ML20052D037

Safety Evaluation Re Induced Neutron Flux Errors for B&W Reactors

ML20052D037
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/21/1982
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20052D033	List: ML20052D032 ML20052D037
References
NUDOCS 8205060211
Download: ML20052D037 (7)

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SAFETY EVALVATXON REPORT BY THE OFFICE OF NRR CONCERNING INDUCED NEUTRON FLUX ERROR FOR BABC0CK AND WILC0X REACTORS Introduction In October 1980 Babcock & Wilcox (34W) indicated (Ref.1) that stu

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recently perforned had concluded that event induced errors in the neutron

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flux detector readings and thus effective flux trip levels could be larger The staff resporided, f

for soma events than those noraally assumed in analyses.

n folicwing conversations with B&W, by requiring infornation from utilities The utilities with operating BAW reactors have responded (Pef. 3) fr f. 2).

e The response and review are sumarized and the re ponse has been revie. cad.

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4 evants the colder water in In brief the problems are (1) for some co:ldcu.

7 the dwcence rojien increases neutron flux.sttenuation thus potentially 1

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f increasing the transient flux crcor en the excore nuclear instruientation

( "I) 5:ycnd ti,e 23 nor ally used in analysis, and (2) far ccntrol rod f

eJoction e.~ ots t',e n:atron flux distribution change resulting feca the

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l atncemal coatrol rod rattern ecuses effective icvels in the excore de i

90th effects affect trip levels to chcnge (for a given core averaqe level).

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and potentially in an a-cunt beyond that norca i

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required.

4 This, along with other I

? omally SW has used a 2t transient flux error.

assuaed errors and a trip setpoint of 105.5% of full porter gives a trip in

%s2d on MlIS'l calculations (from the '?PSS stady) to analyses of 112.i.

a'll the mxinal transient inlet i

translate dowr.coc.er te. perature changes to l

teuperature reduction of about 16*F corresponds to 13% a'I, giving an effe Duke exaiained data fro.1 a nu.r,5er of test programs j

trip point of 123%.

Based on these tests they developed

'I relating tenperature and flux readings.

a relationship (linear with tenperature) between inlet tenperature and S HI (at a 95% confidence level).

It would provide a 12% a tlI at 16*F.

For nuch Af41 at 16*F).

of their analysis, however, they used a 1% a NI/1*F factor (16%

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],i Using the calculated dll vs inle.t tenperature relationship, B&W developed, for each reactor, at its minimum pressure (trip setpoint) a (graphical) relationship between reactor power, outlet temperature, trip lines (high flux with error and variable low pressure - outlet te.nperature) and thus (This is best described regions protected by the reactor protection system.

in the Davis-Besse submittal). They superimposed on this D!!BR values calculated using design power distributions.

The results, -which of course take advantage 1

of the improved D"2R value at the lower inlet tenperature conditions, demonstrate i

that D'!3R limits (both 1.30 and 1.43 which includes a 10.2% rod bouing iil penalty) fall within the protected region for overcooling conditions out to,

'J Pouer distribution calculation for 125% full i

and beyond,16*F overcooling.

'I poitor conditions ware also done to check perturbations in distributions at j

TI,ase were also used to de.~onstrate urgin to j

these li aitirig conditions.

i D':3 and center fuel nalt (CF!1) liaits.

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Duke perfori.ied plant specific analyses for each cv2rcooling transient, 1;j including '.he turbine bypass event (also giving the ni).ina ecercocling as j

abcve) and the lar gar stcaaline 5.caks occidents (astu;ing a hich flux trip t

4 is rcquired). They used 1%.WI/*F to identify n3xinn (non trip) power I

'i UI for the turbine hyp;ss) Sud ass:.aed ICS failures 3

levels (gidng a50ut lit 1

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to naxinire overcooling and analyzed for O!!B using design peaking factors.

They fcund that D!!3 and CF:4 limits v:re not excccded, even without the 1 1 i l reduction which would have been provided by a lower trip level which would 4

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occur using the derived afil - tenperature error rather than 1% SNI/*F.

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, The review of all of the submittals has lead to the conclusion that the nagnitude and extent of the effect and its consequences during events of 1

interest have been suitably examined.

The B3W calculations and the Duke 3i neasurements co1plenent each other on the nagnitude of ANI vs temperature as-do the compleacntary calculations for the nagnitude of temperature decrease to be c'onsidered during naximum events. Using this information the protection ~

system appears to be able to provide protection before exceeding limits on j?

[j US and CFM.

Hevever, all future submittals which require analysis of cvarcooling events by B&W reactors should include in the analysis and presenta-tion an equivalent of the information involved in the present su%ittals and I

the ase of the penalties resulting frc.a inlet cooling siallar to these unless nea v31;2s are justified.

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The other cvent involving a potential indication error for the flux signal, which in turn is involved in terainating the event by a trip signal, is the rod ejection accident.

In this case the error arises froa the change in po cr distribution cused by tSe ejected rod making the erfective power level as sacn by the flux detactor different froa the average used in (point kinetics) analyses.

The ?rc51en, as related to trip, would only exist for small sesrth roels (neig5Scrhood of 0.7% Ak or less) since the rise in flux j

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level is too large to significantly af fect trip cecurrdace end tiaing for 4

1l lar ger rarctivity insertions.

Since the BSW "zero pwar" event oualyses noraally involve high pressure trips rather tbn high flux trip for smaller

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rod worths, the problem is only relevant to the full poi.er analyses which

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are nonnally analyzed as tripping on high flux.

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5-1 The B&W subnittals argued on the basis of e,ngineering judgment that, if heat transfer out of the fuel pin during the transient were included in the a

l ejection analysis (as has not been the case in past submittals), 'the power and peaking increases for the range of reactivity insertion that might not cause flux trips would not result in peak enthalpies exceeding limits (280 Duke presented results of calculations of flux errors resulting W

cal /gm).

fro, a nuaber of rod configurations, providing a basis for a correlation of error with rod worth, and also presented typical power histories as a function From these it can be concluded that there would be a high of rod worth.

for a rod worth above about 0.1% Lk at a trip level of about 120%

flux trip (rather then the usually ass)ned 112%).

For rods under this level there night not Se a flux trip, he aver, power le.'els and peaking factors associated

'f with these rod worths are sufficiently lu that the limit for the event (280 j

I The initial transient is ainor and the quasi-l]a cal /3a) is not approached.

The steady state is sicilar to that of the single rod withdrawal cvent.

latter is described in the Midland S*A u'iere it is indicated, in an analysis i

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,I with heat transfer, that 230 cal /ga is not approached (nor is D'!3 reached) for even larger s od earths than are involved here (e.g., greater than 0.3% ak).

l The raview of the subaittals has lead to the conclusion that the flux error

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associated with the changed power distribution for rod ejection does not significantly af fect the trip functica for the larger rod worth eveats and

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that the consequences for the saaller worth. events are not of a aagnitude to i

approach limits when considering the heat transfer that occurs.

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o la-1 if Suniary, and Conclusions The effective neutron flux trip level in BaW reactors may be raised above 1

that normally used in analyses because of increased flux attenuation in~ the

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downconer in cooldown events and because of power (flux) distribution changes Hm;ever, analyses of extrene cooldovin events in the rod ejection event.

that sufficient nargin exists in the trip requiring high flux trip indicate levels, as augmented by the inprovenent in D"BR provided by the cooldown, The that linits on 0"3 and CFM are not exceeded in operating reactors.

review of this analysis has resulted in agreenent with this conclusion for

>I Heuever, all future analyses of these events for B&W cperating reactors.

rcactors should include in the effective trip level fcr cooldown events a I

i suita51e flux error tera of a njnitude as discuss.$d in this review, e.g.

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13% L!ll for a 16*F cooldoven, or as specifically derived for the reactor as 1

For the rod ejection event the analysis of the has been dor,e by Duke.

1 increased error indicat.>s that the only events which nay 50 significantly il

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af fected are these,,ith snaller rod vorths fqr s/aich the consequences are

^l The ravicw has concluded that i

bc;1cw li..its cvcn uitbo;t a hi;h flux trip.

no chtn;es see nee kd in op; rating paraneters for currently operating reactors i

because of this error.

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References d

'F Letter from James Taylor (B&W) to Victor Stello OlRC), October 29, 1930:

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"Results of Recent Induced Flux Error Investigations."

i'emorandum from L. S. Rubenstein OlRC) to T. Novak (NRC), November 28, 1930:

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"B%' Induced Flux Error."

'3.

Letters from the following utilities on the indicated dates to the

. i, NRC, Operating Reactor Branch 4 a

Toledo Edison, March 18, 1931 v.

Duke Power Co., March 19, 1981 Sacramnto I'Jnicipal Utility District, " arch 20, 1931.

"2tronlitan Edison Co., Se;)tenher 29, 1981.

Florida Fo. tar Corp., " arch 20, 1981 i

i Arkansas Pc..er & Light Co., January 30, 1981.

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