ML20235T995

Proposed marked-up Tech Specs,Reflecting Revised RCS Total Flow Rate

ML20235T995
Person / Time
Site:	Wolf Creek
Issue date:	10/02/1987
From:	WOLF CREEK NUCLEAR OPERATING CORP.
To:
Shared Package
ML20235T984	List: ML20235T981 ML20235T995
References
NUDOCS 8710130315
Download: ML20235T995 (12)

Text

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Attachment III to WM 87-0254 Page 7 of 11 October 2, 1987

~ ~'A POWER OISTRTBUTION L1MfTS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RC5) total flow rate and R shall be maintained within the region of allcwable cperation shown on Figure 3.2-3 for four loop operation.

Where:

N FAH R = 1.49 [1.0 + 0.2 (1.0 - P)]

l a*

b THERMAL POWER and b*

P =

RATED THERMAL POWER H

N c.

FaH.= Measured values of F btained by using the movaole incore aH detectors to obtain a power distribution mao.

The measured values of F shall be used to calculate R since Figure 3.2-3 H

includes measurement uncertainties of 2.0% for flow and 4%

l for incore measurement of F H' APPLICABILITY:

MODE 1.

ACTION:

With the combination of RCS total flow rate and R outside the region of acceptable operation shown on Figure 3.2-3:

a.

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

1.

Restore the combination of RCS total flow rate and R to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setooint to less than or ecual to 55% of RATED THERMAL POWER witnin the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comoarison that the combination of R and RCS total flow rate are restored to within the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER 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 WOLF CREEK - UNIT 1 3/4 2-8

~.

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Attachment.III to WM 87-0254 Page. 8 of 11 October 2,1987

~

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.0-P)]

FIGURE 3.2 3 l

RCS TOTAL FLOW RATE VERSUS R FOUR LOOPS IN OPERATION 4

WOLF CREEK UNIT 1 3/429 4

l l $

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n 11 to WM G7-0254 Page 9 of 11 October 2,1937 2.1 SAFETY LIMITS BASES s

2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

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

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

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the W-3 correlation (R-GRID).

The.W-3 DNB correlation (R-GRID) has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

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

)

The minimum value of the DNBR during steady-state operation, normal i

(

operational transients, and anticipated transients is limited to 1.30.

This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

N These curves are based on an enthalpy hot channel factor, F f 1.55 AH, and a reference 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 expression:

F g = 1.55 [1+ +ct (1-P)]

Where P is the etion 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 f

(al) function of the Overtemperature trip.

When the axial power imbalance 2

is not within the tolerance, the axial power imbalance effect on the Over-temperature aT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

WOLF CREEK - UNIT 1 B 2-1 lt

b 1

Attachment III to WM 8M0254 Page 10 of il Octobcr 2,1987 F0WER CI5TF.'EUTI N !. D'! T 5 BASES HEAT FLUX HOT CHANNE'. FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALF: RI5E HCI CHAtmEL F AC TOR (Con:t nuec) c.

The cart.ci red insertion limits of Specification 3.1.3.6 are maintained, and l

c.

The axial pcwer distribution, expressed in terms of AX!AL FLUX DIFFERENCE, is maintained within the limits.

N I

F will be maintained within its limits provided Conditions a. tnrcugn

,g N

d. above are maintained.

AsnotedonFigure3.2-3,RCSflowrateandFy may be " traded off" against one another (i.e., a low measured RC5 ficw N

rate is acceptable if the measured F is also low) to ens m uat me c cu g

lated DNBR will not be below the design DNBR value.

TherelaxationofF},g as a function of THERMAL POWER allows changes in the radial power shape fer all permissible rod insertion limits.

QascalculatedinSpecification.3.2.3andusedinFigure3.2-3. accounts

(~

forFhH less than or equal to 1.49.

This value is used in the various accicent

(

analyses where F influences parameters other than DNBR, e.g.,

peak clac tem 3g perature, and thus is the maximum "as measured" value allowed.

)

Fuel rod bowing reduces the value of DNB ratio.

Credit is availacle to offset this reduction in the generic margin.

The generic margins, totaling 9.1% DNBR, completely offset any rod bow penalties.

This margin incluces tne l

following In addition, part of the 1)

De s '.qn limit DNSR of 1.30 vs. 1.28, remaining generic margin is used to support the 2)

Grid spacing (K ) of 0.046 vs. 0.059, current RCS minimum 5

lowable flow rate.

3)

Therma. Diffusion Coefficent of 0.038 vs. O.

e 4) 0.NSR Multiplier of 0.86 vs. 0.88, and 5)

Pitch Reduct. ion.

The applicable values of rod oow penalties are referenced in the FSAR.

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 a full-core map taken with the Incore Detector Flux Mapping System, and a 3% allowance is appropriate for manufacturing tolerance.

WOLF CREEK - UNIT 1 B 3/4 2-4 o

• Attachment III to WM 87-0254' Page 11 of'11 October 2 } 1987 GEE C:HRIEUTIM t?MITS Ea5ES 1

HEAT FLUX HOT CHA';NEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE

M CHANNEL FACTOR (Continuec)

The Radial Peating Factor, fxy(Z). is measured periodically to provide assuran:e that the Hot Channel Factor, F (z), remains within its limit.

The forRATEDTHERMALPOWER(Ffy)q F, limit as providec in the Radial Peaking Factor Limit Report per Specification 6.9.1.9 kas determi'ned from expected power control manuevers over the full range of burnup conditions in the core.

When RCS flow rate and F are measured, no additional allowances are g

necessary to comparison with.the limits of Figure 3.2-3.

Measurement l

errors of..

for RCS total flow rate and 4% for F have been allowed for in 3g determination of the design DNBR value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate incicators.

Potential fouling of the feedwater venture which might not be

(

detected could bias the result from the precision heat balance in a non-conservative manner.

Therefore, an inspection is performed of the feedwater venture each refueling outage.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to cEtect only flow degradation which could lead to operation outside the acceptacle region of operation shown on Figure 3.2-3.

This surveillance also provides adequate monitoring to detect any, core crud buildup.

3/4.2.4 0UADRANT 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.

4 Radial power distribution measurements are made during STARTUP testing and periodically dur<ng 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' limit of 1.02 was selected to provide an allowance for the uncertainty associated with the indicated power tilt.

The 2-hour time allowance for operation with a tilt condition greater inan 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 sucn ACTION does not correct the tilt, the margin for uncertainty on F is reinstated Dy reducing q

tne maximum allowed power by 3% for each percent of tilt'in excess of 1.

i l

1 I

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