THE USE OF THE ELECTRON-PROBE MICROANALYSER TO DETERMINE THE RADIAL DISTRIBUTION OF FISSION PRODUCTS IN IRRADIATED THERMAL REACTOR FUEL

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THE USE OF THE ELECTRON-PROBE

MICROANALYSER TO DETERMINE THE RADIAL DISTRIBUTION OF FISSION PRODUCTS IN

IRRADIATED THERMAL REACTOR FUEL

J. Pearce

To cite this version:

J. Pearce. THE USE OF THE ELECTRON-PROBE MICROANALYSER TO DETERMINE THE RADIAL DISTRIBUTION OF FISSION PRODUCTS IN IRRADIATED THERMAL REACTOR FUEL. Journal de Physique Colloques, 1984, 45 (C2), pp.C2-829-C2-832. �10.1051/jphyscol:19842190�.

�jpa-00223866�

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THE USE OF THE ELECTRON-PROBE MICROANALYSER TO DETERMINE THE RADIAL DISTRIBUTION OF FISSION PRODUCTS IN IRRADIATED THERMAL REACTOR FUEL

J.H. Pearce

UKAEA, WindscaZe MucZear Power Development Laboratories, SelZafieZd, Seasca.Ze, krmbria, CAZO IPF, U.K.

Rksumk

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Les mkthodes utiliskes 2 Windscale Nuclear Laboratories (UKAEA) pour mesurer la distribution des produits de fission dans les combustibles des rkacteurs thermiques sont dkcrites. Pour amkliorer la sensibilitk analytique en prksence de radiation 6-y on rgduit au minimum les dimensions de l'kchan- tillon de U02. On discute les considkrations qui affectent la prkcision et la sensibilitk de l1.analyse de Xe, Cs, Zr, Nd, Te, I, Mo et Ru et des rksultats pour des taux de combustion de 5 5 55 GWj/tU sont prksentks. On dkcrit aussi l'emploi d'un systkme de contr6le par ordinateur.

Abstract - Techniques used at Windscale Nuclear Laboratories (UKAEA) to measure the distribution of fission products in irradiated thermal reactor fuel are described. The size of the U02 fuel sample is minimised to reduce the loss in analytical sensitivity due to beta-gamma radiation reaching the X-ray detectors. Factors affecting the accuracy and sensitivity of analysis for Xe, Cs, Zr, Nd, Te, I, Mo and Ru are discussed. Results are given for fuel with burn-ups in the range 5 to 55 GWd/tU. Use of a recently installed computer-control system for this work is also described.

1

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INTRODUCTION

In order to verify the mathematical models used to predict behaviour of nuclear fuel rods in reactor, eg the swelling behaviour of intact fuel rods, or the activity release from defective fuel rods under transient conditions, it 1s necessary to measure the radial distribution of volatile or gaseous fission products, such as Cs, Xe and I. Other authors/l,2/ have measured Xe concentration profiles in high burn- up reactor fuels; subsequently lower concentrations of Cs and Xe down to 0.05 wt-%

have been measured in thermal reactor fuels of lower burn-up in this laboratory

/ 3 / .

These early results were obtained using a manually operated Cambridge Microscan

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probe. More recently a JEOL 733 probe fitted with LINK 'Specta' computer control and energy-dispersive/wavelength-dispersive spectrometers has been used at WNL to extend the range of fisslon products measured to include Zr, Nd, Te, Mo, Ru. The equipment can also detect variations in I content, at about 200 ppm level, in high burn-up fuel ( ^%50 MWd/kgU)

.

All analytical data for fission products is obtained using wavelength dispersive spectrometers which have no direct path between specimen and X-ray counter and can therefore be easily protected against By radiation from the specimen. T k rela- tively low By-irradiation levels of the small specimens used has also enabled the use of an energy-dispersive spectrometer (EDS) for qualitative identification of unknown phases. The technique requires a high electron beam current to give a high signal/noise ratio and a variable aperture on the ED detector to control the radiation received.

Article published online by EDP Sciences and available at http://dx.doi.org/10.1051/jphyscol:19842190

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JOURNAL DE PHYSIQUE

2 - SAMPLE PREPARATION

The first step is to cut a 1 mm thick transverse slice of the irradiated fuel rod;

this is embedded with silver-load epoxy resin in a plastic mount, recessed by 0.7 mm to hold the sample. The transverse slice is then ground flat and cuts s 4 mm deep are made in the face to provide a radial strip of fuel plus cladding, about 6 mm long by 1 mm wide. The final transverse cut to free the sample is made 2-3 mm from the ground face, to include part of the mount, which supports the cracked fuel and makes the small sample easier to handle remotely. The sample is then re-embedded face down in a conducting mount, again recessed by 0.7 mm and using a silver-loaded resin. The sample is metallographically prepared using non-aqueous solutions to provide a polished radial strip < 0.5 mm thick, having a y-activity of typically

s 0.5 R /h at 50 mm.

Fig. 1 - Preparation of small radial strip (A) from fuel rod section

It is not necessary to put a conductive coating on the fuel sample, as the U02 has sufficient electrical conductivity to prevent serious charging effects.

3 - ANALYTICAL TECHNIQUE

Only the techniques now used with computer-controlled operation are described, since details of the earlier work using a manually controlled instrument have already been published /3/. It has been found that about 20 analyses using spots or areas (typically 100 prn x 20 urn) along the fuel pellet radius are required to determine the element concentration profiles. The area analyses generally show less scatter, par- ticularly for Xe.

The electron image is used to select each analysis position (X,Y) avoiding large cracks or pores in the fuel. The specimen height along the fuel radius is also checked with an optical microscope to ensure spectrometer focus, and these data are stored on disc to control the subsequent automated analysis.

In some instances the fission product L line used overlaps with minor uranium lines, this is corrected for by measuring the apparent concentration of the fission product at that wavelength in an unirradiated U02 standard and subtracting this from the L line obtained from the irradiated sample. This correction can be adjusted to take account of the lower uranium content in the irradiated sample. Each analysis is individually ZAF corrected under computer control.

For some elements one of the two background measurements normally taken is eliminated to avoid resetting spectrometers on a steeply changing part of the uranium spectrum.

Similarly, the peak-seek computer routine is suppressed for fission product data to prevent an incorrect spectrometer recalibration due to the underlying uranium spectrum. Cornparison of spectra from unirradiated and high burn-up thermal reactor U02 fuel illustrates the peak overlap problem and also the small increase in background caused by y radiation from the samples used.

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Where solid standards are not available, as in the case of Xe, interpolation using elements of adjacent atomic number (ie Sb or Te) is necessary /4/. The LINK

'specta' computer program has a pseudo-standards routine which calls in a pre- selected adjacent element standard, the necessary spectrometer 'offsets' and cali- bration factor. It has also been found convenient to use this procedure where, otherwise a compound standard would have to be used, eg for Cs, I and Ba. When interpreting the xenon results it is necessary to remember that a deficiency in the X-ray signal can occur when this element is present as gas bubbles /5/. This effect is usually small for thermal reactor fuel.

4 - RESULTS

Part of a computer print-out for analysis on 18 GWd/tU fuel close to the pellet rim is given below (15 Kv, 150 nA, 100 S counting time on peak and background, a 40,000 cps on Sb pseudo-standard).

Element C apparent Error (Wt%) P-B (c/s) B (c/s) ZAF % Element U :MA 86.286 .I62 17610.6 410.210 .994 86.748 Xe : LA .292 .007 133.687 282.990 .858 .340 Cs :LA .257 .007 125.327 343.057 .851 .302 Te : LA .092 .006 36.8802 200.880 .831 .I11 Zr : LA -238 .011 19.9716 24.3709 .997 .239

0 (By difference) .I29 12.261

The error column is fractional error x apparent concentration and is calculated from the expected variances on peak and background sample measurements. The esti- mated detection limit for fission products (95% confidence limit) is about 200 ppm under these conditiops. The correction for Cs and Xe in this analysis due to over- lapping minor uranium peaks is -0.120 wt % and -0.020 wt % respectively as deter- mined from the apparent concentrations found in unirradiated U02- Similarly a correction for a Te La overlap with the P u M y is necessary (Pu is at present at 2. 0.5 wt % /3/). Nd and Ce specta also partially overlap but these elements are expected to have similar distributions in the fuel and deconvolution is not strictly

necessary.

Typical results using area and spot analysis are shown in Figs 3 and 4 respectively.

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100s each on X = Xe P and B

a =

Cs

o !

^I ^I I

RIM

CENTRE

C900°C1 O RIM CENTRE ~ ~ 0 0 ~ ~ 1 ~ Fig. 3

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Radial profiles using area analysis (low burn-up and high burn-up fuel)

For Xe, Cs, data obtained using area analysis the width of the scatter band (40 = 280 ppm) is about that expected purely from counting statistics. For spot analysis the beam was positioned in grain centres using the back-scattered electron image to select the analysis point. The scatter in these spot analyses arises mainly from counting statistics when the element is chemically stable (as in the case of Zr), but also from migration and concentration into separate phases (eg Ru, Mo) in the hotter parts of the fuel.

100s each on ₀= MO

P and B = Zr

1

^400s^P^and^{each on}^B

600

, = I

I

CENTRE> d i ~

Fig. 4 - Radial profiles using grain centre spot analysis on fuel at 55 GWd/tU.

5 - CONCLUDING REMARKS

The sensitivity of the analysis is sufficient to detect changes of about 100 ppm in fission product concentration at 400s counting times. Comparison of calculated and measured concentrations suggest that for Xe and Cs the accuracy is better than 10% /2,3/. Results confirm that the gaseous and more volatile elements (Xe, Cs and I) tend to be lost from the hotter central parts of the U02 fuel and that there is an enhanced build-up of fission productsin anarrow zone close to the pellet surface due to fission of additional Pu in this region arising from epithermal neutron capture effects.

6 - REFERENCES

1. H. Kleykamp. J. Nucl. Mater. 80 (1979) 13.

2. H. Kleykamp. J. Nucl. Mater. 84 (1979) 109.

3. J. H. Pearce, R. Surnerling and R. Hargreaves. J. Nucl. Mater. 116 (1983) 1.

4. C. T. Walker. J. Nucl. Mater. 80 (1979) 190.

5. C. Ronchi and C. T. Walker. J. Phys. D. 13 (1980) 2175.