always belo~ the core safety limit as shown in Figure 2.1-1. If axial peaks are more severe 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 PIANT - UNIT 2 B 2-4 AMENDMENT NO. SR, 134

LIMITING SAFETY SYSTEM SETTINGS BASES Over ower Delta T The Overpower Delta T reactor trip pxovides 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.

No credit was taken for operation of this trip in the accident analyses;

however, the functional capability of the Overpower Delta T trip 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 control rod assembly bank withdrawal at power event.

COOK NUCLEAR PLANT - UNIT 2 B 2-5 AMENDMENT NO, 82~

134

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LIMITING SAFETY SYSTEM SETTINGS BASES Undervolta e and Underfre uenc

- Reactor Coolant Pum Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or underfrequency to more than one reactor coolant pump.

The specified set points assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients, For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip set point is reached shall not exceed 0.3 seconds.

The total response times for these functional units include an additional 0.3 seconds for trip breaker operation and CRDM release.

Turbine Tri A Turbine Trip causes a direct reactor trip when operating above P-7.

Each of the turbine trips provide turbine protection and reduce the severity of the ensuing transient.

No credit was taken in the accident analyses for operation of these trips.

Their functional capability at the specified trip settings is required to enhance the overall reliability of the Reactor Protection System.

COOK NUCLEAR PIANT - UNIT 2 B 2-7 AMENDMENT NO. ~~~

134

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3 4.1 REACTIVITY CONTROL SYSTEMS BASES 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 from all operating conditions,

2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be 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 operaHng temperature, and is ave'vR associated with a postulated steam line break accident and resulting uncontrolled RCS cooldown.

In the analysis of this accident, a minimum SHUTDOWN MARGIN of 1.6% 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 reactivity transients resulting from a 0

postulated st5am line break cooldown are minimal and a 1% -Delta k/k SHUTDOWN av MARGIN provides adequate protection for this event.

The SHUTDOWN MARGIN requirements are based upon the limiting conditions 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 will be gradual during boron concentration reductions in the Reactor Coolant System.

A flow rate of at least 2000 GPM will circulate an equivalent Reactor Coolant System volume of 12,612 cubic feet in approximately 30 minutes.

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

COOK NUCLEAR PLANT - UNIT 2 B 3/4 1-1 AMENDMENT NO. 82,18S 134

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3/4.1 REACTIVITY CONTROL SYSTEMS BASES With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable.

Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The limitation for 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, unless the reactor vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability of either system is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after xenon decay and cooldown to 200 F.

The maximum expected boration capability usable volume requirement is 7715 gallons of 20,000 ppm borated water from the boric acid storage tanks or 160,122 gallons of borated water from the refueling water storage tank.

The required RWST volume is based on an assumed boron concentration of 2400 ppm.

The minimum RWST boron concentration required by the post-LOCA long-term cooling analysis is 2400 ppm.

The minimum contained RWST volume is based on ECCS considerations.

See Section B 3/4.5.5.

With the RCS average temperature below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200'F is sufficient to provide the required MODE 5 SHUTDOWN MARGIN after xenon decay and cooldown from 200 F to 140 F.

This condition requires usable volumes of either 4300 gallons of 20,000 ppm borated water from the boric acid storage tanks or 90,000 gallons of borated water from the refueling water storage tank.

The value for the boric acid Storage tank volume includes sufficient boric acid to borate to 2190 ppm.

The required RWST volume is based on an assumed boron concentration of 2400 ppm.

The minimum RWST boron concentration required by the post-LOCA long-term cooling analysis is 2400 ppm.

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 OPERABILITY of boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

COOK NUCLEAR PLANT UNIT 2 B 3/4 1-3 Amendment No.8~N7,G6,

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3/4.1 REACTIVITY CONTROL SYSTEMS BASES With the RCS average temperature above 200 F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems inoperable.

Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from in)ection system failures during the repair period.

The limitation for maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps and safety in)ection

pumps, except the required OPERABLE charging

pump, to be inoperable below 152'F, unless the reactor vessel head is removed, provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

The boration capability of either system is sufficient to provide the required SHUTDOWN MARGIN from expected operating conditions after xenon decay and cooldown to 200 F.

The maximum expected boration capability usable volume requirement is 7715 gallons of 20,000 ppm borated water from the boxic acid storage tanks or 160,122 gallons of borated water from the refueling water storage tank.

The required RWST volume is based on an assumed boron concentration of 2400 ppm.

The minimum RWST boron concentration required by the post-LOCA long-term cooling analysis is 2400 ppm.

The minimum contained RWST volume is based on ECCS considerations.

See Section B 3/4.5.5.

With the RCS average temperature below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single in)ection system becomes inoperable.

The boron capability required below 200 F is sufficient to provide the required MODE 5 SHUTDOWN MARGIN after xenon decay and cooldown from 200 F to 140 F.

This condition requires usable volumes of either 4300 gallons of 20,000 ppm borated water from the boric acid storage tanks or 90,000 gallons of borated water from the refueling water storage tank.

The value for the boric acid storage tank volume includes sufficient boric acid to borate to 2190 ppm.

The required RWST volume is based on an assumed boron concentration of 2400 ppm.

The minimum RWST boron concentration required by the post-LOCA long-term cooling analysis is 2400 ppm.

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 OPERABILITY of boron in)ection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

COOK NUCLEAR PLANT UNIT 2 B 3/4 1-3 Amendment No.g~N7>L76>>

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REACTIVITY CONTROL SYSTEMS BASES BORATION SYSTEMS Continued The charging flowpath of Unit 2 required for Unit 1 shutdown support ensures that flow is available to Unit 1 and addresses the requirements of 10 CFR 5g Appendix R.

The flowpath consists of a charging pump powered from an electrical bus and associated water supplies and delivery system.

Pire watches posted in the affected opposite unit areas (i.e., Unit 1 areas requiring use of the Unit 2 charging system in the event of a fire) may serve as the equivalent shutdown capability specified in the action statements of Specification 3.1.2.3.

In the affected areas, either establish continuous fire watches or verify the OPERA-BILITY of fire detectors per Specification 4.3.3.7 and establish hourly fire watch patrols.

The required opposite unit equipment along with the surveil-lance requirements necessary to ensure that this equipment is capable of fulfillingits intended Appendix R alternate safe shutdown function have been established and are included in a plant procedure.

An additional plant procedure details how the above noted fire watches will be implemented.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod align-ment and insertion limits.

The ACTION statements which permit limited variations from the basic requirements

'are accompanied by additional restrictions which ensure that the original design criteria are met.

Misalignment of a rod requires measurement of peaking factors or a restriction in THERMAL POWER; either of these restrictions provide assurance of fuel rod integrity during continued operation.

In addition, those accident analyses affected by a misaligned rod are reevaluated to confirm that the results remain valid during future operation.

The maximum rod drop time restriction is consistent with the assumed rod drop time used in the accident analyses.

Measurement with Tavg greater than or equal to 541 F and with all reactor coolant pumps operating ensures that the measured drop times will be representative of insertion times experienced during a reactor trip at operating conditions.

Control rod positions and OPERABILITY of the rod position indicators are required to be verified on a nominal basis of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months with more frequent verifications required if an automatic monitoring channel is inoperable.

These verification frequencies are adequate for assuring that the applicable LCO's are satisfied.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 1-4 Amendment No.

$9,7/6, 134

POWER DISTRIBUTION LIMITS BASES 3 4.2.2 and 3 4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR

. The limits on heat, flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance 0

criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2.1, 4.2.2.2, 4.2.3, 4.2.6.1 and 4.2.6.2.

This periodic surveillance is sufficient to ensure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than plus'r minus 12 steps from the group demand position.

b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6.

c.

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

d.

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

F will be maintained within its limits as specified in the COLR pxovidect conditions a. through d. above are maintained.

The relaxation of i

as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. The form of this relaxation for

~P DNBR limits is discussed in Section 2.1.1 of this basis.

When an F measurement is taken, both experimental error and manu-facturing tolerance must be allowed for.

Si is the appropriate allowance on F

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

COOK NUCLEAR PIhNT - UNIT 2 B 3/4 2-4 MENT NO. 82)72P-

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(Continued)

When RCS flow rate and F

are measured, no additional allowances are necessary prior to comparison with the limits of Specification 3.2.3 Measurement errors of 2.1% for RCS flow total flow rate and 4% for F H have been allowed for in determination of the design DNBR value and in th5 determination of the LOCA/ECCS limit.

Margin between the safety analysis DNBRs and the design limit DNBRs is maintained.

(Safety analyses DNBRs: 1.69 and 1,61 for the Vantage 5 typical and thimble cells, respectively, and 1.43 and 1.40 for the ANF fuel typical and thimble cells, Design limit DNBRs: 1.23 and 1.22 for the Vantage 5 typical and thimble cells, respectively, and 1.39 and 1.36 for the ANF fuel typical and thimble cells.)

A fraction of this margin is utilized to accomodate applicable transition core penalties and the appropriate fuel rod bow DNBR penalty for the Vantage 5 fuel (equal to 1.3% per WCAP-8691, Rev. 1).

The remainder of the margin between design and safety analysis DNBR limits can be used for plant design flexibility.

COOK NUCLEAR PIANT - UNIT 2 B 3/4 2-4a AMENDMENT NO. 334

4.2 POWER DISTIBUTION'IMITS BASES 3 4.2.4 UADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during startup testing and periodically during power operation.

The limit of 1.02 at which corrective action is required provides DNB and linear heat generation rate protection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing the power by 3 percent from RATED THERMAL POWER fear each percent of tilt in excess of 1.0.

3 4.2.5 DNB PARAMETERS The limits on the DNB-related parameters ensure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The T less than 0

avg or equal to 578.7 F and pressurizer pressure greater than or equaY to 2200 psig are consistent with the UFSAR assumptions and have been analytically demonstrated adequate to maintain the core at or above the design DNBR throughout each analyzed transient with allowance for measurement uncertainty.

The T greater than or equal to 543.9 F is conservative to a 0

av Iz safety analysis perrormed to demonstrate that the plant may operate on a linear control program where the analytical limit of T at 100%

RATED THERMAL POWER may range from 541.4 F to 580.1 F.

The Hki.t of 543.9 F

'contains a margin of 1.1 F.

The core may be operated with indicated vessel average temperature at any value between the upper and lower limits.

Pressurizer pressure is limited to a single nominal setpoint, with the lower limit of the indicated value setpoint set forth in the specifications.

The T/S value was selected for consistency with Unit 1 and contains a margin of 6 psi.

The limits are consistent with the UFSAR assumptions and have been analytically demonstrated to be adequate to maintain the core at or above the applicable design limit DNBR values for each fuel type (which are listed in the bases for Section 2.1.1) throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation, The 12-hour surveillance of the RCS flow measurement is adequate to detect flow degradation.

The CHANNEL CALIBRATION performed after refueling ensures the accuracy of the shiftly flow measurement.

The total flow is measured after each refueling based on a secondary side calorimetric and measurements of primary loop temperatures.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 2-5 AMENDMENT NO. H2, 134

3 4.2 POWER DISTIBUTION LIMITS BASES 3 4.2.6 ALLOWABLE POWER LEVEL - APL Constant Axial Offset Control (CAOC) operation manages core power distributions such that Technical Specification limits on F -(Z) are not violated during normal operation and limits on MDNBR are no9 violated during steady-state, load-follow, and anticipated transients.

The V(Z) factor given in the Peaking Factor Limit Report and applied by the Technical Sp'ecifications provides the means for predicting the maximum F (Z) distribution anticipated during operation using CAOC taking intro account the incore measured equilibrium power distribution.

A comparison of the maximum F (Z) with the Technical Specification limit determines the power level (APL) below which the Technical Specification limit can be protected by CAOC.

This comparison is done by calculating APL, as defined in Specification 3.2.6.

COOK NUCLEAR PLANT - UNIT 2 B 3/4 2-6 AMENDMENT NO. 134

I

~

I

~ g 3 4.7 PLANT SYSTEMS BASES 3 4.7.1 TURBINE CYCLE 3 4.7.1.1 SAFETY VALVES The OPERABILITY of the main steam line code safety valves ensures that the secondary system pressure will be limited to within 110% of its design pressure of 1085 psig during the most severe anticipated system operational transient, The maximum relieving capacity is associated with a turbine trip from 100%

RATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e.,

no steam bypass to the condenser).

The specified valve liftsettings and relieving capacities are in accordance with the requirements of Section III of the ASME Boiler and Pressure

Code, 1971 Edition.

The total relieving capacity of all safety valves on all of the steam lines is 17,153,800 lbs/hr which is at least 105 percent of the maximum secondary steam flow rate at 100%

RATED THERMAL POWERS A minimum of 2 OPERABLE safety valves per steam generator ensures that sufficient relieving capacity is available for the allowable THERMAL POWER restriction in Table 3.7-1.

STARTUP and/or POWER OPERATION is allowable with safety valves inoperable within the limitations of the ACTION requirements on the basis of the reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip settings of the Power Range Neutron Flux channels.

The reactor trip setpoint reductions are derived on the following bases:

For 4 loop operation Where:

SP reduced reactor trip setpoint in percent of RATED THERMAL POWER V - maximum number of inoperable safety valves per steam line X

total relieving capacity of all safety valves per steam line in lbs./hours 4,288,450 Y - maximum relieving capacity of any one safety valve in lbs./hour - 857,690 109 Power Range Neutron Flux-High Trip Setpoint for 4 loop operation COOK NUCLEAR PIANT - UNIT 2 B 3/4 7-1 AMENDMENT NO. g7-> 134

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