Refuelling Analysis of Pu-Fuelled HWR-FUGEN

September, 1981

Hidemasa KATO Shigeru OHTERU Toru HAGA

and

Hiroshi KAMIKAWA*

Power Reactor and Nuclear Fuel Development Corporation

9-13, l-Chome, Akasaka, Minato-ku, 'Tokyo! Japan

* Century Research Center Corporation 2, 3-Chome, Honcho, Nihonbashi,,Chuo-ku, Tokyo Japan

I. INTRODUCTION

The refuelling analysis of a Pu-fuelled heavy water reactor FUGEN has been performed by using a coarse mesh; 3 dimensional diffusion code "POLESTAR-3D". The program was previousely reported:' based on a modified l-group model but its 2-group version has been added recently and both were used in the present paper.

For direct verification of this code, the calculated thermal neutron flux distributions were compared with the measured values obtained from a scanning in-core monitor, during power operation from the initial core to the 2nd cycle refuelled core.

First, the two group(coase-mesh 3D) calculations were compared with the measurements, and it was found that the calculations had sufficient accuracy.

Next, the modified one group(coarse mesh 3D) calculations were compared with the measurements,.where the accuracy was found quite reasonable with only a small difference compared with the two group cases. The "modified one groupn model considers the correction of reflector transients over the power distributions.

The procedure, accuracy and computing time are also discussed.

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II. CORF NANAGEMENT AND MONITOR LAYOUT

In the FUGEN core, sixteen insrument tubes are provided for travelling neutron monitor as shown in Fig.1. A micro fission chamber scans through these instrument tubes and provides the information of thermal neutron f:Lux distributions. This monitor is called power calibration monitor(PCM) ~

For attaining easy core management, the symmetrical refuelling and control rod patterns were selected in FUGEN reactor- Considering a quadrant rotational symme-try, the information of thermal neutron flux distributions are obtained at center positions surrounded by every four fuel assemblies.

Fuel loading patterns from the initial core to the 2nd a

cycle refuelled core are shown in Fig.2, wherle P is MOX fuels(O.66% Pu fissile + Nat.U), U is UO2 fuels(l.5%235U)

and SP is speci.al UO2 fuels(108%235U) providead for irradiation of materials.

Lower column in the figure shows the loaded fuels in the initial core, and the upper column shows the new fuels loaded inthe 1st and 2nd refuelling- The numbers 2 and 3 mean core cycles which correspond to the 1st and the 2nd refuelling, respectively*

The control rods patterns from the initial core to the 2nd cycle refuelled core during power operation are shown in Fig- 3. The reactor has 49 control rods, with the same absorber element- These control rods are identified

by the name A -to Mm The E rods are located in the center of each quadrant and used for automatic power regulation* In the figure: the numbers in the circles shctw the

withdrawn percentage stroke of control rods, and the zero corresponds to the lowest end, and 82% to the upper end

of the effective fuel zone-

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The control rod pattern was changed once in the middle of the initial core burn-up for flattening power distributions. The excess reactivities for burn-up were compensated by adjusting 1OB concentration in D20 moderator. Fig.3-1 shows the control rod pattern(LlS%, E25%, C75% withdrawn, and other rods fully withdrawn) in the initial core prior to an average burn up 4.2GDW/T (25OFPD), while Fig.3-2 shows the control rod pattern (E65% withdrawn, others fully withdrawn) after 4.2GWD/T in the initial core.

Similarly Fig.3-3 shows the pattern(E65%, C70 withdrawn, others fully withdrawn) in the 1st refuelled core, while Fig.3-4 shows the pattern(E65%,C70% withdrawn, others fully withdrawn) in the 2nd refuelled core. The C rods were at almost withdrawn(70% or 75%), only used for adjusting small reactivity.

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III. COMPARISON OF CALCULATED AND MEASURED FLUX DISTRIBUTION

The calculated and measured axial thermal neu.tron flux distributions are illustrated in Figs.4 and 5. Figures 4 and 5 show the comparisons at the beginning of the 1st and the 2nd refuelled cores, respectively. The caluculation was

performed by the two group(coarse mesh 3D) model. The calculated and measured radial flux distributions at

the beginning of the 1st and the 2nd refuelled cores are illustrated in Fig.6 and 7, respectively.

The axial flux distributions at the center monitor

(~CM#l872) and the peripheral monitor(PCM#l856) are shown* And, the radial flux distributions are shown on the two planes of 7th and 15th height(l6th corresponds to a

the top and zero to the lowest end of effective fuel zone). The flux reduces in the center of the axial direction for the aseismatic plate of Al-

From these comparisons of the calculated and the measured thermal neu.tron flux distributions, it is found that the two group(coarse mesh 3D) calculation has sufficient accuracy.

Next, in order to make the computing time shorter? a modified one group procedure is studied taking into consideration of two group feature.

One of the main difference between two group and one group calculation was considered in the reflector effect" Therefore

for the trial# a correction factor on the neutron flux distribution was introduced, which was obtained from the

l comparison of the thermal neutron flux of the two group calculation and the flux of one group calculation on the ideal core-

Figure 8 shows the distribution of the flux ratiocwhich is called Form Factor) along radial direction of the core as a parameter of BlO concentration in D20 calculated on the initial core with no burn-up and no control rods inserted.

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The form factor obtained from the calculation on the ideal initial core with no control rods insertion and with no burn-up was assumed to be valid for the correction of neutron flux distribution for both the burn-up and the refuelled cores with control rods inserted. Th,e validity of this assumption is evaluated by the comparison of the calculated neutron flux distribution with the measured values(PCM).

In the Fig.8, the radial form factor was shown. For the axial direction, it is known that no correction is necessary for the FUGEN analysis, since the axial reflectors have strong neutron absorption by the structure meterial of both upper and lower ends of the fuel assembly as shown in Table 1.

The comparisons of the calculated results by the one group procedure corrected with the form factor(which is called modified one group) with the measured ones on the 1st and the 2nd refuelled cores, are shown in Fig.9 and 10, respectively.

In the figures, upper column shows the measured values, middle column shows the modified one group calculation, and the lower column shows the two group calculation. The numbers in the parentheses are the difference(%) of the calculations from the measured value ((calculated - measured)/measured).

From these comparison, the standard deviation of the modified one grouptcoarse mesh 3D) calculation from the measured values was found 3% and the maximum difference was usually in the periphery.

Figures 11 and 12 show the channel power distributions calculated by three methods, which are the two group, the: modified one crouo, and the one croun, on the 1st and 2nd. refuelled core, resnectivelv.

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In the case of one qro"o, the core burn-up was calculated by the modified one qroup. The upper, the middle and the lower columns show the values by the ,two group, the modified one group and the one group, respectively. The numbers in the parentheses show the difference(%) from the two group calculation. From tnis result, the difference between the modified one group and the two group calculations is 2% in standard deviation. However, the one group(non- modified) calculation differs 5% in standard deviation (where the burn-up is calculated with the modified one

g=w), and it will differ more than this case when the burn-up is caiculated with non-modified one group.

Figure 13 and 14 illustrate channel power distributions along the radial direction calculated with the two group and the modified one group on the 1st and the 2nd refueled cores, respectively.

The peaks of the channel power in the 1st refueled core are at the reloaded channel shown as 2P(Fig.l3). And, the peaks of the channel power in the 2nd refueled core are at the reloaded channel in the 2nd refuelling(3?1 and the 1st refuelling(2P). From these fiqures, the modified one qrouo procedure was found qood approximation to the two qrouo calculation.

Figure 15 shows the fuel burn-up distributionscalculated with the two group and the modified one group at the end of the 2nd refuelled core.

1n this figure, the modified one group calculation gives a good approximation to the two group calculation On tine fuel

burn-up distribution, and its standard deviation is 2%. Table 11 shows the comparison of the cpu time for the two group and the modified one group calculation. The cpu times

of the burn-up calculation on the FUGEN initial core and the refuelled core are 27.4 and 11.5 seconds on a CRAY-1 for the modified one group calculation. This cpu times

correspond to 0.8 seconds per burn-up step for the modified one group. The cpu time of the modified one group is half of the two group calculation.

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IV coNcLusIoN

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On the initial and refuelled cores of Pu fuelled heavy water reactor, FUGEN, the measured thermal flux distributions were compared with the calculated results by the two group and the modified one group methods based on coarse mesh(one mesh per channel) 3 dimension diffusion approximation, using POLESTAR-3D code.

From the comparisons,.it was confirmed that the two group coarse mesh 3D calculation had sufficient accuracy to the measurement. And, the modified one group calculation using Form Factor has also sufficient accuracy(3% standard devi3,Aionj C" Tqith short computing time. The cPu time per burn-up step is 0.8 seconds in the typical case of FUGEN, on the modified one group calculation on a CRAY-1 computer.

REFEPZNCE

1. T. HAGA, N. AIHARAr H. KAMIKAWA: POLESTAR-2F/3? code for power mappinq and refuelling analysis of HWR-FUGEN, Calculation of 3-dimensional ra5ng distributions in operating reactors, Proceedinq of a specialist's meeting, Paris, 26-28 NOV. 1979 NEA-OECD p319-343(1979).

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P: 0.66@u+Nat.U U: 1.5@--235

sp: Spsial fuel(for irradiation of material)

Fig.2 Fuel loading pattern from the initial coi?~! to the second refuelled core. (l'he figure shovs a quadrant of the core which has rotational symmetry.)

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Fig.3-1 The initial loade,A core b?fore 4.2 GWD/?

Fig.3-2 The initial loaded core cif'ter 4.2GVD/T

Fis.3-4 The second refuelled, ccre

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Fig.4 Comparison of measured and calculated axial thermal flux distribution at the beginning of the first refuelled ~0re.t

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Fig-5 Comparison of measured and calculated axial thermal flux distribution at the 'beginning of the second refuelled core.

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Q 3 ------- I

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- --I----- -- , ,is* plane 1

Fig.6 Thermal neutron flux distribution on 7th,plane and 15th plane at the beginning of the fast refuelled core.

Fig.7 Thermal neutron flux distribution on 7th plane and 15th plane at the beginning of the second refuelled core.

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Fig.8 Form facto:? along core radial direction and its dependence on B-10 concentration in D2D moderator for modified one group calculation.

Table. i Dependence of Form Factor on reflector thernal neutron absorption.

0

1

2

3

!-A

5

6

x---

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UPEY

~~KLTxIO-~

b.?8x la-3

7.12 .x lC'-3

7.zb x IF3

7.37 x lo-3 7.5-3x/0-3

7.67~ !D-3

2.80X lo-3

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Fig.9-1 Thermal neutron flux distribution on 7th plane at the beginning of the first refuelled core.

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Fig.lO-2 Thermal neutron flux distribution on 15th plane at the beginning of second refuelled core.

Fig,11

Fig.12

TUJO yroup (Tim yruup 6tUrIup)

Modif;ed one yrottp tHod;fied one qroW butn up)

One group

1: Difference frm Iwo 9fou.p ca./culation (%I

Comparison of two group, modified one group :nd one group calculated channel power distribution at the beginning of the first refuelled core. (core averaged.0)

Upper: TUO group tJwo iroui bum up)

Middle: tiod;fiedone yrou? (tfodified me group k~71 alp)

Lotuefl he qroq

ConqYarison of two group, modified one group and one group calculated channel power distribution at the beginning of the second refuelled core.

The first refuelled core

p p iP P iP il ZIJ u PPSPPPU~U ! 2Pplupluu u u Pp PU uu Ll 7.P pYJ UZLJU u

u-u u u u u J

9 yuuu

I I 1 I I t--.-L--

u %J IJ IP p 2p p p UUPPPPPP u UZPPZPP P P IJLJP’PPPPP .bY- LJ u 2u P 2u SP 2p p UIJUUPPPP

u 2u u 2u P 2.P p UUUU~tlP iuu U ux.lu

PPPPPPfJU 0.5 P p Pvp2Puu

.----I PPPPPPUU

Plotted direction

cent er Radial direction

Fig.13 Cmnel power distribution at the beginning of the first ref>elled core, l core avera ~~~1.0) b l

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The 2nd refuelled core

Conirdtds: EG%, C ‘75%

ULJ u3up3up P2PP2ULJZU UL u U U2Up2U p2P3p p3up3U u LJU u Ll 3U P 3u 9 3p P p z@‘SP2U p 2u u u Ll A IJ ZP P 2P P P 3P p 3p p 3u p u u v tJ ?JlJ P 3P P 3P P p p P 2P p 2P IJ IJ U2PPzlPP P P P3Fp3pp3uu iJ U IJ f’3Up3Pp3P p P2PPZP’J2lJ’J v u 2u P 2u SPlp p p 3P w3u p 3lJ u Ll

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u2uLl 7.u PlPP p3u p3lJu u u IJLJuU~UU~U~PPZUV~~U~

r!Ju u v2Ju !J !J uu u v uLlLJLluuuu~

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Fig.14 Channel pow;r distribution at the beginning of the second relfuelled core. (core nvernge=l.O)

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