Text

Amend 44 to License NPF-30,revising Tech Specs to Support Cycle 4 Core Reload.Revs Include Increased Peaking Factors, Positive Moderator Temp Coefficient & Increased Refueling Water Storage Tank & Accumulator Boron Concentrations
	ML20245D751
Person / Time
Site:	Callaway
Issue date:	04/19/1989
From:	Hannon J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20245D754	List: ML20245D751; ML20245D761;
References
	NUDOCS 8905010150
	Download: ML20245D751 (38)
	v • d • e

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UNITED STATES I

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NUCLEAR REGULATORY COMMISSION I

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,j WASHINGTON, D. C. 20655 p

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i UNION ELECTRIC COMPANY CALLAWAY PLANT, UNIT 1 q

f DOCKET NO. STN 50-483 i

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.44 License No. NPF-30 1.

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

A.

The application for amendment filed by Union Electric Company (UE,thelicensee)datedOctober 25, 1988 complies with the standards and requirements of.the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

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

Thereisreasonableassurance(1)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in complianc' with the Comission'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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment, and para-graph 2.C.(2) of Facility Operating License No. NPF-30 is hereby amended to read as follows:

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2 (2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendmant No.44, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into.the license. UE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

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

FOR THE NUCLEAR REGULATORY COMMISSION John N.Hannon, Director Project Directorate 111-3 Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 19, 1989 9

+

ATTACHMENT TO LICENSE AMENDMENT NO. 44 OPERATING LICENSE NO. NPF-30 DOCKET NO. 50-483 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Corresponding overleaf pages are provided to maintain document completeness.

REMOVE INSERT B 2-1 8 2-1 B 2-2 B 2-2 3/4 1-4 3/4 1-4 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12 3/4 2-1 3/4 2-1 3/42-2(a) 3/42-2(a) 3/4 2-4 3/4 2-4 3/4 2 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/42-7(a) 3/42-7(a) 3/4 2-7(b) 3/4 2-7(b) 3/4 2-8 3/4 2-8 3/4 5-1 3/4 5-1 3/4 5 3/4 5-10 3/4 6-14 3/4 6-14 B 3/4 1-2 B 3/4 1-2 B 3/4 1-3 B 3/4 1-3 B 3/4 2-1 B 3/4 2-1 B 3/4 2-4 8 3/4 2-4 B 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 B 3/4 5-2 B 3/4 5-2

1

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A 2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the 'uel and possible cladding perforation which would result in the release of f1 sion products to the reactor coolant.

Overheating of the fuel cladding is prevented 4

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 from nucleate boiling (DNB)g temperatures because of the onset of departure result in excessive claddin i

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 Tempe ature and Pressure have been related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNB heat flux ratio (DNBR) defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indic-ative 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-1 correlation for Optimized fuel (0FA) and the WRB-2 correlation for VANTAGE 5 fuel in this application). The correlation DNBR limit is estab-lished based on the entire applic.able 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 both the WRB-1 and WRB-2 correlations).

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

The uncertainties in the above plant parameters are used to determine the plant DNBR uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analyses using values of input parameters without uncertainties.

For Callaway, the design DNBR values are 1.33 and 1.35 for thimble and typical cells, respectively, for 0FA, and 1.33 and 1.34 for thimble and typical cells, respectively, for VANTAGE 5 fuel.

In addition, margin has been maintained in both fuel designs by meeting safety analysis DNBR limits of 1.42 and 1.45 for thimble and typical cells, respectively, for 0FA, and 1.61 and 1.69 for thimble and typical cells, respectively, for VANTAGE 5 fuel.

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

CALLAWAY - UNIT 1 B 2-1 Amendment No. JE, 28,44

SAFETY LIMITS.

1 BASES I

2.1.1 REACTOR CORE (Continued) l The curves are based on a measured nuclear enthalpy rise hot channel factor', F f 1.49 for 0FA and 1.59 for VANTAGE 5 fuel, and a reference aH, cosine with.a peak of 1.55 for axial power shape. An allowance is included for an increase in F"H at reduced power based on the expressions:

1 FAH = 1.49 [1+ 0.3 (1-P)] for 0FA, and F

= 1.59 [1+ 0.3 (1-P)] for VANTAGE 5 fuel H

where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f3 (AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtem-perature AT trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety limit protects the integrity of the Reactor Coolant System (RCS) from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor vessel, pressurizer, and the RCS piping and valves are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

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

The entire RCS is hydrotested at greater than or equal to 125% (3110'psig) of design pressure to demonstrate integrity prior to initial operation.

CALLAWAY - UNIT 1 B 2-2 Amendment No. JE, 44

REACTIVITY CONTROL SYSTEMS SHUTDOW MARG?% - T,yg 5 200*F

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LIMITING CONDITION FOR OPERATION l to 1% Ak/k.

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The SHUTDOWN MARGIN shall be greater than or e

[(

C 3.1.1.2 I[i MODE 5.

APPLICABILITY:

t lC immediately initiate and continue

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ACTION:

f a solution containing greater

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With the SHUT 00WN MARGIN less than 1% Ak k, lC boration at greater than or equal to 30 gpm ot until th Q(

l than or equal to 7000 ppm boron or equiva en is restored.

Z SURVEILLANCE REQUIREMENTS ater than or equal The SHUTOOWN MARGIN shall be determined to b 4.1.1.2 ntrol rod (s) and I to 1% Ak/k:

l Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperab e coile at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter whtrippable, the If the inoperable control rod is immovable or u a.

ith an increased SHUTDOWN MARGIN shall be verified acceptable allowance for the withdrawn worth of the i

control rod (s); and f the following fac !

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by consideration o Reactor Coolant System boron concentration, b.

1)

Control rod position, 2)

Reactor Coolant System average temperature,

tion, 3)

Fuel burnup based on gross thermal energy g

[pgg 4)

C(

Eenon concentration, and (7 g 5)

Ii Samarium concentration.

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9 3/4 1-3 CALLAWAY - UNIT 1

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REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1. 3 The moderator temperature coefficient (MTC) shall be:

a.

Less positive-than +5 pcm/ F for power levels up to 70% RATED THERMAL POWER and a linear ramp from that point to 0'pcm/ F at 100% RATED THERMAL POWER for the all rods withdrawn, beginning of cycle life (B0L) condition; and b.

Less. negative than -4.1 x 10-4 Ak/k/*F. for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.

APPLICABILITY:

Specification 3.1.1.3a. - MODES 1 and 2*#.

Specification 3.1.1.3b. - MODES 1, 2 and 3#.

ACTION:

a.

With.the MTC more positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established and maintained sufficient to restore the MTC to within the above limits l

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to tne insertion limits of Specification 3.1.3.6; 2.

The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition; and 3.

A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits, and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

l b.

With the MTC more negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CALLAWAY - UNIT 1 3/4 1-4 Amendment No. 44 l

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REACTIVITY CONTROL SYSTEMS B0 RATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR.0PCRATION 3.1.2.5 As a minimum, one of the following borated water sources shal1 be OPERABLE; a.

A BoHc Acid Storage System with:

1)

'A minimum contained borated water volume of 2968 gallons, 2)

Between 7000 and 7700 ppm of boron, and 3)

A minimum solution temperature of 65 F, b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 55,416 gallons, 2)

A minimum boron concentration of 2350 ppm, and l

3)

A minimum solution tt..perature of 37 F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source GPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

l SURVEILLANCE REQUIREMENTS 4.1.'2. 5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Verifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when it-is the source of borated water and the outside air temperature is less than 37 F.

CALLAWAY - UNIT 1 3/4 1-11 Amendment No. 44

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REACTIVITY' CONTROL SYSTEMS 1

B0 RATED WATER SOURCES - OPERATING LIMITING CONDITION FOR OPERATION 3.1.2.6 As a minimum -the following borated water sources shall be OPERABLE as required by Specification 3.1.2.2 for MODES 1, 2 and 3 and one of the following borated water sources shall be OPERABLE as required by Specifica-tion 3.1.2.1 for MODE 4:

a.

A Boric Acid Storage System with:

1)

A minimum contained borated water volume of 17,658 gallons, 2)

Between 7000 ~and 7700 ppm of boron, a'nd 3)

A minimum solution temperature of 65 F.

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 394,000 gallons, 2)-

Between 2350 and 2500 ppm of boron, l

3)

A minimum solution temperature of 37 F, and 4)

A maximum solution temperature of 100 F.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTION:

a.

With the Boric Acid Storage System inoperable and being used as one of the above required boratei water sources in MODE 1, 2, or 3, 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% ak/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 RWST inoperable in MODE 1, 2, or 3, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

c.

With no borated water source OPERABLE in MODE 4, restore one borated L

water source to OPERABLE status within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 l

J CALLAWAY - UNIT 1 3/4 1-12 Amendment No. 44 1

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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the following target band (flux difference units) about the target flux difference:

a.

+3%, -12% for Normal Operation b.

+3% for RESTRICTED AFD GPERATION deviate outside the applicabggequired targTED THERMAL The indicated AFD ma t band at to 50% but less than 0.9 APL or 90% of greater than or equa POWER, whichever is less, provided the indicated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the cumulative penalty deviation times does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD may deviate outside the applicable required target band at greater than 15% but less than 50% of RATED THERMAL POWER provided the cumula-tive penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

APPLICABILITY: MODE 1, above 15% of RATED THERMAL POWER *,#

ACTION:

a.

With the indicated AFD outside of the applicable required taraet band and with THERMAL POWER greater than or equal to 0.9 APLND**

or 90% of RATED THERMAL POWER, whichever is less, within 15 minutes, either:

1.

Restore the indicated AFD to within the applicable required target band limits, or

See SpecTal Test Exception Specification 3.10.2.

Surveillance testing of the Power Range Neutron Flux channel may be per-formed pursuant to Specification 4.3.1.1 provided the indicated AFD is maintained within tg, Acceptable Operation Limits of Figure 3,2-1 and THERMAL POWER <APLN A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation may be accumulated with the AFD outside of the applicable required target band during testing without penalty deviation.

- APLND is the minimum allowable power level for RESTRICTED AFD OPERATION and will be provided in the Peaking Factor Limit Report per Specification 6.9.1.9.

- - APLN0 is equal to the minimum

" 2.50 K(Z)

100 over Z

,F (Z)

W(2)N0 and F (Z) and W(Z)N0 are defined in 4.2.2.2.c.

CALLAWAY - UNIT 1 3/4 2-1 Amendment No. 2g/, 44 l

-_-_________-__-_____-______________n

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1 POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION ACTION (Continued) 2.

Reduce THERMAL POWER to less than 0.9 APLND** or 90% of RATED THERMAL POWER, whichever is less, and discontinue RESTRICTED AFD OPERATION (if applicable).

b.

With the indicated AFD outside of the applicable. required target i

band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during ths previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL POWER less than 0.9 APLND**

or 90%, whichever is less, but equal to or greater than 50% of. RATED THERMAL POWER, reduce:

1.

THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes, and 2.

The Power Range Neutron Flux-High Setpoints to less than or a

equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

c.

With the indicated AFD outside of the applicable required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated AFD is within the applicable required target band.

SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AFD 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:

1.

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, b.

Monitoring and logging the indicated AFD 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 AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed to exist during the interval preceding each logging.

I CA%)' LAY - UNIT 1 3/4 2-2 Amendment No. 28

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.2 The indicated AFD shall be considered outside of its target band when

' two or more OPERABLE excore channels are indicating the AFD to be outside the target band.

Penalty deviation outside of the above required target band 1

shall be accumulated on a time basis of:

a.

One minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels equal to or above

-50% of RATED THERMAL POWER, and

)

b.

One-half minute penalty deviation for each 1 minute of POWER OPERA-TION-outside of the target band at THERMAL POWER levels between 15%-

and 50% of RATED THERMAL POWER.

4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days. The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effec-tive Full Power Days by either determining the target flux difference pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and the calculated value at the end of the cycle life.

l The provisions of Specification 4.0.4 are not applicable.

l l

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CALLAWAY - UNIT 1 3/4 2-2(a)

Amendment No. /p, 44

i FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER 1

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i CALLAWAY - UNIT 1 3/4 2-3 l

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a POWER DISTRIBUTION LIMITS 3/4.2.2: HEAT FLUX HOT CHANNEL FACTOR - F (Z) n LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall.be limited by the following relationships:

9 F (Z) 1 [2.50] [K(Z)] for P > 0.5, arid l

n P

F (Z) 1 [5.00] [K(Z)] for p 1 0. 5.

l g

Where:

p _ THERMAL-POWER

, and RATED THERMAL POWER K(Z) = the' function obtained from Figure 3.2-2 for a given core height location.

APPLICABILITY: ' MODE 1.

ACTION:

With F (Z) exceeding its limit:

g a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit 9

within 15 minutes and similarly reduce the Power Range Neutron Flux-

'High Trip Setpoints within.the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; subsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpointc have been reduced at least 1% for each.1% F (Z) exceeds the limit; and g

b.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by ACTION a.

above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within its limit.

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l ll' l-CALLAWAY - UNIT 1 3/4 2-4 Amendment No. 44 f

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.sc84 carnynwoN-(IIM CALLAWAY - UNIT 1 3/4 2-5 Amendment No. 44

u POWER DISTRIBUTION LIMITS' SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 For Normal Operation, Fg(z) shall be evaluated to determine if F (2)

Q is within'its ' limit by:

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

.b.

Increasing the measured Fg(z) component of the power distribution map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c.

Satisfying-the following relationship:

Fg (z) s'2.50 x K(z) for P > 0.5 l

M P x W(z)N0-M Fg (z) s'2.50 x K(z) for P 1 0.5 W(z)N0 x 0.5 manufac$(z)isthemeasuredF(z)increasedbytheallowancesfor where F Q

turing tolerances and measurement uncertainty, 2.50 is the l

F0 limit, K(z) is given in Figure 3.2-2, P is the relative THERMAL POWER,andW(z)NO is the cycle dependent, Normal Operation function that accounts for power distribution transients encountered during Normal Operation. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.

M d.

Measuring FQ (z) according to the following schedule:

1.

Upon achieving equilibrium conditions after exceeding, by 10%

or more of RATED THERMAL POWER, the THERMAL POWER at which F (z).was last determined,* or Q

2.

At least once per 31 Effective Full Power Days (EFPD), whichever occurs first.

l

During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.

i i

CALLAWAY - UNIT 1 3/4 2-6 Amendment No. P#, 44

f l

1 POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 '(Continued),

~

e.

With measurements indicating Fh(z)

-maximum over z K(z)

M l

hasincreasedsincethepreviousdeterminationofFg(z),either of the following actions shall be taken:

M Fg (z) shall be increased by 2% over that specified in 1.

Specification 4.2.2.2c.. or M

FQ (z) shall be measured at least once per 7 Effective Full 2.

Power Days until two successive maps indicate that h(z) maximum is not increasing.

over z K(z) f.

With the relationships specified in Specification 4.2.'2.2c. above not being satisfied:

1.

Calculate the percent Fg(z) exceeds its limit by the following 4

-expression:

(max. over z of Fh(z) x W(z)N0 )-1 x 100 for P > 0. 5 K(z) l 2.50 v p

(max. over z of Fh(z)x W(z)N0

)-1 x 100 for. P < d.5 2.50 x K(z) 0.5 2.

Either one of the following actions shall be taken:

(a) Comply with the requirements of Specification 3.2.2 for Fo(z) exceeding its limit by the percent calculated aBove, or (b) Verify that the requirements of Specification 4.2.2.3 for RESTRICTED AFD OPERATION are satisfied and enter RESTRICTED AFD OPERATION.

g.

The limits specified in Specifications 4.2.2.2.c., 4.2.2.2.e., and 4.2.2.2.f. above are not applicable.in the following core plane regions:

1.

Lower core region from 0 to 15%, inclusive.

2.

Upper core region from 85 to 100%, inclusive.

CALLAWAY - UNIT 1 3/4 2-7 Amendment No. ?#, 44

y l

I POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.3 RESTRICTED AFD OPERATION (RAFDO) is permitted at powers above APLND if the following conditions are satisfied

a.

Prior to entering RAFD0, maintain THERMAL POWER above APLND and less than or equal to that allowed by Specification 4.2.2.2 for at least the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Maintain RAFD0 surveillance (AFD within +3%

1 of target flux difference) during this time period. RAFD0 is then l

APLgd providing THERMAL POWER is maintained between APLND and per or between APLND and 100% (whichever is more limiting) and A L g illance is maintained' pursuant to Specification 4.2.2.4.

F s is defined as:

APLRAF00 = minimum '2.50 x K(z) x 100%

l

~

M(z)xW(z)kAFD0 over z

,p manufactu(ring)tolerancesandmeasurementunc where:

F (z) is the measured Fg(z) increased by the allowances for The FQ limit

]

is 2.50 K(z is given in Figure 3.2-2.

W(z)RAFD0 is the cycle l-dependent function that accounts for limited power distribution transients encountered during RAFD0. This function is given in the q

-Peaking Factor Limit Report as per Specification 6.9.1.9.

1l b.

During RAFD0, if the THERMAL POWER is decreased below APLND then the conditions of 4.2.2.3.a shall be satisfied before re-entering RAFDO.

During RAFD0, F (z) shall be evaluated to determine if F (z) is

'4.2.2.4 Q

Q within its limits by:

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER above APLND, Increasing the measured F (z) component of the power distribution b.

Q map by 3% to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties, c.

Satisfying the following relationship:

ND F (z) < 2.50 x K(z) for P > APL

_ P x W(z)RAFD0 Fh(z)isthemeasuredF(z). The Fg limit is 2.50.

K(z l

where:

Q is given in Figure 3.2-~2.

P is the relative THERMAL POWER. W(z RAFD0 is the cycle dependent function that accounts for limited power distribution transients encountered during RAFDO. This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.9.

I f

1 I

I CALLAWAY - UNIT 1 3/4 2-7(a)

Amendment No. /, 44 L____-___________-____.

l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.4 (Continued) d.

MeasuringF$(z)inconjunctionwithtargetfluxdifferencedetermi-nation according to the following schedule:

1.

Prior to entering RAFD0 af ter satisfying Section 4.2.2.3 unless a full core flux map has been taken in the previous 31 EFPD with the relative thermal power having been maintained above APLND for the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to mapping, and 2.

At least once per 31. Effective Full Power Days.

e.

With measurements indicating Fh(z) maximum over z

,K(z).

hasincreasedsincethepreviousdeterminationofFh(z)eitherof the following actions shall be.taken:

Fh(z) shall be increased by 2 percent over that specified in 1.

4.2.2.4.c, or 2.

F$(z)shallbemeasuredatleastonceper7EFPDuntiltwo s0ccessive maps indicate that

'h(z)~isnotincreasing, maximum F

over z

.K(z) f.

With the relationship specified in 4.2.2.4.c above not being satisfied, comply with the requirements of-Specification 3.2.2 for F (z) exceeding its limit by the percent calculated with the Q

following expression:

_(max.overzof[Fh(z)xW(z)RAFD0 )

-1 x 100 for P > APLND 2.50 x K(z) kP

)

g.

The limits specified in 4.2.2.4.c, 4.2.2.4.e, and 4.2.2.4.f above are not applicable in the following core plane regions:

1.

Lower core region from 0 to 15 percent, inclusive.

)

l 2.

Upper core region from 85 to 100 percent, inclusive.

4.2.2.5 When Fo(z) is measured for reasons other than meeting the require-

)

ments of Specification 4.2.2.2 or 4.2.2.4, an overall measured F (z) shall be Q

obtained from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measure-ment uncertainty.

CALLAWAY - UNIT 1 3/4 2-7(b)

Amendment No. /M 44 i

1 POWER DISTRIBUTION LIMITS NUCLEARENTHALPYRISEHOTCHANNELFACTOR-Fh 3/4.2.3 LIMITING CONDITION FOR OPERATION 3.2.3 F shall be limited by the following relationship:

H F

i 1.59 [1 + 0.3 (1-P)]

H where P = THERMAL POWER RATED THERMAL POWER F H = Measured values of F H obtained by using the movable incore detectors to obtain a power distribution map.

The measured values o'f Fh shall be used since an uncertainty of 4% for incore measurement of Fh has been included in the above limi APPLICABILITY:

MODE 1 ACTION:

WithFhexceedingitslimit:

a.

Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

RestoretheFlH to within the above limits, or 1.

2.

Reduce THERMAL POWER TO LESS THAN 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux-High Trip Setpoint to i 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Demonstrate through in-core flux mapping that Fh is within b.

its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and c.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may pro-reed provided that Fh is demonstrated through in-core flux mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

CALLAWAY - UNIT 1 3/4 2-8 Amendment No. /l/7,44

(-

f.

L l

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4. 5.1 ACCUMULATORS-LIMITING CONDITION FOR OPERATION 3.5.1 Each Reactor Coolant System accumulator shall be OPERABLE with:

a.

The isolation valve open and power removed, b.

~A contained borated water volume of between 6061 and 6655 gallons, c.

A boron concentration of between 2300 and 2500 ppm, arid l

d.

A nitrogen cover-pressure of between 602 and 648 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 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 reduce RCS pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one accumulator inoperabis due to the ' isolation valve being closed, either immediately open the isolation valve or be in at least H0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce RCS pressure to less'than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.1.1 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, by the absence of alarms, the contained borated water volume and nitrogen cover-pressure in the tanks, and 2)

Verifying that each accumulator isolation valve is open.

RCS pressure above 1000,psig.

One accumulator isolation valve may be closed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in mode 3* for surveillance testing per 4.0.5 or 4.4.6.2.2.

CALLAWAY - UNIT 1 3/4 5-1 Amendment No. 2E,/4, 44

l 1

j j

EMERGENCY CORE COOLING SYSTEMS j

1 SURVEILLANCE REQUIREMENTS (Continued) b.

At least once per 31 days and within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution j

l volume increase of greater than or equal to 70 gallons by verifying the boron concentration of the accumulator solution; and c.

At least once per 31 days when the RCS pressure is above 1000 psig

' by verifying that the circuit breaker supplying power to the isola-tion valve operator is open.

4.5.1.2 Each accumulator water level and pressure channel shall be demonstr-ated OPERABLE at least once per 18 months by the performance of a CHANNEL L

CALIBRATION.

l l

CALLAWAY - UNIT 1 3/4 5-2 e_________

EMERGENCY CORE COOLING SYSTEMS 3/4.5.4 ECCS SUBSYSTEMS - T

200*F avg -

LIMITING CONDITION FOR OPERATION l

3.5.4 All Safety Injection pumps shall be inoperable.

APPLICABILITY: MODE-5 with the water level above the top of the reactor vessel flange, and MODE 6 with the reactor vessel head on and with the water level above the top of the reactor vessel flange.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.4 All Safety Injection pumps shall be demonstrated inoperable

by verifying that the motor circuit breakers are secured in the open position at least once per 31 days.

)

1

An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

CALLAWAY-UNIT 1 3/4 5-9 Amendment No. 42

EMERGENCY CORE COOLING SYSTEMS 3/4.5.5 REFUELING WATER STORAGE TANK LIMITING CONDITION FOR OPERATION 3.5.5 The refueling water storage tank (RWST) shall be OPERABLE with:

a.

A minimum contained borated water volume of 394,000 gallons, b.

A boron concentration of betwee 2350 and 2500 ppm of boron, l

c.

A minimum solution temperature of 37 F, and d.

A maximum solution temperature of 100*F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION.

With the RWST inoperable,. restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least H0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS

'4.5.5. The RWST shall'be demonstrated OPERABLE:

a.

At least once per 7 days ty:

1)

Verifying the contained borated water volume in the tank, and 2)

Verifying the boron concentration of the water, b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the outside air temperature is either less than 37*F or greater than 100 F.

CALLAWAY - UNIT 1 3/4 5-10 Amendment No. 44 1

9 CONTAINMENT SYSTEMS 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM LIMITING CONDITION FOR OPERATION 3.6.2.1 Two independent Containment Spray Systems shall be OPERABLE with each Containment Spray System capable of taking suction from the RWST and transferring suction to the containment sump.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With one Containment Spray System inoperable, restore the inoperable Containment Spray 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 ; restore the ir. operable Containment Spray System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.2.1 ' Each Containment Spray System shall be demonstrated OPERABLE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; b.

By verifying, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 250 psig when tested

.~

pursuant to Specification 4.0.5; c.

At least once per 18 months during shutdown, by:

1)

Verifying that each automatic valve in the flow path actuates j

to its correct position on a Containment Pressure-High-3 (CSAS) test signal, and 2)# Verifying that each spray pump starts automatically on a l

Containment Pressure-High-3 (CSAS) test signal.

(

)

d.

At least once per 5 years by perfoming an air or smoke flow test through each spray header and verifying each spray nozzle is unobstructed.

)

i

The specified 18 month frequency may be waived for Cycle I provided the l

surveillance is performed prior to restart following the first refueling outage or June 1,1986, whichever occurs first. The provisions of Specification 4.0.2 are reset from performance of this surveillance.

CAllAWAY - UNIT 1 3/4 6-13 Amendment No. 8 l

o

CONTAINMENT SYSTEMS SPRAY ADDITIVE SYSTEM LIMITING CONDITION FOR OPERATION i

3.6.2.2 The Spray Additive System shall be OPERABLE with:

a.

A spray additive tank containing a volume of between 4340 and 4540 gallons of between 31% and 34% by weight Na0H solution, l

and b.

Two spray additive eductors each capable of adding Na0H solution from the chemical additive tank to a Containment Spray System pump flow.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

With the Spray Additive System inoperable, restore the 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 H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore the Spray Additive System to OPERABLE status within the next 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.5.2.2 The Spray Additive System shall be demonstrated OPERA 3LE:

a.

At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position; b.

At least once per 6 months by.

1)

Verifying the contained solution volume in the tank, and 2)

Verifying the concentration of the Na0H solution by chemical

analysis, c.

At least once per 18 months during shutdown, by verifying that each automatic valve in the flow path actuates to its correct position l

on a Containment Pressure-High-3 (CSAS) test signal; and d.

At least once per 5 years by verifying 1)

Each eductor flow rate is greater than or equal to 52 gpm using the RWST as the test source throttled to 17 psig at the eductor inlet, and 2)

The lines between the spray additive tank and the eductors are not blocked by verifying flow.

CALLAWAY - UNIT 1 3/4 6-14 Amendment No. 44 i

i L

l-1 3/4.1 RtACIIV11Y CONIROL SYSTEMS BASFS 3/4.1.1 BORAT10N 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 subtritical 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 subtritical to preclude inadvertent criticality in the shutdown condition.

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

The most restrictive condition occurs at EOL, with T,yg at no load operating temperature, and is associated with a postulated steam'line break accident and resulting uncon-trolled RCS cooldown.

In the analysis of this accident, a minimum SHUTOOWN MARGIN of 1.3% ok/k is required to control the reactivity transient.

Accordingly, the SHUTDOWN MARGIN requirement is based upon this limiting condition and is consistent with FSAR safety analysis assumptions. With T,yg less than 200 F, the reactivity transients resulting from a postulated steam line break cooldown are minimal and a 1% Ak/k SHUTDOWN MARGIN provides adequate protection.

3 /4.1.1. 3 MODERATOR TEMPERATURE COEFFICIENT Ihe limitations on moderator temperature coefficient (MTC) are provided to ensure that the value of this coefficient remains within the limiting conclition assumed in the FSAR accident and trangient analyses.

lhe MIC values of this specification are applicable to a specific set of plant conditions; accordingly, verification of MlC values at conditions other than those explicitly stated will require extrapolation to those conditions in order to permit an accurate comparison.

CAllAWAY - (INil i B 3/4 1-1

\\

1 REACTIVITY CONTROL SYSTEMS BASES I

MODERATOR TEMPERATURE COEFFICIENT-(Continued)

L i

The most negative MTC value equivalent to the most positive moderator

{

l density coefficient (MDC), was obtained by incrementally correcting the MDC i

used in the FSAR analyses to nominal operating conditions. These corrections l

involved subtracting the incremental. change in the MDC associated with a core condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting MTC value -4.1 x 10-4 ok/k/ F.

The MTC value of -3.2 x 10-4 Ak/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.1 x 10-4 Ak/k/ F.

The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551 F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature range (2) the trip instrumentation is within its normal operating range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RT temperature.

NDT 3/4.1.2 B0 RATION SYSTEMS The Boration Systems ensure 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) centrifugal charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, and (5) an e"iergency power supply from OPERABLE diesel generators.

With the RCS average temperature equal to or greater than 350 F, a minimum I

of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The Boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% ak/k after xenon decay and cooldown to 200 F.

The maximum expected boration capability require-ment occurs at E0L from full power equilibrium xenon conditions and requires 17,658 gallons of 7000 ppm borated water from the boric acid storage tanks or 83,745 gallons of 2350 ppm borated water from the RWST.

With the RCS average l

temperature less than 350 F, only one boron injection flow path is required.

CALLAWAY - UNIT 1 B 3/4 1-2 Amendment No. 44

i REACTIVITY CONTROL SYSTEMS BASIS 1

B0 RATION SYSTEMS (Continued)

With the RCS temperature below 200 F, one Boration 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 changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable in MODES 4, 5, and 6 provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV or an RHR suction relief valve.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% ak/k after xenon decay and cooldown from 200 F to 140 F.

This condition requires either 2968 gallons of 7000 ppm borated water from the boric acid storage tanks or 14,076 gallons of 2350 ppm borated water from the l

RWST.

The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 11.0 for the solution recirculated within Containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes th( effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one Boration System during REFUELING ensures that 'this system is available for reactivity control while in MODE 6.

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 main-tained, and (3) the potential effects of rod misalignment on associated acci-dent analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position Indicator agrees with the demanded position within i12 steps at 24, 48,120 and 228 steps withdrawn for the Control Banks and 18, 210 and 228 steps withdrawn for the Shutdown Banks provides assurance that the Digital Rod Position Indicator is operating correctly over the full range of indication. Since the Digital Rod Position System does not indicate the actual shutdown rod position between 18 steps and 210 steps, only points in the indi-cated ranges are picked for verification of agreement with demanded position.

Shutdown and control rods are positioned at 225 steps or higher for fully with-drawn.

CALLAWAY - UNIT 1 B 3/4 1-3 Amendment No. U, 44

I i

-REACTIVITY CONTROL SYSTEMS 1

BASES MOVABLE CONTROL ASSEMBLIES (Continued)

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 and a restriction in THERMAL POWER. These restrictions provide assurance of fuel rod. integrity during continued operation.

In addition, those safety 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 safety analyses. Measurement with T greater than or avg equal to 551 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

.LCOs are satisfied.

CALLAWAY - UNIT 1 B 3/4 1-4 L

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(1) maintaining the minimum DNBR in the core at or above the safety analysis DNBR limits during normal operation and in short-term transients, and (2) limiting the fission gas release, fuel pellet temperature, and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded.

The definition of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux 9

on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; and N

FaH Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power aloag the rod with the highest integrated power to the average rod power.

3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F envelopes of 2.32 and 2.50 for 0FA and VANTAGE 5, respectively, g(Z) upper bou times the normalized axial peaking factor are not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions. The full-length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady-state operatiNi at high power levels. The value of the target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value by the appropriate fractional THERMAL POWER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

The limits on AXIAL FLUX DIFFERENCE (AFD) are given in Specification 3.2.1.

Two modes of operation are permissible. One mode is Normal Operation, where the applicable AFD limit is defined by Specification 3.2.1.a.

The AFD limit for this mode of operation is a +3, -12% target band about the target flux difference.

After extended load following maneuvers, the AFD limits may result in restric-tions in the maximum allowed power to quarantee operation with Fg(Z) less than its limiting value. To prevent this occuwence, another operating mode which CALLAWAY - UNIT 1 B 3/4 2-1 Amendment No. TE, y,44

POWER DISTRIBUTION LIMITS BASE'.

3/4.2.1 AXIAL FLUX DIFFERENCE (Continued) restricts the AFD to a relatively small target band and does not allow signif-icant changes in power level has been defined. This mode is called RESTRICTED AFD OPERATION, which restricts the AFD to a +3% target band about the target flux difference and restricts power levels ti between APLND and either APLRAFD0 or 100% of RATED THERMAL POWER, whichever is less.

Prior to entering RESTRICTED AFD OPERATION, a 24-hour waiting period at a power level (t2%) above APLND and below that allowed by Normal Operation is necessary.

During this time period load changes and cor. trol rod motion are restricted to that allowed by the RESTRICTED AFD OPERATION procedure. After the waiting period, RESTRICTED AFD OPERATION is permitted.

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duration of tne deviation is limited. Accordingly, a 1-hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer deter-mines the 1 minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for two or more OPERABLE excore channels are outside the target band and the THERMAL POWER is greater than 90%

of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50%

and 90% and between 15% and 50% RATED THERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of i

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4.2-1 shows a typical monthly target band.

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 desi minimum DNBR are not exceeded, and 2) gn limits on peak local power density and in the event of a LOCA the peak fuel clad temperature will not exceed the 2200 F ECCS acceptance criteria limit.

j l

CALLAWAY - UNIT 1 B 3/4 2-2 Amendment No. JE, 28

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INDICATED AX1AL FLUX DIFFERENCE FIGURE B 3/4.2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE YERSUS THERMAL POWER CALLAWAY - UNIT 1 B 3/4 2-3 Amendment No. 15

9 POWER DISTRIBUTION LIMITS BASES l

3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued)

Each of these is measurable but will normally only be determined period-ically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveil-lance-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 + 12 steps, indicated, from the group demand position.

b.

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

c.

The control rod insertion limits of Specification 3.1.3.6 are maintained.

d.

The axial power distribution, expressed in tenns of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FlH will be maintained within its limits provided conditions a. through d.

N above are maintained. The relaxation of F as a function of THERMAL POWER allows changes in the radial power shape fh all permissible rod insertion limits.

When an F measurement is taken, an allowance for both experimental error q

and manufacturing tolerance must be made. An allowance of 5% is appropriate for I

a full-core map taken with the incore detector flux mapping system and a 3%

allowance is '.opropriate for manufacturing tolerance.

When F is measured (i.e., inferred), no additional allowances are H

necessary prior to comparison with the limits of Section 3.2.3.

An error allow-ance of 4% has been included in the limits of Section 3.2.3.

Specifications 3.2.2 and 3.2.3 contain the F and F-delta-H limits appli-q cable to VANTAGE 5 fuel. The OFA fuel is analyzed to lower limits since it will have experienced burnup, thereby reducing the attainable OFA-specific hot channel factors such that the expected peak power levels and peak radial power of the OFA fuel will be much less than that necessary to approach the 0FA Fg and F-delta-H analysis limits.

j Margin between the safety analysis DNBR limits (1.42 and 1.45 for the Optimized fuel thimble and typical cells, respectively, and 1.61 and 1.69 for the VANTAGE 5 thimble and typical cells) and the design DNBR limits (1.33 and 1.35 for the Optimized fuel thimble and typical cells and 1.33 and 1.34 for the

]

VANTAGE 5 thimble and typical cells, respectively) is maintained. A fraction of this margin is utilized to accommodate the transition core DNBR penalty l

l CALLAWAY - UNIT 1 B 3/4 2-4 Amendment No. JE 28,44 L__________--___--

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 (Continued)

(12 % for VANTAGE 5 fuel) and the appropriate fuel rod bow DNBR penalty (less I

than 1.5% per WCAP-8691, Rev. 1). The margin between design and safety analysis DNBR limits of 6.3% for Optimized fuel and 17.4% for VANTAGE 5 fuel includes greater than 3% margin for both Optimized fuel and VANTAGE 5 fuel for plant design flexibility.

The hot channel factor F (z) is measured periodically and increased by a cycle and helght dependent power factor appropriate to either Nonnal Operation or RESTRICTED AFD OPERATION, W(z)NO r W(z)RAF00, to provide assurance that the limit on the hot channel factor, F (z), is met. W(z)NO accounts for the g

effects of normal operation transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

W(z)RAFD0 accounts for the more restrictive operating limits required by RESTRICTED AFD OPERATION which result in less severe transient values.

The W(z) functions are provided in the Peaking fcetor Limit Report per Specifica-tion 6.9.1.9.

Provisions to account for the possibility of decreases in margin to the FQ(z) limit during intervals between surveillance are provided. Any decrease in the minimum margin to the Fg(z) limit compared to the minimum margin determined from the previous flux map is determined by comparing the ratio of:

maximum Fh(z) over z K(z) taken from the current map to the same ratio from the previous map. The ratios to be compared from the two flux maps do not need to be calculated at identical z locations.

Increases in this ratio indicate that the minim m margin to the F (z) limit has decreased and that additional penalties must be applied to the Q

measured F (z) to account for further decreases in margin that could occur Q

before the next surveillance. More frequent surveillance may also be substi-tuted for the additional penalty.

3/4.2.4 QUADRANT POWER TILT RATIO The QUADRANT POWER TILT RATIO limit assures that the radial power distri-bution 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. A l

l l

CALLAWAY - UNIT 1 B 3/4 2-5 Amendment No. U.N.44

POWER DISTR!BUTION LlMlTS BASES 3/4.2.4 QUADRANT POWER TILT RATIO (Continued) limit of 1.02 was selected to provide an allowa nce for the uncertainty associ-ated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correc-tion of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fn is reinstated by reducing the maximum allowed power by 3% for each percent of tYlt in excess of 1.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO. The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles. The two sets of four symet-ric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-ll. H-3, H-13, L-5, L-11, N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the param-eters is maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain the safety analysis DNBR limit throughout each analyzed transient. The indicated Tavg value of 592.6*F and the indicated pressurizer pressure value of 2220 psig correspond to analytical limits of 595.2'F and,

2202 psig respectively, with allowance for measurement uncertainty.

The 12-hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their liaits following load changes and other expected transient operation.

When RCS flow rate is measured, no additional allowances are necessary prior to comparison with the limits of Section 3.2.5.

A measurement uncer-tainty of 2.2% (including 0.1% for feedwater venturi fouling) for RCS total flow rate has been allowed for in determination of the design DNBR value.

The measurement uncertainty for the RCS total flow rate is based upon perform-ing a precision heat balance and using the result to normalize the RCS flow rate indicators.

Fotential fouling of the feedwater venturi which might not be detected could b as the result from the precision heat balance in a non-conservative manner. Therefore, an inspection is performed on the feedwater venturi each refueling outage.

CALLAWAY - UNIT 1 B 3/4 2-6 Amendment No. U,2E,44

e 3/4.5.EMERG'ENCY CORE COOLING SYSTEMS BASES 3/4. 5.1 ACCUMULATORS The OPERABILITY.of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core prcyides the initial cooling mechanism'during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure

.that the assumptions used for accumulator injection in the safety analysis are met.

The accumulator power operated isolation valves are considered to be

" operating bypasses" in the conte'xt of IEEE Std. -279-1971, which requires that bypasses of a protective function be removed automatically whe.never permissive conditions are not met.

In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required.

The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.

If a closed

-isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action-1: required to place the reactor in a MODE where this capability is not required.

In order to perform check valve surveillance testing per 4.0.5 or 4.4.6.2.2 above 1000 psig RCS pressure, one accumulator isolation valve may be closed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in mode 3 only.

The requirement to verify accumulator isolation valves shut with power removed from the valve operator when the pressurizer is solid ensures the accumulators will not inject water and cause a pressure transient when the Reactor Coolant Syster is on solid plant pressure control.

3/4.5.-2, 3/4.5.3, m d 3/4.5.4 ECCS SUBSYSTEi4S The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any single failure consideration.

Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended breek of the largest RCS cold leg pipe downward.

In addition, each ECCS subsystem provides long-term core cooling capability in the recircula-l tion mode during the accident recovery period.

With the RCS temperature below 350 F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the basis of the stable reactivity j

condition of the reactor and the limited core cooling requirements.

4 CALLAWAY - UNIT 1 B 3/4 5-1 Amendment No. 40 4

EMERGENCY CORE C0OLING SYSTEMS

_ BASES ECCS SUBLYSTEMS (Continued)

The limitation for a maximum of one centrifugal charging pump to be OPERABLE and the Surveillance Requirement to verify all charging pumps except the required OPERABLE charging pump to be inoperable in MODES 4 and 5 and in MODE 6 with the reactor vessel head on, provides assurance that a mass addi-tion pressure transient can be relieved by the-operation of a single PORV or RHR suction relief valve.

In addition, the requirement to verify all Safety Injection pumps to be inoperable in MODE 4, in MODE 5 with the water level above the top of the reactor vessel flange, and in MODE 6 with the reactor vessel head on and with the water level above the top of the reactor vessel flange, provides assurance that the mass addition can be relieved by a single PORV or RHR suction relief valve.

With the water level not above the top of the reactor vessel flange and with the vessel head on, Safety Injection pumps may be available to mitigate the effects of a loss of decay heat removai during partially drained conditions.

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

Surveillance Requirements for. throttle valve position stops and flow balance testing provide assurance that proper ECCS G ows will be maintained in the event of a LOCA. ilaintenance of p_ roper flow resistance and pressure drop in the piping system to each injec-tien point is necessary to:

(1) prevent total pump flow from exceeding runout conditions venen the system is in its minimum resistance configuration. (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. The Surveillance Requirements for leakage testing of ECCS check valves ensure that a failure of one valve will not cause an inter-system LOCA.

The Surveillance Requirement to vent the ECCS pump casings and accessible, i.e., can be reached without personnel hazard or high radiation dose, discharge piping ensures against inoperable pumps caused by gas binding or water hammer in ECCS piping.

3/4.5.5 REFUELING WATER STORAGE TANK The OPERABILITY of the refueling water storage tank (RWST) as part of the ECCS ensures that a sufficient supply of borated water is available for injec-tion by the ECCS in the event of a LOCA.

T'he limits on RWST minimum volume and boron concentration ensure that:

'l) sufficient water is available within containment to permit recirculation coclina flow to the core, and (2) the reactor will remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes assuming all the control rods are out of the core.

These assumptions are consistent with the LOCA analyses.

CALLAWAY - UNIT 1 B 3/4 5-2 Amendment No. A2, 44

_ _ _ _ _ - - _ __

ML20245D751

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