ML20247P197

Ro:On 890725,fuel Element 141 Had Estimated U-235 Burnup to 18.9 G.Operator Instructed That in Addition to Quarterly Calculations Required by Tech Specs,Operator Will Make Projections to Insure That No Element Will Reach Limit

ML20247P197
Person / Time
Site:	Rhode Island Atomic Energy Commission
Issue date:	07/25/1989
From:	Dimeglio A RHODE ISLAND, STATE OF
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 8908040084
Download: ML20247P197 (3)

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TATE OF RI ODE ISLAND AND PROVIDENCE PLANTATIONS Rhode Island ' Atomic Energy Commission NUCLEAR SCIENCE CENTER South Ferry Road,

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. Narragansett, R.I. 02802-1197 July 25,1989

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. U.S. ' Nuclear Regulatory Commission Document ' Control Desk W'

Washington, D.C.

20555 License R-95 Docket 50-193 m.

Gontlement On, Tuesday, morning, July 25, 1989, I was informed by the operator responsible

. for fuel = burn-up records that fuel ' element number 141 had an estimated U-235

- burn up of L18.9 grams and that this may-be in violation of section K.3.e(4)(f)l.

of the-Technical Specifications which requires that the fission ' density limit shall be 0.5 x 1021: f ssion/cc.

Since the technical specifications do not qualify howD this' requirement 7 shall. be met, we have performed.

a.. conservative

. calculation converting this fission s density limit to an element fuel burn-up limit' considering. flux distribution end ' other uncertainties.

A. copy. of the 2

memorandum showing the calculations is ' attached.

From "the 5 memorandum, it is noted that by spplying all the " hot-spot"~ factors.

. the burn-up limit is' 17.94 grams in a fuel element.

While it is ou. intention and practice to remove from service an' element which has ac hi...:d 17.94 grams burn-up, to go. beyond this limit is not a violation of the technical

- specifications. since it is unreasonable to expect all the hot spot factors to apply

< at - the i same point.

In addition, the-hot spot factors have been combined using the. multiplicative - techn'.que rather than a statistical technique.

For these reasons,' we believe that while we have violated an in-house requirement.. we have not violated the technical spe,.ification.

To prevent a reoccurrence of this, the operator has been instructed that in addition to the

' quarterly calculations required by the technical specifications, he will make

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projections to insure that no element will reach its limit during the next quarter.

y truly yours, I

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A. Francis DiMeglio Director L'

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FROM:

A'.' Francis.DiMeglio, Director W

SUBJECT:

. Tech Spec. Fuel Burnup Limit t

.I.' Since" July' 31, 1980,- the technical specifications have contained a limit on fission density for' all types of fuel elements of 0.5x10"~ fissions /cc.

At-~that time a calculation. -(attachment 1) was performed to convert the f.ission density. limitation to a burnup limit in grams since records are

. maintained.of total burnup in each element. The following is a detailed i

. discussion of the method to be used in obtaining the appropriate burnup ]

- limit. Although not explicitely stated, it is assumed that the limit imposed. is. the maximum permitted at any " spot'.' in 'a fuel plate and not an average.over.an entire plate.

113 The following adjustments will be made to.the tech spec limit:

.1.

Peak to average flux along.the length of a fuel element - At the center 1

of the core the peak to average is 1.37.

In a peripheral element the-peak to average would probably be'somewhat less.

.o, 2.

Centerline'to outside plate correction - The percent of power developed' f

-in an elemert is determined by flux determinations along the length o :

the center channel. However, burnup will not occur in each plate of an element at a uniform rate since the plate closer to;the core center is generally in a higher flux than the centerline flux.

(This effect will

- tend to cancel in practice-because elements are frequently rotated 1800 when being moved from one grid position to another). Based on inter-polations of' flux plot curves, the ratio of outside plate flux to center-line flux.is a maximum of 1.25.

3.

Accuracy of fractional power developed per element - The fractionD1 power developed in each element is determined by extrapolation of-fliix plot cdata.

Since an element moves about the core during the lifetime of the element and since the entire core generates 100% of power,' underestimates 'l

. of burnup in one core location tends' to be compensated by the necessary.

overestimate while the element is In another location. However, since a~

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. flux plot is a i 10% procedure, the ratio of actual: fractional power to Y'

the' fraction used for calculating may be 1.2.

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Ratio'of true reactor power to power level estimate from instrument's -

.; An allowance oft 20% (see II.3 above)_for flux plot determination and ey extrapolations is-suf ficient to allow for instrument error, ri(# L

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- 5.1. Volume of fue1~ core based on fabrication specifications - Assuming ea'ch'

' meat dimension is everywhere as.small as, the fabrication specifications permit would lead.to ratio of, spec volume to true volume of 1.083.'

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That.11sp the actual volume'may be 8.3%'3ess than the' volume based on

,. nominal dimensions.

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. 6 Fuel' loading

.The fabrication specifications permit a' loading variation

• n per' plate of..i.2%.. Therefore,1 the: ratio is 1.02.

, 7.'

Fuel inhomogeneity - Uniformity.of fuel density in'a plate.is determined

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.by x-ray.fluros' copy. The specifications permit =4% discrepancy.. There-fore,.the' ratio is 1.04.

III'.

Using a fission to. captive. cross section ratio of.8'4, the uncorrected burn-up:per element in'gm/ element is:

_ f [Xff A *

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• A 4 sw +A -

= va.Mr e~/As IV. ' Applying the correction factors in II above, this becomes:

W. Y ]

n 2 7 x t. a s A i. ). x j. W 3 x /. c' a. x / A y

/7. N f 28, J V.

Assuming that all the. correction factors. apply to the " spot" with greatest

, fission density, the maximum calculated burnup permitted in an element is 17.94 gm.

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