ML20059D434

Proposed Tech Specs Re Reactor Trip Sys Instrumentation Trip Setpoint

ML20059D434
Person / Time
Site:	Farley
Issue date:	07/31/1990
From:	ALABAMA POWER CO.
To:
Shared Package
ML20059D432	List: ML20059D427 ML20059D434 ML20059D437
References
NUDOCS 9009070039
Download: ML20059D434 (10)

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Unit 2 Revision 2

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8. .I , .2 .5 .4 .5 .6 , , 7 .0 .9 1. 1.1 12 POWCR Iftoction of nominell Figure 2.1-1 Reactor Core Safety Limit' Three Loops in Operation 4

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I 22 Amendment No. )

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.,, TABLE 2.2-1 ,

g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS l

FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES d'

1. Manual Reactor Trip Not Applicable Not Applicable w

2. Power Range,-Neutron Flux Low Setpoint - f 25% of RATED Low Setpoint - f 26% of RATED THERMAL POWER THERMAL POWER High Setp? int - f 109% of RATED High Setpoint - f 110% of RATED THERMAL POWER THERMAL POVER

3. Power Range, Neu e mn Flux, f 5% of RATED THERMAL POWER vith f 5.5% of RATED THERMAL POWER High Positive Es .

a time constant 1 2 second with a time constant 1 2 second

4. Power Range, Neutron Flux, f 5% of RATED THERMAL POWER vith f 5.5% of RATED THERMAL POWER High Negative Rate a time constant 1 2 second with a time constant 1 2 second

5. Intermediate Range, Neutron

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f 25% of RATED THERMAL POWER $ 30% of RATED THERMAL POUER t'.

w Flux.

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6. Source Range, Neutron Flux $ 10 counts per second 5 1.3 X 10' counts per second' .

7. Overtemperature aT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3

9. Pressurizer Pressure-Lov 1 1865 psig 1 1855 psig

10. Pressurizer Pressure-High f 2385 psig 5 2395 psig

11. Pressurizer Vater f 92% of instrument span f 93% of instrument span Level-High

12. Loss of Flov .1 90% of design flow per loop

• 189% of design flow per loop

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• Design flow is 87,200 gpa per loop. l E

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$ 1' E REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

• NOTATION I

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@ Note 1: Overtemperature aT f AT, [K 1-K 3 1+TS3 (T - T') + K3 (P - P') - fg (al)]

H 1 +T 2 8 w

where: ar, = Indicated AT at RATED THERMAL POWER T = Average temperature, "F T' f 577.2*F (Maximum Reference T,,, at RATED THERMAL POWER)

P = Pressurizer pressure, psig P' = 2235 psig (Nominal RCS operating pressure)

I+T3 1 = The function generated by the lead-lag controller for y T,,,. dynamic compensation g 1+T8 2 T1 &T 3

= Time constants . utilized in the lead-lag controller 'for T,,, T3 = . 30 secs, T 3

- 4 secs.: .

S = Laplace transform operator, sec *.

Operation with 3-loops Operation with 2 loops K1 = 1.18 K3 = (values blank pending l K2 = 0.0154 K2 = NRC approval of K3 = 0.000635 K3 = 2. loop operation) and f 1 (AI) is a function of the indicated difference between top and. bottom detectors of the power-range nuclear ion..

chambers; with gains to'be selected based on-measured instrument response during plant startup tests such that:

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, TABLE 2.2-1 (Continued) .

SE' p REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS -

" NOTATION continued i

g (i) for q - q between -35 percent and +9 percent, f (M) = 0 (where q and q are percent RATED y THERMIL P0 b in the top and bottom halves of the3 core respectively,,and q,,+ q, is total THERMAL m

POVER in percent of RATED THERMAL POVER).

(ii) ~for each percent that the magnitude of (q, - q,) exceeds -35 percent, the E trip setpoint shall' be automatically reduced by 1.37 percent of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (qt - q,) exceeds.+9 percent, the & trip setpoint shall be automatically reduced by 1.75 percent of its value at RATED THERMAL POWER. .. l-TS3 Note 2: Overpower E f E ,[K, - K 3 1+T3 s T - K, (T-T")-f 2( M)}

where: E ,= Indicated E at RATED THERMAL POVER T = Average temperature, "F

& T" = Reference T,,, at RATED THERMAL POWER (Calibration temperature for E instrumentation, 5 577.2*F) .

K, = 1.08 K3 = 0.02/*F for increasing average temperature and 0 for decreasing average temperature -

K, - 0.00109/*F for T > T"; K, = 0 for T f T* ,

T3 3 - The function generated by the rate lag controller for T,,, dynamic compensation 1+T3S k

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TABM 3.2-1 ,

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I LIMITS 4 PARAMETER 3 Loops in Operation 2 Loops in Operation g

Reactor Coolant System T,,, f 581.2*F Pressurizer Pressure 1 2220 psia

• 1 261,600 gpa **- j.

Reactor Coolant System Total Flow Rate E

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• Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER' step in excess of 10% of RATED THERMAL POWER.

• • Values blank pending NRC approval of 2. loop operation.

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5 EVALUATION OF SIGNIFICANT HAZARDS CONSIDERATION i L

Pursuant to the requirements in 10CFR50.92, each application for amendment to an operating license must be reviewed to determine if the modification involves- '

a significant hazard. The amendment as defined in WCAP;12659, describing the

" Alabama Power J. M. Farley Unit 2 Increased Steam Generator' Tube Plugging and Reduced Thermal Design Flow Licensing Report" has been reviewed and deemed not to involve a significant hazard based on the following evaluation.

-The proposed amendment involves an increase in equivalent plugging limits from the current licensed value of 10% uniform plugging to a new licensed value of 15% avera decrease ge with a 20% peak in any.one steam generator. Also included is a of- approximately'1.5% in Reactor Coolant System (RCS) total flowrate-from the current' licensed value of 265,500 gpm to a new licensed value of ,

261,600 gpm. As: discussed in WCAP-12659, a comprehensive evaluation of the effects.of the-increased tube plugging and reduced RCS flowrate has been completed, and no adverse safety implications have been identified.

Neither the increased tube plugging nor the reduced RCS flowrate involve a significant increase in-the probability or consequences.of any accident previously evaluated.

Tte LOCA and non-LOCA-accidents were reviewed in-WCAP-12659 verifying that the effects'of' increased tube plugging and reduced RCS flowrate do not invalidate the current analyses of record and-that all design basis conclusions are still met.

Several non-LOCA accidents [ main feedwater pipe rupture, uncontrolled bank withdrawal: from subcritical, partial -

loss of flow, single rea; tor coolant pump locked rotor,-' steam generator tube i rupture (doses)) were reanalyzed using the~ revised conditions associated with increased tube-plugging and reduced RCS flowrate, and acceptable _results were obtained. For those nor.-LOCA accidents for which evaluations we a performed, acceptable results were obtained by use of existing sensitivity studies, existing margins, or allocation of generic DNB margin. In all cases, current licensing criteria of r3 cord is met. -In addition, . effects of asymmetrical flow distributions have been evaluated and found acceptable. The most limiting large Break LOCA analysis (C d =0.4) was reanalyzed for the new configuration, and the analysis demonstrated a calculated PCT less.than.the Appendix K limit of 2200'F. ' Evaluations of Small Break LOCA, LOCA hydraulic forcing functions, post-LOCA long-term core cooling, and hot leg switchover to prevent boron precipitation wt_.re performed, and all current conclusions for the J. M.

Farley snit 2 remain valid.

Evaluations of MSLB tnd LOCA mass and energy releases concluded the present-mass and energy releases are-applicable and the containment responses remain valid and all licensing conclusions remain valid.

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Neither the increased tube plugging nor the reduction in RCS total flowrate

• creates the possibility of a new or different kind of accident from any accident previously evaluated. These effects were reviewed in detail'in WCAP 12659 for any adverse effects on RCS components. Major Nuclear Steam Supply System components (i.e., The reactor vessel, internals, loop piping, reactor coolant pump, pressurizer, and the CRDMS) were reviewed, and no adverse safety effects were found. No new single failures were found. No new accident initiators were found. Since increased tube plugging can physically alter the steam generator, the possible effects on the steam generator were evaluated in ,

detail. The results of thermal and hydraulic evaluation, U-bend tube vibration assessment, and the structural evaluation have concluded that the current components of the J. M. Farley Unit 2 steam generators satisfy the requirements of the ASME B&PV,Section III for the increased tube plugging and reduced RCS flowrate. No significant reduction in a margin of safety is involved with the increased 0.be plugging and reduced RCS flowrate since all. current acceptance criteria continue to be met.

The evaluation of the effect of these variables on non-LOCA and LOCA transients has verified that plant operation will be maintained within the bounds of safe, analyzed conditions as defined in the FSAR with the revised technical specifications, and the conclusions presented in the FSAR remain valid. Safe operation with revised core limits, a revised K1 of 1.18 and an approximate 1.5% reduction in thermal design flow has been demonstrated.The impact on K4 was small and no change to K4 was required since adequate-margin exists in the Technical Specification value. In addition, component evaluations have been performed to demonstrate compliance to current acceptance criteria. As such, no reduction in a margin of safety as defined in the basis for any technical specification is required for operation of J. M. Farley Unit 2 with' increased steam generator tube plugging and reduced RCS flowrats  :

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