ML17320A771

Amend 74 to License DPR-58,approving Cycle 8 Reload & Changing Related Tech Specs

ML17320A771
Person / Time
Site:	Cook
Issue date:	09/20/1983
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML17320A772	List: ML17320A771 ML17320A773
References
NUDOCS 8310110334
Download: ML17320A771 (65)

Text

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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 INDIANA AND MICHIGAN EL'ECTRIC COMPANY DOCKET NO. 50-315 DONALD C.

COOK NUCLEAR PLANT UNIT NO.

1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

74 License No.

DPR-58 V

g, 1.

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

a A.

The application for amendment by Indiana and Michigan Electric Company (the licensee) dated May ll, 1983, as supplemented by letter dated July 25, 1983, complies with the standards and re-

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

The facility.,will operate in conformity with the application, the-provisions.,of;Xhe,'Adt, 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.

83i0110334 830920 PDR'ADOCK 05000315 P

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PDR

a I

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h 2.

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

(2)

Technical S ecifications The Technical Specifications contained in Appendices

'A and B, as revised through Amendment. No.

74, 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

'a Ch ev

~

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1 Operating Reactors B

n h ¹1 Division of Licensing

Attachment:

Changes to the Technical Specifications, Date of Issuance:

September 20, 1903

ATTACHMENT TO LICENSE AMENDMENT Af'1ENDMENT NO. 74 TO FACILI Y OPERATING LICENSE tf0.

.DOCKET'NO.; 50-315 DPR-.58 Revise Appendix A as fol 1 ows:

.B B

B B

B B

B Remove Pa es IV 1-7 2-1 2-2 2-5 2-7 2-8 2-9 3/4 1-1 3/4 1-21 3/4 1-22 3/4 1-23 3/4 1-24 3/4 1-25 3/4 2-5 3/4 2-6 3/4 2-7 thru 3/4 2-24 3/4 3-1 3/4 3-2 thru 3/4 3-4 3/4 3-5 3/4 3-6 thru 3/4 3-8 3/4 10-1 B 2-1 B 2-2 B 2-3 B 2-4 B 2-5 B 2-6 3/4 1-1 3/4 1-2 3/4 2-4 3/4 2-5 3/4 2"6 3/4 3-3 3/4"4-1 "InsertI~Pa es IA, IV 1-7 1-10 1*

2-2 2-5 2-/

2-8 2-9 3/4 1-1 3/4 1-21 3/4 1-22*

3/4 1-23*

3/4 1-24 3/4 1-25 3/4 2-5 3/4 2-6 3/4 2-7 thru 3/4 2-24 3/4 3-1*

3/4 3-2 thru 3/4 3-4 3/4 3-5*

3/4 3-6 thru 3/4 3-8 3/4 10-1 B 2-1 B 2-l(a) 8 2-2 B 2-2(a)

B 2-3*

B 2-4 B 2-5*

B 2-6 B 3/4 1-1 B'3/4 1-2*

B 3/4 2-4 B 3/4 2-5 B 3/4 2-6 B 3/4 3-3 B 3/4 4-1 convenience

• Included for

DEP

~EDITIONS Si i:ON SONCE C~CK

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PAGE 1 - 6 1 - 6 1 - 6 CPt SITE DOSE C.Q,CULAZION ~NUAL (ODQi)

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GASEOUS RADMASiE TPZP~NT SYSTEM

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DESIGN THERMAL POWER 0

OPERATIONAL MODES (Table 1.1}.

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C FREQUENCY NOTATION (Table lo2}.

SAFETY ANALYSIS BASIS-POWER LEVELS (Table 1.3}

1-7 1-8 1-9 1-10 D. C.

COOK - UNIT 1 Amendment N0.7V

INDEX ECTION Paoe

/4.2 POWER DISTRIBUTION LIMITS

/4,2.1 Axial Flux Difference..................................

/4.2.4

/4.2.5 quadrant Power Tilt Ratio..................

D fh W C HB Parameter..........................;..........'4.2.6 Axial Power Distribution............,..................

/4.2.2 Heat Flux Hot Channel Factor...........................

/4.2.3 Nuclear Enthalpy Hot Channel Factor........

3/4 2-1 3/4 2-5 3/4 2-'l2 3/4 2->O 3/4 2-16" 3/4 2-18

/4.3 INSTRUMENTATIOH.

/4.3.1

/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION......................................

3/4 3-15 REACTOR TRIP SYSTEM IHSTRUMEHTATIOH....................

3/4 3-1

/4.3. 3 MONITORING INSTRUMiEHTATION Radiation Monitoring Instrumentation.......:....,......

Movable'ncore Detectors...............................

Seismic Instrumentation..;..............................

Meteorological Instrumentation.........................

Remote Shutdown Instrumentation.........................

Fire Detection Instrumentation.........................

Radioactive Liquid Effluent Instrumentation..............

3/4 3-35

.3/4 3-39 3/4 3-40 3/4 3-43 3/4 3-46 3/4 3-51 3/4 3-57 Radioactive Gaseous Process and Effluent Monitoring Instrumentation...............................

3/4 3-62

/4.4 REACTOR COOLANT SYSTEM

/4.4.1 REACTOR COOLANT LOOPS

/4,4. 2 Normal Operati.on.....,.....

SAFETY VALVES - SHUTDOWN...

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/4.4. 5 RR CCCll

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STEAMi GENERATORS.......................................

/4.4.3 SAFETY VALVES - OPERATING..............................

r 3/4 4

3/4 4-4 3/4 4-5 3/4. 4-.6 3/4 4-7 C:

COOK -."HIT 1

IV Amendment No. 74.'

AP w% ~ ~ '0 fQNg a

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CF ir'.".""U""LIC i".~'"'.""3',5j GF 7:";E PU"=LIC shaI'1 include all pe. sons who are not cc u"at anally associat d with:he-plant.

7nis cat gory does not include e.;.playa s o; the utility, its contractors or its vendors.

Also ~wcluce'rom this category are persons who enter the si.e to service'equipment or to aaka deliveries.

This category does include pmcns wno use portions g=.the site for re reational, oc upaticnal cr o""er puroses no associa.ad with the plant.

i = HCL410'(AY 1.:-6 ine 5!T-

"=CUNDARY shall be that line beyond which the land is nat owned, leased ar otherwise controlled by the license; U'<R:~ iRICT~D r'.R2

' 3/

An UN~EST.=. iiH AR"-4 shall be any are a

or beyond the SITE EOUNDARY t" which ac"ess is not controlled by the licensee for pu o es o p: c-.ion o=

hand-'y'idual s rcm expos re to r.dia:ion and r dicactiv rater'als or any ar wi.hin the site bcundary used f:r r s-',d ntial quar-. rs or indus-". al, ccmzrcial, institutional and/or r

" anal purposes.

DESIGN THERMAL!.POWER',

1.38 DESIGN THERMAL"..POWER shall, be adesign total r'eactor core.heat transfer rate to the reactor coolant of 3411.MWt'.-<< See..Table 1.3.

D. C.

COOK -, UNIT 1 1-7 Amendment No.,74

~ ~

~ TABLE 1:3'.

Safet Anal sis Basis-Power Levels The approved maximum power. operation and RATED THERMAL POWER is 3250 MWt.

However, certain portions of the safety analysis provided for Donald C.

Cook Nuclear Plant Unit 1

have been based on a design power of 3411 MWt.

The safety analysis for which 3411 NWt has been used is as follows:

(1)

Uncontrolled Rod Control Cluster Assembly (RCCA) withdrawal from a subcritical condition.

0 (2)

Uncontrolled control rod assembly withdrawal at power.

(3)

RCCA misalignment.

(4)

Chemical and volume Control System malfunction.

(5)

Loss of reactor f]ew (including'locked rotor).

(6)

Loss of external electrical load.

(7)

Loss of normal feedwater flow.

(8)

Excessive heat removal due to feedwater system malfunctions.

(9)

Excessive increase in secondary steam flow.

t (f0); Loss of all AC power to the plant auxiliaries.

(ll)

Rupture of a steam pipe.

(12)

Rupture of control rod drive mechanism housing (RCCA ejection).

(13)

Small break Loss of Coolant Accident (LOCA).

The'.zated thermal'ower of 3250 t1llt was the basis for the..safety.,analysis

.. l used foi'the large. break Loss Of Coolant Accident.

D.

C.

Cook Unit 1

1-10 Amendment No.

74

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer

pressure, and the highest operating loop coolant temperature (T

) shall not exceed the limits shown in Figures 2.1-1 and 2.1-2 for 4 Fkd 3 loop operation, respectively.

APPLICABILITY:

MODES 1

and 2.

ACTION:

Whenever the point defined by the combination of the highest operating loop average temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2735 psig.

APPLICABILITY:

MODES 1, 2, 3, 4 and 5.

ACTION:

/

MODES 1

and 2

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, be in HOT STANDBY with the Reactor Coolant System pressure within its limitWithin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

MODES 3, 4 and 5

Whenever the Reactor Coolant System pressure has exceeded 2735 psig, reduce the Reactor Coolant System pressure to within its limit within 5 minutes.

D. C.

COOK UNIT 1 2-1

• I

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FIGURE 2.1-1 REACTOR CORE SAFETY LIMIT I

4 LOOPS IN OPERATION

~ ~ ~

C)

C4m tD 670 UNACCEPTABLE OP ERATION 650 2400 PSIA U

0 Ol0'30 2250 PSIA hl

'J ~

Ox

)C $

0 Qv Z+

Ih

<<lh

'lsl Oa C> Ul Xg a~x 610 590 2000 PSIA

~176 PSIA 570 ACCEPTABLE OPERATION 550 0.0 0.2 0.4 0.6 ~~ 0.8 1.0 1.2 D.C.

COOK, UNIT.1

'FRACTION OF RATED THERMAL POWER 2-2 AYENDMENT NO.

74

TADLf 2. 2-1 REACTOR TRIP SYSTEI1 INSTRUIIEHTATIOH TRIP SETPOItlTS n

Pl C)

C)

FUtlCT IOHAL UHIT 1.

Ilanual Reactor Trip 2.

Power Range, Heutron Flux Hot Applicable Low Setpoint -

< 25'f RATED '

TIIERISL POWER Hot Applicable 4

Low Setpoint -

< 26%of RATED TIIERtiAL POI(fR TRIP SETPOIHT ALLOWABLE VALUES I

CJl 3.

Power Range, Heutron Flux, lligh Positive Rate 4.

Power

Range, Neutron Flux, lligh Negative Rate 5.

Intermediate

Range, tleutron Flux 6.

Source

Range, Heutron Flux 7.

Overtemperature aT 8.

Overpower.aT 9.

Pressurizer Pressure--Low 10.

Pre ssur i zer Pre ssure--lli gh 11.

Pressuri zer Water Level --lligh 12.

Loss of Flow High Setpoint - <'-109K'of RATfD

.TIIERHAL POWER

< 5g of RATED TIIERIOL POWER with a time constant

> 2 seconds

< 5g of RATED TIIERNL POWfR with a time constant

> 2 seconds

< 25" of RATED TIIERNL POWER

< 105 counts per second See ttote 1

r See tlote 2

> 1865 psig

< 2305 psi 9

< 92K of instrument span

> 90$ ol'esign flow per loop*

lligh Setpoint -

< 110%of RATED TllfRtIAL POWER

< 5.5 ~of RATED TIIERIIAL POWER with a time constant

> 2 seconds

< 5.5 %of RATED TIIERIIAL POWER with a time constant

> 2 seconds 4

< 30 %of RATED TIIERIIAL POWER

< 1.3 x 105 counts per second See Note 3 See tlote 3

> 1855 psig

< 2395 psi g

< 935 of instrument span

> 89$ of desiglt flow per loop*

• Design flow is 91,600 gpm per loop.

O

TAOLE 2.2-1 Conti'nued REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NO fATION Note 1:

Overtemperature hT < hT

[K -K 0

1+v1S (T-T')+K (P-P')-f (hl)]

1+r2S 3

1 where:

hTo Extrapolated hT at DESIGN THERMAL POllER T

D TI p

Average temperature,

. f 577.1 F (indicated Tavg at DESIGN THERMAL POllER)

Pressurizer

pressure, psig P'235 psig (indicated RCS nominal operating pressure) 1+viS 1+tpS T1 T

)

2

~

S The function generated by the lead-lag controller for T dynamic compensation.,

ayg Time constants utilized in the lead-lag controller for T x1 33 secs, T2 4 secs ayg 1

Laplace transform oper ator

TABLE 2.2-l Continued REACTOR TRIP SYSI'EH INSTRUHENTATION TRIP SETPOINTS NOTA1ION'ontinued Operation with 4 Loops K1 1.135 K2 0.0130 K3 0.000659 Operation with 3 Loops K1 0.99 K2

~

0.01026 K3

~

0.000617 and f1 (a I) 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 qt qb between -37 percent and +2 percent, f~ (aI) 0 (where qq and qb are percent DESIGN THERMAL POWER in the top and bottom halves of the core

'espectively, and qt + qb is total TNERHAL POWER in percent of DESIGN THERHAL POWER).

(ii) for each percent that the thagnitude of (qt qb) exceeds

-37 percent, the aT trip setpoint shall he automatically reduced by 2.3 percent of its value at DESIGN TllERHAL POWER.

(iii) for each percent that the magnitude of (qt qb) exceeds

+2 percent, the aT trip setpoint shall be automatically reduced by 1.8 percent of its value at DESIGN THERHAL POWER.

n TAl3LE 2.2-1 (Continued)

REACTOR TRIP SYSTEH IN TRUHENTATION TRIP SETPOINTS NOTATION (Conti nued) n C)

C)

Note 2:

Overpower aT

< aT

[K4-K<

~3S

~+vg where:

ATo Extrapolated zT at DESIGN TNERHAL POWER T

Average temperature, F.

Indicated Tavg at DESIGN THERNL POWER 577 1 F K4 a

1.0S9 Kg KG 0.0177/

F for increasing average temperature and 0 for decreasing average temperature I

I 0 0011 for T

> T"; KG = 0 for T

< T" v3S T+v~Y S

The function generated by the rate lag controller for $

dynamic compensation a

g Time constant utilized in the rate lag controller for Tag v3 = 10 secs.

Lapl ace transform operator f2(AI)

=

f1 (aI) as defined in Note 1 above.

Note 3:

The channel's maximum trip point shall not exceed its computed trip point by more than 4 percent.

3/4.1 REACTIYITY CONTROL SYSTEM 3/4.1.1 BORATION CONTROL SHUTOOMN MARGIN T

~ 200'F I ~

e LIMITING CONDITION FOR OPERATION 3.1.1.1 The SHUTDDMN MARGIN shall be

~ ].60', hkfk.

APPLICABILITY:

MOQES 1, 2, 3, and 4.

AVION:

Mith the SHUTOGMN MARGIN c 1.60K hk/k, immediately initiate and continue j

boration at > 10 gpm o, 20,000 ppa boric acid solution or equivalent until the required SHUTCOMN MARGIN is restored.

SURYEILLANCE REQUIRPIBPS 4.1.1.1.1 Co d.

The SHUTDOMN MARGIN shall be det rmined to be

~ 1.60" hk/k Nthin one hour after detection of an inoperable contml 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 /> t'hereafter while the rod(s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above required SHUTOOMN MARGIN shall be increased by an amount at least equal to the withdrawn worth of the immvable or untripoable control rod(s).

Mhen in MODES 1 or 2, at leas 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 wiDin he limits of Spe ifica-tion 3.1.'3.5.

Qhen in MODE 2, at least once during control rod withdrawal and at least once per hour thereafter until the reactor is critical.

Prior ta initial operation above 5

RATED THERMAL PO~c, 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.

Se Special Test Exception 3.10.1 Mith K ff

~ 1.0 gfMith K ff <I.O 0.

C.

CQOK - UNIT 1 3/4 1-1 Amendment th.

74

REACTIVITY CONl'ROL SYSTEMS ROD DROP TIME LIMITING CONDITION FOR OPERATION 3.1.3.3 The individual full length (shutdown and control) rod drop time from the fully withdrawn position shall be < 2.4 seconds from beginning of decay of stationary gripper coil voltage to dashpot entry with:

a.

T

) 5414F, and avg b.

All reactor coolant pumps operating.

APPLICABILITY:

Mode 3.

ACTION:

a.

With 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 proc eding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceec provided THERMAL POWER is restricted to < 76 percent of RATED i~ERMAL POWER.

SURVEILLANCE REQUIREMENTS

4. 1.3.3 The rod drop time of full length rods shall be demons-".ated through measurement prior to reactor criticality:

a ~

b.

For all rods following each removal of the reactor vessel

head, 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.

D.

C.

COOK-UNIT 1 3/4 1-21 Amendment No.

74

REACTIVITY CONTROL SYSTEMS SHUTDOWN ROD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.4 All shutdown rods shall be fully withdrawn.

APPLICABILITY:

MODES 1* and 2*¹ ACTION:

With a maximum of one shutdown rod not fully withdrawn, except for surveillance testing pursuant to Specification 4. 1. 3. 1.2, within one hour either:

a.

Fully withdraw the rod, or b.

Declare the rod to be inoperable and apply Specification 3.1.3.1.

SURVEILLANCE RE UIREMENTS 4.1.3.4 Each shutdown rod shall be determined to be fully withdrawn:

a.

Within 15 minutes prior to withdrawal of any rods in control banks A, B, C or D during an approach to reactor criticality, and b.

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.

See Special Test Exceptions 3.10.2and 3.10.4.

¹Wzth K

) 1.0 D.

C.

COOK - UNIT 1

3/4 1-22

REACTIVITY CONTROL SYSTEMS CONTROL ROD.INSERTION LIMITS LIMITING CONDITION FOR OPERATION 3.1.3.5 The control banks shall be limited in physical insertion as shown in Figures 3.1-1 and 3.1-2.

APPLICABILITY:

MODES 1* and 2*5'.

ACTION:

With the contr'ol banks inserted beyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2; either:

a.

Restore the control banks to within the limits within two hours,,or b.

Reduce THERMAL POWER within two hours to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the group position using the above figures, or c.

Be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.1.3.5 The position of each control bank shall be determined to be within the insertion limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the Rod Insertion Limit Monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

See Special Test Exceptions 3.10.2 and 3.10.4 With K

> 1.0.

D.

C.

COOK - UNIT 1 3/4 1-23

(Fvlly Withdrawn) 228 200 C

BANK C CD CA CD 150 BANK D 100 CD CD 50 0 0.0 0.'2 0;4 0.6 0.7 (Fully Inserted}

FRACTION OF RATED THERMAL POWER FIGURE 3.1-1 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER 3

LOOP OPERATION D.

C.

COOK-UNIT 1

314 1-24 Amendment No.

74

(FULLY WITHDRAWN.)

228

%a'IJ 200 cD150 BANK C 8100 CD C7 BANK D

.50 0

0.2 (FULLY INSERTED) 0.4 0.6 0.8 1.0 FRACTION OF RATED THERMAL POWER (3250 MWt)

FIGURE 3.1-2 ROD GROUP INSERTION LIMITS VERSUS THERMAL POWER 4 LOOP OPERATION D. C.

Cook - Unit 1 3/4 1-25 1

AMENDMENT NO. 74

POWER OISTRIBUTION LIMITS HEAT FLUX HOT CHANNEL FACTOR",'6 (4)

LIMITING CONOITION FOR OPERATION 3.2.2 F (Z,E) shall. be limited by the following relationships:

Westin house Fuel Exxon Nuclear Co.

Fuel

>g(z.

) ':

'1 F(z)1 L

F (E )

~q(Z <) '~p fK(z))

P >,0.5 Fq(ZS~)

< P g4j r.K(Z)j Fq(Z, a)

< 2 [F (E~) K(Z)]

P

< 0.5 p

THERMAL POWER FA tD~E F~ (E<) is the exposure dependent F~ limit for rod a and L

is defined in,Figure 3.2-4 for Exxon, Nuclear Co. fuel and in Figure 3.2"5 for Westinghouse fuel.

E< is the maximum pellet exposure in rod K(Z) is the func ion obtained from Figure 3.2"3 for Westinchouse fuel and Figure 3 2-2 for Exxon Nuclear Co. fuel.

F~ is defined as the F~(Z,E) with the s'mallest margin or the oreatest excess of the limi..

APPLICABILITY:

M9QE I ACTION:

'With F~ exceeding its limit:

a.

Comply with eit'her of the following ACTIONS:

1.

Reduce, THERMAL POWER at least I for each I

F~ exceeds the limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4

hours; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower hT Trip Setpoints have been reduced at least I for each I" F~ exceeds the limit.

The Overpower 4T Trip Setpo'.nt reduction shall be performed with the reactor in at least HOT STANOBY.

O.C.

Cook Unit 1 3/4 2-5 Amendment No.

74

POWER GIS)RIBUTION LIMITS V

~

LIMITING CONDITION FOR OPERATION'(Continued) 2.

Reduce THERMAL POWER as necessary to meet the limits of Specification 3.2.6 using the APOMS. with the lates-fncore map and updated R.

b.

'dentffy and correct the cause of the out of liming corcftion prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F~ fs 4emonstrated through fncore mapping ro be within fts limit.

SURVEILLANCE Rc UIREMENT 4.2.2.1.

Tne provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (Z,a) shall be determined to be within fts limit by:

a.

b.

Using the movable fncore detectors to obtain a power distribution map at any THERMAL POWER greater than 5X c RATED THERMAL POWER.

Increasing the measured F~(Z,a) component of the power dfstrfbutfon map by 3~ to account for manufac-urfng to erances and further increasing the value by 5~ to account for measurement uncertainties.

This product fs defined as F~~(Z).

c.

Satisfying the following y elatfonshfps at the time of =he target fl.ux determination.

Westfnohouse Fuel Exxon Nuclear Co.

Fuel Fq(Z) <

PxE (Z

~KZ V(Z)

Fq(Z) <

FO(Z)

FxX (Z)

K(Z)

P >0.5 Fq(Z) <

3. 94:.

E (Z)

+K(Z 2

F (Z)

P

+K(Z P <O,5 D.C.

Cook Unft 1

3/4 2"6 Amen~nt No.

74

POWER GISl'RIBUTION LIMlTS 8

LIMITING CONDITION FOR OPERATION'(Continued) 2.

Reduce THERMAL POWER as necessary to meet the limits of Specification 3.2.6 using the APDMS with the lates-incore map and uodated R.

b.

I'dentify and correct the cause of the out of limit cor. ition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided FQ is demonstrated through incore mapping to be within its limit.

SURVEILLANCE RE UIREMENTS 4.2.2.1.

Tne provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (Z,E) shall be determined to be within its limit by.

a.

b.

C.

Using the movable incore detec.ors to obta',n a power distribution map at any THERMAL POWER greater than 5,o c=

RATED THERMAL POWER.

Increasing the measured FQ(Z,E) component o

the power distribution map by 3 o to account for manufac-uring to rances and further increasing the value by 5~ to account for 0

measurement uncertainties.

This product is defined as Q(Z).

Satisfying the following relationships at the time of =he target fl.ux determinati on.

Westinohouse Fuel Exxon Nuclear Co.

Fuel F'(Z) ';0' PxE (Z)

P

~KZ V(Z)

FQ(Z)

~Q(Z) -'x(Z) p K(Z)

P )0.5 FQ(Z) (

4.0-E (Z)

X(Z)

V(Z) 2 F (Z)

Q~

~

EW(z p

~K~Z V(Z)

P O.5 D.C.

Cook Unit 1

3/4 2"6 Amendment No-74

POWER OIS fRIBUTIGN LIMITS SURVEILLANCE REQUIREMENTS (Continued) where FQ(Z) = FQ(Z,E) at a for which F (Z,a) is a maximum FQ(Z) = FQ(E,)

at f for which FO(Z,a) is a maximum T(E~)

FQ(Z) and FQ(Z) are functions of core height, Z, and L

FO(")

CarreSPOnd at eaCh Z tO he rad S far uhiCh ~~

is a

E maximum at that Z

V(Z) is a cycle dependenl; function and is provided. in.thePeaking Factor.

Limit Repor~.

!C(Z) is defined in Figure 3.2-2 for Exxon Nuclear Company

• f'u'el~and in Figure 3.2"3 for Westinghouse fuel.

T(E<) is defined in

- Figures 3.2-4 and 3.2-5.

E (Z) is an uncertainty factor to account P

for the reduction in the FQ (E ) curve due to accumulation of L

exposure prior to the next flux map.

Westin house Fuel E (Z) = 1.0' E (Z) = 1.0 P

E (Z) = i.o P

Exxon Nuclear Co.

Fuel E (Z) = 1.0 P

E (Z) = 1.0 + L.0040 x FQ(Z)]

P E (Z) = 1.0 + I'.0093 x FQ(Z)]

P 0<

E< 17,62 17.62 E

< 34.5 34.5 EE < 42.2 O.C.

Cook Unit 1 3/4 2-7 Amendment No.

74

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) d.

Measuring FQ(Z,t) fn conjunction with a target flux differenc and target band determination, according to the following schedule:

2.

Upon achieving equilibrium conditions after exceeding by 10'r more of RATED THERMAL POWER, the THERMAL POWER at which FQ(Z) was last determined",

or At least once 'per 31 effective full power days, whichever occurs first; "During power escalation at the beginning of each cycle, the design target may be used until a power level for extended operation has been achieved.

With successsive measurements indicating an increase in max over FQ(Z)

Z of [~K Z ] with exposure, either of the following addi ional actions shall be taken:

1.

F'(Z) shall be increased by 2.o over tha specified in 4.2.2.2.c, or 2.

FQ(Z) shall be measured and a target axial flux difference reestablished at least once per 7 effective full power days until 2 successive maps indicate that max over Z

FQ(Z) of [

] is not increasing.,

K(Z)

With the relationship specified in 4.2.2.2.c not being satisfied, either of the following actions shall be taken:

1.

Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied and remeasure the target axial flux difference.

D.C. Cook Unit 1 3/4 2-8 Amendment No.

74

POWER OISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 2.

Comply with the requirements of Specification 3.2.2 for F (Z,t) exceeding its limit by the maximum pe'rcent calculated with..he following expressions with V(Z) corresponding to the target band and P > 0.5:

M F (Z) x V(Z) x E (Z) max. over Z ef Q

F(

(<~) x P(z)3 P

Exxon

-1 x 100 Nuclear Co FO(Z) x'(Z) x E (Z) max. over Z of p,

MESTINGHOUSE FUEL "1 x 100 The limits'pecified in 4.2.7,.2.c and 4.2.2.2.f applicable in the following core plane regions:

1.

Lower core region 0 to 10 inclusive.

g aoove are not 2.

Upper core. region 90>> to 100>> inclusive.

4.2.2.3 When FQ(Z,f) is measured for reasons other than meeting the requirements of Specification 4.2.2.2, an overall measured FQ(Z,a) shall be obtained from a power distribution map and increased by 3" to account for manufacturing tolerances and further increased by 5>> to account for measurement uncertainty.

D.C. Cook Unit 1

,', 3/4 2-9 Amendment No.

74

Cl 1.2 Cl OO 1.0 0)

(6.0, 1.0)

(11.01,

.936)

'.8 tV QJN Of 0.6 E

O 0.4 (12.0,

.4902).

0.2 O

~

O 0

2

'6 Core Ileight (Ft.)

.8 10 12 Figure 3.2-2 K(Z) - Normalized F0(Z) As A Function of Core lleight for Exxon Nuclear Coihpany Fuel

1.2 1.0

~ Oy il ~

(0 0

(6.0, 1.0)

(11 407 g32) 0.8 (12.0..761) 0.6 E

0.4 I

0.2 0.0 6

8 10 12 Core Height (FT)

Figure'3, 2-3 k(Z)-- Normal'ized Fq (Z) As A Function of Cori Height For westinghouse Fuel D. C. Cook - Unit 1

3/4 2-11 Amendment iNo.'<

POWER DISTR IBUTI IMITS

. ~

NUCLEAR ENTHALPY HOT CHANNEL FACTOR -

F~H LIMITING CONDITION FOR OPERATION 3.2.3 F shall be 11mited by the following relat1onshfps:

F>H

.1.49 [I + 0.3 (I-P)]

(for West1nghouse fuel) and F<H

~ 1.45 [I + 0.2 (I-P)]

'for Exxon Nuclear Co. fuel) where P's the fract1on of RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

With F~ exce ding its limit:

.a.

Reduce THERMAL POWER to less than 50K of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints to < 55K of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.

Demonstrate through in"core mapping that F>H is'ithin its limit with1n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceed1ng the limit.or reduce THERMAL POWER to 'less than 5X of RATED THERMAL POWER within the next, 2

hours, and c.

Identify and correct the cause of the out-of-limit condition prior to 1ncreasing THERMAL POWEP.;

subsequent POWER OPERATION may

proceed, provided that F<H is'emonstrated through in-co're.

mapping to be within its limit at a nominal 50~ of RATED THERMAL POWER pr1or to exceeding this THERMAL POWER, at a nominal 75K of RATED THERMAL POWER pr1or to exceeding this THERMAL power and with1n 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after atta1ning 95~ or greater RATED THERMAL POWER.

D.C; Cook Unit I 3/4 2-l2 Amendment No.74

POSER OISTRIbUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 a.

F shall be determined to be within its limit by using the movable incore detectors to obtain a power distribution map:

Prior to operation above 75~ of RATED T))cRMAL PCMER af.er each fuel loading, and b.

At least once per 31 Effective Full Power Days.

C.

i'he provisions of Specification 4.0.4 are not applicable.

O.C.

Cook Unit I 3/4 2-13 Amendment No.

74

POWER DISTRIBUTION Lll1ITS

. QUADRANT POWER TILT RATIO Lll'IITING COteDITION FOR OPERATION 3.2.4 THE QUADRANT POWER TILT RATIO shall not exceed 1.02 APPLICABILITY llODE 1 ABOVE 50"OF RATED THERfNL POWER*

ACTION:

a.

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 but

< 1.09:

1.

Mithin 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

a)

Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERlNL POWER at least 3" from RATED THERlNL POWER for each 1" of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

2.

Yerify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduce THERlNL POKIER to less than 50>> of RATED THERlNL POMER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip setpoints to

< 55" of RATED THERlNL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.

Identify and correct the cause of the out of limi condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above $0$

of RATED THERtNL POWER may proceed provided that the'UADRANT POWER TILT RATIO is verified within its limit at least ence per hour until verified acceptable at 95or greater

?ATED THERfNL POWER.

b.

Mith the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod:

I.

Reduce THERIaAL POWER at Ieast 3E from RATED THERIIAL POMER for each 35 of indicated QUADRANT POWER TILT RATIO in excess of 1.0, within

. 30 minutes.

2.

Yerify that the tlUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit or

• See Specia Test Exception 3. 10.2 D.C.

COOK - UNIT 1 3/4 2-14

'.amendment No.

74

POWER OISTRI""UiION LIMITING CQNQI i iON FOR OPERATION (Conti nued) r duce 7ifHFAL POMER to less than 50" of RAI CD 7iic~L PGAER within the nex 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range

'i urn Flux-High trip Setpoints to <:=" o-.

RA~i-is!EEL POWER wi Sin the nex 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

3.

I'densify and carr ct the cause of the out c, limit can-dftion prior to ',ncrmsinq THUHAL PCMER; subsequent POMME OPHATI'ON above 50>> or RA1<D 7nEPPAL Pic~

may prcca d

provided tha" the gUAQRANT PCMcZ TILT'ATIQ is verified within fts limit at least once per hour un-'1 v rfffed acceptable at 46 or gr ater RAZZ 7r'. MAL POMME.

c.

Mfth

'ie gUAQRANT POMER TILT RATIO det rmined to xceH 1.09 due to causes other than the mfsalfgnment cf ei-ver a shu--

down cr c"n. ol rod:

1'.

Reduc 7i!CHAL PGMER to less than 50~ cf R"TH 7iiHPAL PEEK within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ancf i duc the Power?ange tteu ran Rux-Hfgn Trip Setpofnts to

< =">> of RAiB AQNL PUB within.he next 4-hours.

2.

Ident,y and cor, "

'"." "".""e c,'h cu c= lfm'., c"o-s dition prior to fnc. wsing 78pHAL POACH; s bseru n:

PCAER OP%ATION above 50 o= RATH 7i!EWAL P$ 4&c.

may pr.c provided that the gUAQRANT PGWH TILT RAii& fs ve.1 mthin fts limit at. leas cnc per hour un=. i verified ai 95>> ot-grea er RAi~D 7n "ML PCMER.

SURVEILLANCE REOUIR tElTS 4.2.4 The gUAQRANT PGMER TILT RATIO shaTl be det rmfned m be within ze 1 fmft above 50" o, RAi~D iriERMAL PONER by:

CaTcula. ing the ra~io at least'nc per 7'ays wnen the alarm fs OP&ABLE.

b.

Calculati'nq the ratio at leas onc per 12 hcurs during s-dy s.m'>> operation when the. alarm fs fnoper aole.

c ~

Usfne the movabl incore detec ors

.o det rmine,e CUAQPANT POMER TILT RATIO at leas.

onc per 12 hcurs when one Pcwer Range Channel is inoperable and 7':-.='VL PW'E.=. is 7"= percen of RATA~ 7'llAL r"C~ER.

Q.

C.

COOK -

UNji 1 3/4 2-15 RQC-alt Ho. 74

POWER DISTRIBUTION LIMITS DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2"1:

a.

Reactor Coolant System T

b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate APPLICABILITY:

MODE 1 ACTION:

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

SURVEILLANCE REQUIRBlENTS 4.2.5. 1 Each o'f the parameters of Table 3.2-1 shall be veri-ied to be within tneir limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The Reactor Coolant System total flow rate shall be determ'ined to be within fts limit by measurement at least once per mon-h.

3/4 2-16 D. C.

COOK-UNIT 1 Amendment No.

TABLE 3.2-1 DNG PARAMETERS O'

7C PARAMETER Reactor Coolant System Ta>

Pressurizer Pressure Reactor Coolant System Total Flow Rate

. 4 Loops In Operation at RATED THERMAL POWER

< 570,5'F

> 2220 psia*

> 1.386 x 108 lbs/hr L!MITS 4 Loops In Operation a t DESIGN THERMAL POWER

< 579.8 F

> 2220 psia*

> 1.386 x 108 lbs/hr 3 Loops in Operation at RATED THERMAL POWER

<'Syo, gl'F

> 2220 psia*

>0.9917 x 100 Ibs/br imit not appl cab e

ur ng either a THERMAL POWER ramp increase in excess of 5 percent RATED THEfUIA4 POWER per minute or a THERMAL POWER step increase'in excess of 10 percent RATED THERMAL POWER.

POWER DISTRIBUTION LIMITS AXIAL POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION.

3.2.6

- The axial'.power distribution shall be limited by the following relationship:

Westin house Fuel 1.97 I K(Z 1 J

s (P )(1.03)(1

+ a.)(1.07 F

(RJ)

L J

Exxon Nuclear Co.

Fuel I'2.041 IK Z

~s P (1.037(1

+ a-1.07 F

(R))

L J

P where:

a.

F.(Z) is the normalized axial power distribution from.himble Jj at core elevation Z.

b.

PL is the fraction of RATED THERMAL POWER.

c.

K(Z) is the function obtained for a given core height location from Fioure 3.2-2 for Exxon Nuclear Company fuel and from Figure 3.2"3 for Westinghouse fuel.

d.

R., for thimble 3, is determined from at least n=6 in-core J'lux maps covering the full configuration of permissible rod patterns at 100~ or APL (whichever is less) of RATED THERMAL POWER in accordance with:

n R.=-

E R

n where:

Meas F i

/T(B,)

~ i

%ax R

ard its associated ai may be calculated cn a full core or a limiting fuel batch basi's as defined on page 8 3/4 3-3 of basis.

D.C..Cook Unit 1 3/4 2-18 Amendment No.

74

POWER OISTRIBUTlON LIMITS LIMITING CONOITION FOR OPERATION (Continued) e.

F~i, is the limiting total oeaking factor

~n flux Yeas map i.

The limiting total peaking fac.or is that factor with least margin to the F~(Ea) curve defined in Figure 3.2-4

-L for Exxon Nuclear Company fuel and in Figure 3.2-5 for Westinghouse fuel.

For Exxon Nuclear Company fuel, T(ER) is the ratio of =he exposure dependent F~(E) to 2.04 and is defined in Figure 3.2-4.

T(B) is equal to 1.0 for fuel suoolied by Westinghouse Electric Corporation as given in Figure 3.2-5.

[Fi (Z)3M is the max'mum value of the normalized axical ij Max distribution at elevation Z from thimble j in map i which had a

limiting total measured peaking fac.or without uncertaianties or densification allowance of F~i Reas a.

',s the standard deviation associated with thimble j, j

expressed as a.fraction or percentage of R.,

and is derived from n flux mans from the relationsnip below, or 0.02, (c o) wnichever is greater.

n 2

]/2 n"1 5 ~1 i i j

R-The factor 1.07 is comprised of 1.02 and 1.05 to account for the axial power dis.ribution instrumentation accuracy and the measurement uncer ainty associated with F~ using the movable detector system respectively.

The factor 1.03 is the engineering uncertainty factor.

Amendment Nc.

74 3/4 2-19 g.

F is an uncertainty factor for Exxon fuel to account

=or the p

reduction in the F~(L)curve due ro an accueulation or exposure prior to the next flux map.

The following F factor p

shall apply:

D.C.

Cook Unit 1

0, POWER DISTRIBUTION LIMITS V

LIMITING CONDITION FOR OPERATION (Cohti.nued)

Westin house Fuel F

= 1.0 p

F

= 1.0 P

F

= 1.0 ENC Fuel F

= 1.0 P

F

= 1.0 + [.0015 x W]

p F

= 1.0 + [0.0033 x W]

P 0<

E

<17.62 17.62 E

< 34.5 34.5 E

< 42.2 where W is the number of effective full power weeks (rounded up to the next highest integer) since the last full core flux map.

A APPLICABILITY:

Mode 1 above the minimum percent of RATED THERMAL POWER indicated by the relationships."

APL, = min over Z of 1.97 x

X Z)

F<(z,a) x V(Z) x 100 ~

Westinghouse Fuel L

FO (E ) x K(Z)

APL = min over Z of F (Z E) x V(Z) x E (Z) x 100,.

P Exxon Nuclear Co.

Fuel wnere F (Z,a) is the measured F~(Z,E), including a

3 o manufacturing tolerance uncertainty and a 5;o'easurement uncertainty, at the time of target flux determination from a gower distribution map using the movable incore detectors.

V(Z) is the function given in the Peaking Factor imit Report.

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

1.

Lower core region 0 to 10~ inclusive.

2.

Upper core region 90~ to 1005 inclusive.

"The APOMS may be out of service when surveillance for determining power distribu.ion maps is being performed.

O.C.

Cook Unit 1

,gC'/4 2-20 Amendmen No.

74

POWER OISTR1BUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

ACTION:

a.

With a F.(Z) facto exceeding (F.(Z)]S'y < 4 percent, J

j S

reduce THERMAL POWER 1 percent for every percen-by which the F-(Z) factor exceeds its limit within 15 minutes and within J

the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either reduce the F.(Z) factor to within its J

limit or reduce THERMAL POWER to APL or less of RATED THERMAL POWER.

b.

With a F.(Z) factor exceeding [F.(Z)]

by > 4 percent, J

J reduce THERMAL POWER to APL or less of RATED THERMAL POWER within 15 minutes.

O.C.

Cook Unit 1 3/4 2-27 Amendmen No.

74

POWER DISTRIBUTION LIMITS SURVEILLANCE REGUIRB1ENTS

.4.2.6.1 a

F.(Z) shall be determined to be within its limit by:

Either using the APOMS to monitor the thimbles required per Spec i ication 3.3.3. 6 at he ro 1 1 owing rr quenc ies.

1.

At least onc per.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 2;

Ijrmediately and at intervals of 10, 30, 60, 90, 120, 240

'and 480 minutes fol iowing:

a)

Increasing the THE&PL POWER above APL of RATED THUHAL POWER, or b)

Movement of control bank "O" more than an accumu1:ted total of 5 steps. in any one direction.

b.

Or using the movable incore detec ors a t. the ol lowing rre-quencies when the APOMS is inoperable:

1.

A least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and 2;

At intervals of 30, 50, g0, 120, 240 and 480 minutes 7 0 1 lowing

')

Increasing the THER/!AL POWER above APL of RATED THER!AL POWER, or b'ovement of control bank "0" more than an ac=umulat d

total of 5 steps in any one direction.

4.2.6.2 When the movable incore detectors are used to monitor F. (Z), at least 2 thimbles shall be monitored and an F.(Z) accuracy equivalent to tha obtained rrcm the APOMS sha 1 1 be ma inta) ned.

O.

C.

COOK - UNIT 1

3/4 2-22 Amendment No.

74

2.2 2.0 (0. 2.04)~ (17.62, 2.04)

(34.5, 1.,55)

UJ 1.9 L4 1.8 1.7 1.6

~FL Q

-'l FL (E<)

= 2.04 (42.2, 1.86)

~0 Ea

< 17,6Z I

(EE)

= 2.134-0.005333 E2 = 17.62<

Ea-'< 34.5 (Ei) = 2.353-.01169 Ei

'34,5

< Ez < 42,2 (0, 1.0) ~

(>> -"2, 1

"-)

1.0.

(34.6,

.966)

(42,2

.912)

.8-

.7 T(ER) =,1.0 T(Ex) = 1.046-.002614 Ex T(Eg)

= 1.154-.00573 Eg,

~(0 <

E

< 17.62 f17.622 Eg 'c 34 6

34.5 Eg

< 42.2 10 20 30 Peak Pellet Exposure in MMO/KG 50 FIGURE 3.2-4 Exposure Dependent FQ Limit, FQ (Ex), and Normal ized Limit T(ER) as a function o, Peak Pellet Burnup for Exxon Nuclear Company Fuel D.C. Cook -. Unit 1

3/4 2-23 Amendment No.

74

2 2 2.1 2 0, (0,.'l'.97.)

(42.2,,l.97).

1.7 1.6 1.0 (0,1.00)

(42.2, l.00) 0.8 0.7 0

10 20 30 PEAK PELLET EXPOSURE IH MMO/KG FIGURE 3.2"5 40 44 Exposure Oependent F~ Limit, F~ (E ), and Normalized Limit L

T(E ) as a Function of Peak 'Pellet Burnup for Westinghouse Fuel O.

C. Cook - Unft 1 N

1 A

3/4 2-24 Amendment t(o. 74

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TAOLE 3.3-1 REACTOR TRIP SYSTEH IHSTAUHIEHTATIOH FUttCTIOttAL UttIT Hanual Reactor Trip 2.

Power Aange, tleutron Flux 3.

Power Range, Neutron Flux tligll Positive Rate 4.

Power Aange, Neutron Flux, ttigl> ttegative Aate 5.

Interatediate

Range, Heutrgn Flux 6.

Source

Range, tteutron Flux A.

Startup 0.

Sliutdown TOTAL t<0.

OF CIIAHttELS 2

c CltAHttELS TO TRIP HINIHUtI Ci!AHHELS OPERABLE 3

APPLICABLE HOOES 1,

2 and*

1, 2

1, 2

1, 2

1, 2 and*

2 and

• 3, 4 and 5

ACTIOII 12 28 7.

Overteoiperature aT Four Loop Operation Three Loop Operation 21**

3 1,

2 3

1, 2

O

TABLE 3. 3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION n

C3 I

FUHCT IOHAI UNIT TOTAL NO, OF CIIAHNELS HlNIHUH CllRNNELS CHANNELS APPLICABLE TO TRIP OPERAOLE MORES ACTION Q.

Overpower AT Four Loop Operation Three Loop Operation 9.

Pressurizer Pressure-Low'0.

Pressurizer Pressure lligh 11.

Pressuri zer ala ter Level--lligh 12.

Loss of Flow - Single Loop (Rl ove P-O) 3/loop 2/loop in any oper-ating loop

'3 1,

2 1,

2 1,

2 1,

2 2

1, 2

2/loop in 1

each oper-ating loop

) ~

9 13.

Loss of Flow - Two Loops (Above P-7 and below P-8) 3/loop 2/loop in 2/loop 1

two oper-each oper-ating loops ating loop m

m C)

Cl C

FUHC1 IONAl UNIT 14.

Steam Generator Mater LevelLow-l.ow TOTAL HO...

CllAHNELS OF CHANHELS TO TRIP 3/loop 2/loop in any oper-ating loops TABLE 3.3-1 Continued REACTOR TRIP SYSTEH IHSTRUMENTATIOH HIHIMUH CIIAHHELS APPLICABLE OPERABLE HODES 2/loop in 1,

2 each oper" ating loop ACTION 15.

St,eaai/Feedwater Flow Hismat,ch and Low Steam Generator Water Level 2/1 oop-1 evel and 2/1 oop-f1 ow mismatch in same loop 1/1 oop-1 ovel coincident with 1/loop-f1 ow mismatch in same loop 1/1 oop-1 evel and 2/1 oop-flow mismatch or 2/1 oop-level and 1/1 oop-flow mismatch 1,

2 O

16.

Undervol tage-Reactor

Coolant, Pumps

17. 'nderfrequency-Reactor Coolant Pumps IB.

Turbine Trip A.

Low Fluid Oil Pressure 0.

Turbine Stop Valve Closure 19.

Safety In)ection Input from ESF 4/1/bus 4-1/bus 3

2 4

1, 2

7P

TABLE 3. 3-1 Continued REACTOR TRIP SYSTEM INSTRUMENTATION FUNCTIONAL UNIT 20.

Reactor Coolant Pump Breaker Position Trip A.

Above P-8 B.

Above P-7 21.

Reactor Trip Breakers 22.

Automatic Trip Logic TOTAL NO.

OF CHANNELS 1/breaker 1/breaker CHANNELS TO TRIP MINIMUM CHANNELS APPLICABLE OPERABLE MODES 1/breaker 1/breaker per oper-ating loop 1, 2*

l. 2*

ACTION 10ll

TABLE 3.3-1 Continued TABLE NOTATION

• With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal.

~he channel(s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped condition.

PThe provisions of Specification 3.0.4 are not applicable.

88High voltage to detector may be de-energized above P-6.

ACTION STATEMENTS ACTION 1-ACTION 2-With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />;

however, one channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveill.ance testing per Specification 4.3.1.1. 1.

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/ot POWER OPERATION may proceed provided the following conditions are satisfied.

a.

The inoperable channel js placed in tripped condition within I hour.

b.

C, ACTION 3-The Minimum Channels OPEPABLE requirement is met:

however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.F 1.1.1.

Either, THERMAL POWER is restricted to 4 75K of RATED THERhNL POWER and the Power Range, Neutron Flux trip setpoint is reduced to ~ 855 of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.c.

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

D.

C.

COOK - UNIT 1 3/4 3-6 Amendment No.

74

a.

b.

C ~

ACTION 4-a.

TABLE 3. 3-1 Continued Below P-6, restore the.inbperable channel to OPERABLE status prior to i%crees'3'hg THERMAL POWER above the P-6 Setpoint.

Above P-6 but below 5X of RATED THERMAL POWER, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 5X of RATED THERMAL POWER.

Above 5X of RATED THERMAL POWER, POWER OPERATION may continue.

With the number of channels OPERABLE one less than required by the Minimum Channels'OPERABLE requirement and with the the THERMAL POWER level:

Below P-6, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint.

b.

Above P-6; operation may continue.

ACTION 5-ACTION 6-a b.

With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, verify compliance with the SHUTDOWN MARGIN requirements of Specification

3. 1. 1. 1 or 3. 1. 1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> 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.

With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

~

The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing of the other channels per Specification 4.3. 1. 1. 1.

ACTION 7-With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed until performance of the next required CHANNEL FUNCTIONAL TEST provided the inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, D. C.

COOK - UNIT 1 3/4 3-7 Amendment No. 74

TABLE 3.3-1 Continued ACTION 8 - Mitk the number of OPERABLE channels one less than the Total Numbers of Channels and with the THERHAL POWER level above P-7, p'lace the inoperable channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; operation may continue until per ormance of the next required CHANNEL FUHCTIONAL TEST.

ACTIOH 9 - With a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in 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 />; ho~ever, one channel associated with an operating loop may be bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4, 3.1. l. l.

ACTION 10 - With one channel inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERHAL POWER to below P-8 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Operation below P-8 may continue pursuant to ACTIOH 11.

ACTION ll - With less than the Hinimum Number of'hannels OPERABl.E, operation may continue provided the inoperable ch nnel is placed in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ACTION 12 - With the number.of channels OPERABLE one less than required by the Hinimvm Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in 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/or open the reactor trip. brea.kers.

DES IG>'IATION REACTOR TRIP SYSTEH I>lTERLOCKS CONDITION AND SETPOIHT FUHCTrOH P-6 With 2 or 2 Intermediate Range 11 Neutron Flux Channels 6 x li0 amps.

P-6 pr events or defeats the manUal bloc!( of source ranqe reactor trip.

D.

C.

COOK - UNIT1 3/4 3-8 Amendment Ho.

74

\\

3/-'.10 5PEC!AL K~".=.PT.GNS SHUTDC"H HARG:H.

LDllT.HC CONOli:OR FOR OicRATTO'l 3.10.1 The SHUT)DCMN VRGIH requirement of Specification 3.1.1.1 may be suspended

-,or me surement o= control rod worth and shutdown margin provided the re c"iv.ty equivalen.

to a. least the h ohes estima control rod worth is available

",or trip inser:ion from OP)=ML"- control rod(s).

APPL'CABAL! ) Y:

MODF 2.

ACTIOi'"

a

~

Mith the reac or critical (K

> 1.0) and wi h less than the above reactivity equivalent availaolh for trip insertion, immediat ly initiaie and continue cora, ion a

> 'l0 gp'c",

20,000 ppm bor',c acid solution or its equivalen..

un"i1 "he SHUTOOk'H PARGlH required by Speci ication 3.1.1.1 is r s ored.

M'.th th reactor subcritical (K,

< 1.0) by less than th above reactivity equivalent, irmodiat31y initia" and cortinue boration at

> 10 gpm o, 20,000 ppm boric acid solu ion or 'w equivzlen. unt 1

the SHUT)DGMH HARGiH required by Speci=',cation 3.1.1.1 is rector d.

SU"VK

)

) AHC"- RENDU:R~"=HTS 4.10.1.1 The position o, each,ull length rod ei.her partially or fully witndrawn shall be determined a

least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

4.10.1.2 Each full length rod not fully inserted shall be demonstra't d

OPERABLZ bv verifying its rod drop time to be

< 2.4 secords within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to reducing he SHUTDOWN HARGiH:o less

.."an

.he limi.s of Speci-,ication 3.1.1,1.

0.

C.

COQV.-UH'.7 1

3/4 10-1

~endm n: Ho, 74

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 restrictino fuel operation to within the nucleate boiling

regime, wnere the heat transfer coefficient is large and the cladding surface emperature is slightly above the coolant saturation emperature.

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

ONB is not a directly measurable parameter during operation and therefore, THERMAL POWER and Reactor Coolant Tempera-ture and Pressure have been related to ONB.

This relation has been developed to predict the DNB flux and the location of GNB for axially uniform and non-uniform heat flux dis.ributions.

The local ONB heat flux

ratio, DNBR, defined as the ratio of the heat flux that would cause ONB at a particular core location to the local heat flux, i.s indicative of the margin to ONB.

The DNB design basis is as follows:

there must be at least a 95 perceni orobability with 95 percent confidence that DNB will not occur when the minimum ONBR is at the design DNSR limit.

In meeting this design basis, uncer ainties in plant operaiing para-

meters, nuclea.

and hermal parameters, and fuel fabrication parameters are considered statistically, such thai there is at least a

95 percent confi-dence that the minimum DNHR for the limiting rod is greater than or equal to the aoplicable design DNBR limit for each fuel type (as def-;ned below).

For 4 loop operation, the improved hermal design procedure is used.

The uncer ainties in the plant parameters are used to determine he plant ONBR uncertainty.

This ONBR uncertainty, combined with the correlaxion ONBR limit (as defined below),

es ablishes a design ONBR limit value, which must be met in plant safety analyses, using values of input parame-ers without uncertainties.

For 3 loop operation, a conservative set of uncertainties are used in the safety analyses.

The table below indicates the relationship between the correlation limit DNBR, design limit DNBR, and the safety analysis limit ONBR values used for this design.

O.C.

Cook Unit I B 2-1 Amencment No.

74

2.1 SAFETY LIMITS BASES 4 Loop Operation 3 Loop Operation (WRB-1 Correlation)

(W-3 Correlation)

(W-3 Correlation)

Westinghouse Fuel (15x15 OFA)

Exxon Nuclear Co.

Fuel (15x15)

W and ENC Fuels Typica 1 Thimbl e Typical Thimbl e Typical Thimbl e Correlation Limit Oesign Limit DNBR afety Analysis Limi DNBR 1.69 1

~ 69 1.17 1.17 1.32 1.31 1

~ 30 1.58 1'8 1.50 1.30 1.30 1.30 1.50 1.30 1.30 1.30 1.30 The curves of Fioures Z. 1-1 and 2. 1-2 show the loci of points of HERMAL POWER, Reactor Coolant System pressure and averaoe temperature for hich the minimum ONBR is no less than the applicable desian ONBR limit, or he averaoe entnalpy at the vessel exit is equal to the enthalpy of saturated liauid.

D.C.

Cook Unit 1

B 2-1(a)

Amenc...ent No.

74

SAFETY LIMITS BASES N

The curves are based on an enthalpy hot channel factor F, of 1.49 for Westinghouse fuel and an F

H of 1.45 for Exxon Nuclear Co. fuel and a

reference cosine axial power shape with a peak of 1.55.

An allowance is included for an increase in F<H at reduced

power, based on the expressions:

~4H 1.49 [1 + 0;3 (1-P)]

and F;H = 1.45 [1 + 0.2 (1"P)]

(for Westinghouse fuel)

(for Exxon Nuclear Co. 7uel) where,P is the fraction of RATED THERMAL POWER

Note, do not include a 4~ uncertainty value, since this measurement uncertainty has been included in the design DNBR limit values; which are list d in the bases for Section

2. 1. 1.

Although the N-loop operation curves are calculated for operation at DESIGN THERMAL POWER, F H values for RATED THERMAL POWER are reported here in order to be consistent wi h Sec"ion 3.2.3.

Tne r

values of Section 3.2.3 are limited by.he LOCA analyses wnich aH were performed at RATED THERMAL POWER.

These limiting heat flux conditions are higher than

".hose calculated

'or the ranoe of all control rods fully withdrawn to the max',mum allowable control rod insertion, assuming the axial power imbalance is wi.hin the limits of the fl (h!) function of the Over.emperature trip.

When the axial power imbalance is not within the tolerance, the axial power imbal-ance effect on the Over temperature hT tr ips will reduce the setpoints to provide protection consistent with the core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protec s the integri y of the Reac or Coolant Sys em from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolan from reaching the containment atmosphere.

O.C.

Cook Unit 1

B 2-2 Amendmenw No. 74

SAFETY LIMITS BASES The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant, which permits a

maximum transient pressure of 110,O (2735 psig) of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pr ssure of 120.O (2985 psig) of component design pressure.

The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 1255 of design

pressure, to demonstrate integrity prior to initial operation.

D.C. Cook Unit 1

B 2-2(a)

Amendment No.

74

2.2 LIMITING FETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each parameter.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits.

Opera-tion with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that each Allowable Value is equal to or less than the drift allowance assumed for each tr ip in the safety analyses.

Manual Reactor Tri The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Ran e, Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry.

The low set point provides redundant protection in the power range for a power excursion beginning from low power.

The trip associated with tHe low setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a power level of above approximately 9

percent of RATED THERMAL POWER) and is automatically reinstated when P-10 becomes inactive (three of the four channels indicate a.power level below approximately 9 percent of RATED THERMAL POWER).

Power Ran e, Neutron Flux, Hi h Rates The Power Range Positive Rate trip provides protection against, rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip complements the Power Range Neutron Flux High and Low trips to ensure that the criteria are met for rod ejection from partial power.

D.

C.

COOK UNIT 1

B 2-3

SAFEeY LIMITS BASES The Power Range Hegative Rate Trip provides protection for control rod drop accidents.

At high power, a rod drop accident could cause local flux oeaking which could cause an unconservative local ONBR to exis-.

The Power Range Heaative Rate Trip will prevent, 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 ac"idents for which the ONBR' will be greater than the applicable design limit ONBR value for each fuel type.

Intermediate and Source

Range, Nuclear Flux The '.ntermediate 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 channels.

The Source Range Channels will initiate a reactor trip at about 10 counts per. second, unless manually blocked when P-5 becomes active.

+5 The Intermediate Range Channels will initiate a reactor trip a-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 ooeration of the trips associa.ed with either the Intermediate or Source Range Channels in the accident analyses;

however, thei r functional capabi 1-ity at the specified rip settings is required by this specification to enhance

.he overall reliability of the Reactor Protection System.

Qvertemoerature hT The Overtemperature 4T trip provides core protect',on to prevent ONB for all combinations of pressure, power, coolant tempera

ure, and axial power distribution, provided tha. the transient is slow with

." spec to piping transit delays from the cor to the temperature detectors (about 4

seconds),

and pressure is within the range between the High and Low Pres-sure reac.or trips.

This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensa-tion for piping delays from the core to the loop temoerature detectors.

With normal axial power distribution, this reactor t". ip limit is always below the core safety limit as shown in Figure Z. 1-I. If axial peaks are greater than design, as indicated by the difference between top and bot=om power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

O.C.

Cook Unit I 8 2-4

~iiendment Ho.

74

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SAFETY LIMITS BASES through the pressurizer safety valves.

No credit was taken for operation of this scrip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Loss OT Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90~ of nominal full loop flow.

Above 51~ (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 905 of nominal full loop flow.

This latter trip wi':1 prevent the minimum value of the DNBR rom going below the applicable safety analysis design limit DNBR value for each fuel type, (as lis ed in the bases fo~ Sect. on Z. l. 1) during normal operational transients and anticipated transien.s when 3 loops are in operation and the Overtemperature 4T trip setpoint is adjusted to the value specified for all loops in operation.

With the Overtemperature hT trip setpoint adjusted to the value specified for 3 loop ooera..on, the P-8 trip at 76,~

RATED THERMAL'POWER will prevent the minimum value of tne DNBR from going below the applicable safety analysis design limit DNBR value for each fuel type, (as listed in the bases for Sec.'.on Z. 1.') curing normal ooerational transients and an icipated transients when 3 loops are in operation.

Steam Generator Water L vel The Steam Generator Mater Level Low-Low rip provides core protection by preventing operation with the steam aenerator water level below the minimum volume required for adequate heac removal capacity.

The specified setpoint provides allowance that there will be sufficient water inventory in the steam generator s at the time of trip, to allow for s arting delays of the auxiliary feedwater system.

Steam/Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam/Feedwater Flow Mismatch in coincidence with a S

earn Generator Low Water Level trip is not used in the transient and accident

analyses, but is included in Table Z.Z-l to ensure the functional capa-bility of the specified trip settings and thereby enhance the overall D.C.

Cook Unit 1

B 2-6 Amendment No.

74

3/4.1 REACTIVITY CDNTROL SY~i 3/4 ~ 1, 1 HORA i ION CONTROL 3/4.1.1.1 and 3/4.1.1.2 SHlPiDOWN MARGIN A sufficient SHUTDOWN MARGIN ensur m that 1) "Ne reactor can be invade subcr tical frcm all operating canditions;

2) the reactivi"y mnsient" ass'ociatad with postulated acciden" conditions ara controllable within accap+ ble limits, and 3) the reactor will be maintained su-.=.iciently subcritical ta preclude inadvertent cri icality in the shuAcw condi ion.

SHUTDOWN MARGIN racuiraaents vary throuchout, care 'life as a func ian

'f fuel depletion, RCS boron concentration, and RCS T The aust rastric ive condition occurs at EQL, with T at no f5d opemting mpera~iara, and is associa+A with a paschal 8 s~ line brmk accident and resulting uncontrolled RCS ccoldown.

In the analysis of His accident, a mininam SHUTDOWN MARGIN of 1.60wk/k is initially require( w contra]

)

the reac.ivity transient.

Accoraingly, the SHlliiMN MARGIN ~uirement is based upon this limiting condition and is ccrsis~t with FEAR ac"ident analysis assumptions.

With 7

<3:-0 F, the r activity ~nsiants r sul ing rom a postulated s~

m line br ak ccoldown are minimal and a

1 N/k shutdown margin provides adequate protection.

3/4.1.1.3 BORON DILUTION A minimum flow rato of at least 3000 GPM provides adequam

mixing, pr vents strati ication and ensures that reac.ivity c.'anges wi11 be gradual during boron canc ntration reduc ions in the Reactor Coolant System.

A flow rata of at least 3000 GPM will cir alata an euivalent Reactor Coolant System volume of 12,6i2 + 100 cubic fe t fn a oroximataly 30 minutes.

The reactivity change rate associa~

with boron rmtuctions will therefore be within the capability for operator recognition and control.

3/4.1.1.4 MODERATOR TEMPERATURE CDErr iCI VT MTC The limitations on MTC are provided to ensure that the assumptions used in the accident and transient analyses remain valid thr ugh each fuel cycle.

The surveillance requir ment for measurenent of We MTC at the beginning, and near the end of each fuel cycle is adecuata ta confirm the NTC value since iis coefficient changes slowly due D.

C.

COOK-UNIT 1 S 3/4 1-1 Amendr~t Ho.

74

3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1.4 MODERATOR TEMPERATURE COEFFICIENT MTC Continued principally to the reduction in RCS boron concentration associated with fuel burnup.

The confirmation that the measured and appropriately compensated MTC value is within the allowable tolerance of the predicted value provides additional assurances that the coefficient will be maintained within its limits during intervals between measurement.

3/4.1.1. 5 MINIMUMTEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541'F.

This limitation is required to ensure

1) the moderator temperature coefficient is within its analyzed temperature

range,

2) the protective instrumentation is within its normal operating

range, and 3) Tav is above the P-12 interlock setpoint.

3/4.1. 2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include 1) borated water sources,

2) charging

pumps,

3) separate flow paths,

4) boric acid transfer

pumps,

5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

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 boration capability of either system is sufficient to provide a

SHUTDOWN MARGIN from all operating conditions of 1.0X hk/k after xenon decay and cooldown to 200'F.

The maximum boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires 5106 gallons of 20,100 ppm borated water from the boric acid storage tanks or 52,622 gallons of 1950 ppm borated water from the refueling water storage tank.

D. C.

COOK-UNIT 1

B 3/4 1-2

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS FQ(Z) and F~H The limits on heat flux and nuclear enthalpy hot channel fac.ors ensure that i) 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 criteria limit.

Each of these hot channel factors are measurable, but will normally only be determined periodically, as specified in Specifications 4.2.2 and 4.2.3.

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

a.

Control rods in a single group move together with no individual rod insertion differing by more than

+ 12 steps from the group demand position.

b.

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

3. 1.3.5.

c.

The control rod insertion limits of Specifications

3. 1.3.4 and

3. 1.3.5 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE is maintained within the 1'.mits.

The relaxation in F

H as a function of THERMAL POWER al ows changes in the radial power shape for all permissible rod insertion limits.

F H will be maintained within its limits, provided conditions (a) a,H through (d) above are maintained.

When an F

measurement is taken both experimental error and manu-Q t

facturing tolerance must be allowed for.

SX is the appropria e allowance for a full core map taken with the incore detector flux mapping sys em, and 3'

s the appropr iate al 1 owance for manufacturing to 1 erance.

When F

H is measured, experimental error must be allowed for, N

aH and 4;'is the appropriate allowance for a full core map taken with the incore detection system.

This 4-measurement uncertainty has been included in the design DNBR limit value.

The specified limit for F

H also contains an additional 4,~ allowance for uncertainties.

The total allowance is based on the following considerations:

D.C.

Cook Unit 1

B 3/4 2-4 Amendment No. "4

POWER OISTRIBUTION LIMITS BASES a.

abnormal perturbations in the radial power shape, such as from rod misalignment, affect F

more directly than F

b.

although rod movement has a direct influence upon limi.ing F~ to within its limit, such control.is not readily a:ailable to limit F<H, and ting axial flux dist". ibutions.

is less readily available.

c.

errors in predic ion for control power shape detected during startup physics tests can be compensated for in F

, by restric-This compensation for r H

A burnup dependent F~ is specified as a result of the ECCS evalua-tion, in accordance with 10 CFR Part 50 Appendix K and to meet the accep" tance criteria of 10 CFR 50.46.

The basis for this dependence is given in document XN"76"51, Supplements 1, 2, 3, and 4 for Exxon fuels and the exemption granted by the Commission on May 18, 1978 for Westinghouse fuel.

3/4.2.4 OUAORANT POWER TILT RATIO The quadrants, power tilt ratio limit assures that he radial power distribution satisfies the design values used in the power capability analysis.

Rad'.al power distribution measurements are made during szartup tes ing and periodically during power operation.

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

A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in F~ is depleted.

The limit of 1.02 was selected to provide an allowance for the uncer ainty associated with the indicated power tilt.

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 for each percent of tilt in exc ss of 1.0.

O.C.

Cook Unit 1 B 3/4 2-5 Amendment No.

74

POWER DISTRIBUTION LIMITS BASES 3/4.2.5 ONB PARAMETERS The limits on the DNB related parameters assure that each of the para-meters are maintained within the normal steady state envelope of operation assumed in the transien and accident analyses.

The limits are consistent with the initial FSAR assumotions and have been analytically demonstrated to be adequate to maintain the applicable design limit ONBR values for each fuel type (which are listed in the bases for Section 2.1. l) throughout each analyzed transient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instru-ment readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The monthly periodic RCS elbow tap flow measurement is adequate to detect flow degradation and to ensure the correlation of the flow indica ion channels with measured flow, as determined at the beginning of each cycle using a power balance around the steam genera:ors, such -that the indicated percent flow will provide sufficient verification of flow rate on a

12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

Measurement uncertainties have been accounted for in determining the ONB parameters limit values.

3/4.2.6 AXIAL POWER DISTRIBUTION The limit on axial power distribution ensures that F will be con-Q

'rolled and monitored on a more exact basis through use of the APDMS wnen operating above APL of RATED THERMAL POWER.

This additional limitation on F

is necessary, in order to provide assurance tha peak clad tern'oera-tures will remain below the ECCS acceptance criteria limit of 2200'F in the event of a LOCA.

The unit may operate with fuel assembl',es supplied by the Exxon Nuclear Company and by Westinghouse Electric Corporation.

An F

limit has been specified for each of these two fuel types.

D.C.

Cook Unit I 8 3/4 2-6 Amendment No.

74

INSTRUt1ENTATION BASES 3/4.3.3.6 AXIAL POWER DISTRIBUTION MONITORING SYSTEM APDMS The OPERABILITf of the APDMS ensures that su ficient c pability is available for the measurement of the neutron flux spa ial distribution within the reactor core.

This capability is required to 1) monitor the core f'ux pat.erns that ar representative of the power p aking factor in the limiting fuel rod.

The limiting fuel rod is the fuel rod that has the least margin to the exposure dependent F limit curve, and 2) limit th cor average axial power profile such Chat the total po~er peaking fac.or F~ in the limiting fuel rod is maintained within accept-able limits.

R, factors ar used to determine the APDMS setpoint limits

f. >(<'}3S.

On a full core basis the R, and o, factors are calculated in accordance with the equations on Piges 3/k 2-18 and 3/4 2-'l9 However, near BOC, thimbles not in the region of fuel which contains the limiting total peaking factor, F~.

, may not follow he axial power distribution of the hot rod.

This ss.uation will manifest itself in he orm of large cr-for thimbles noi in the same region as the total peak F

In this Situation, i, the rod with the limi ing total pe king

=M".or were to move from one fuel region to another,

he neutr"n flux in "he tnimble wi h the smallest; a

~ would not nec ssarily'ollow t.".e axial power distribution o, the power in the n w limiting rod.

In order to cop with this di=f',culty, it is permissible to calculate as many a 's and R 's for each thimble as there are fu 1 types or regions iit the cor~.

Each K and a

for a thimble j is to be calculated from the equations on Pages )/4 2-11t and 2/4 2-19 with the following exception.

For each'K.

and a~ for thimble j, a different F.

and T(E) shall be used.

The diffpg ~t o 's and R 's for thimble j s5kfl be calcu-lated substituting for F'~i an3 T(E) thk values per'.sining to he limiting peak relative power from each fuel region.

Obviously for one of these calculations the limiting peak relative po~er from one region wi11 be the cor 1:miting total peaking factor.

If this option is chosen, the a; se. to use for APDMS thimble selec-tion and the R> set to use for the calculation of LF~(Z}])S shall be the set obtained usinc the limiting peak relative power vrom the same fuel type as the Fqi from th mos recent incore flux map.

D. C.

COOK - Ui'tIT I 8 3/4 3-3 Amend ent No.

74

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4. I REACTOR COOLANT LOOPS The plant is designed to operate with all reactor coolant loops fn operation, and maintain ONBR above the applicable design limit DNBR value during all normal operations and anticfpated transients.

With ~~e reactor coolant loop not in operation, THERMAL POWER is resgrfcted to <

51 percent of RATED THER:NL POWER, until the Overtemperature 4T trip is reset.

Either action ensures that the ONBR wi 11 be maintained above the applicable design limi. ONBR values for each fuel type.

A loss of flow in two loops will cause a reactor trsp if operating above P"7 (ll percent of RATED THERMAL PCWER) while a loss of flow in one loop will cause a reactor trip if operating above P-8 (51 percent of RATED THERMAL POWER).

A single reactor coolant. loop provides sufficient heat removal capa" bilfty for removing core decay heat while in HOT STANDBY; however, single failure considerations require placing an RHR looo into operation in the shutdown cooling mode if comoonent repairs and/or corrective cannot be made within the allowable out-of-service time.

The restrictions on starting a Reac or Coolant Pump below P-7 with one or-more-8CS '@rid 'legs less than or equal to 188'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 1G CFR Part 50.

The RCS will be orotected against overpressure transients and will not exceed the limits of Apoendix G by either (1) restric ing the water vorume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restric.ing starting of the RCPs to when the second-ary water temperature of each steam generator is less than 50 F above each of the RCS cold leg temperatures.

3/4.4.2 and 3/4.<.3 SAUCY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per hour of. saturated steam at the valve setpoint.

The relief capacity of a single safety valve is adequate to relieve any over pressure conditions which could occur during shutdown.

In the event that no safety valves are

OPERABLE, an operating RHR loop, con-nected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.

O.C.

Cook Unit 1

8 3/4 4-1 Amendment No. 74