ML20247P197
| 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
.J
. 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
/
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~
.,. L
. 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
~~ 7
- 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.'
^
That.11sp the actual volume'may be 8.3%'3ess than the' volume based on
,. nominal dimensions.
I
. 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
[
4-
.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 *
/hol*b5 E'
A
(,, - O,1, X to AI &[
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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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