size (\im) ON OooOO O
6. COMPOSITE FUELS
The composite is made up of an inert matrix in which the fuel is dispersed in granular form. This concept is the most innovative and its introduction in nuclear reactors is still far.
Its potential advantages are nonetheless undeniable and attractive. With regard to safety, for example, the matrix could promote thermal transfer so that the fuel temperature and the specific fission gas release would be lowered ; the matrix also forms a second barrier (after the fuel itself) to the migration of fission products.
Our study [8] concerns CERMET-type composite (Fig. 7) with a molybdenum metallic matrix and a CERCER-type composite with a MgAl2O4 ceramic matrix. They contain 36 vol.
% of UO2 particles with a 19.6 wt % of 235U enrichment.
The CERMET fuel stays cold during irradiation (§ 7, Fig. 8). The irradiated pellets are not cracked and remain as they were before irradiation. The fraction of 85Kr released during
Standard UO,
After chemical etching
40 um 40 um
40 jam
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UO2 + A12O3 + SiO2 40 (with an other Si/Al ratio)
Before chemical etching
U02 + A1203 + S.E.M. picture
Si0 15 um
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(with an other Si/Al ratio) S.E.M. picture
F/G. 6: Microstrucrure of improved cesium retention fuels
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. 200
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U02/Mo interface
l O f i m F/G. 7 : Microstructure of the UOMo CERMET fuel
irradiation is 2.1% due to ejection and recoil from UO2 granules emerging at the pellet surface.
The fission gas release during post-irradiation annealings is very low in view of the high burnup, close to 50000 MWd.V1, and the temperatures that it would never reach under irradiation.
The CERCER fuel behaves like a standard UO2 for the thermal (§ 7, Fig. 8). The fraction of 85Kr released during irradiation is only 0.7% due to the existence of a strong contact between fuel and cladding during most of the irradiation period. This swelling seems to be caused by the transformation of the matrix around UO2 particles on the edge of the pellets.
The fission gas release during post-irradiation annealings is rather high because the CERCER cracks during the irradiation period and subsequent post-irradiation relief annealing and no longer forms a tight barrier with respect to fission products.
7. THERMAL ANALYSIS
The thermal analysis was performed on the basis of the experimental and calculated relationships between centre-line temperature and power, as represented on Figure 8.
In view of the low contents of additives introduced in the advanced microstructure fuels with or without Er2O3, their thermal conductivities may be considered to be equivalent to that of standard UO2. The thermal response of a given fuel rod changes during the irradiation cycle
1JLUU '
150,00 200,00 250,00 300,00 35C
o
— — Cs reterticn fuel calcdation
Linear power (Won)
FIG. 8 : Centre line temperature versus linear power diagram
when the power exceeds a certain limit (as can be seen for the reference UO2 fuel) : the thermal conductivity reduces from an "out-of-pile" value to an "in-pile" value ; this is indicative of the degree of fuel fragmentation.
For the fuels with enhanced caesium retention capacity, the high proportion of additives contributes to a reduction in U02 thermal conductivity of about 5% estimated according to the Maxwell-Eucken's law of mixtures. The rods effectively tend to be much hotter than the reference UO2 fuel.
Experimental data show that the thermal conductivity of the CERCER is the same as the UO2 one. The calculation with the Maxwell-Eucken law does not confirm that: either the knowledge of the thermal conductivity of the MgAl2O4 spinel is insufficient or the heat tranfert between UO2 particles and matrix is not good enough.
The thermal conductivity of the CERMET is much higher than the UO2 one and in good agreement with calculation with Maxwell-Eucken law.
8. CONCLUSION
Given the diversity of this research on innovative UO2 fuels, coupled with the originality and efficiency of the experimental procedure, the authors are able to propose solutions to meet the stated requirements for improved fission product retention at high burnups, extension of irradiation durations and minimisation of pellet-cladding interaction effects.
Thanks to the relationships which exist or that can be made between the various studies presented in this article, progress has been made in the definition of a UO2 fuel in which the maximum number of properties are improved.
Fuel with a low doping agent content would seem to hold out most promise and its use is being continued in order to optimize compositions and microstructures.
The composite fuel concept is interesting as a study tool ; for example, it can be used to obtain UO2 at very high burnups having been subject to special irradiation conditions such as low temperatures in the CERMET case.
These basic studies are supported by the FRAMATOME and EDF companies and will be followed by numerical calculation on the behaviour in reactor and irradiations in PWR.
REFERENCES
[1] DEHAUDT, Ph., CAILLOT, L., DELETTE, G., EMINET, G., MOCELLIN, A., Irradiation of UO2+X fuels in the TANOX device, TCM IAEA Tokyo, these Proceedings.
[2] BOURGEOIS, L., Effect of additives on enhanced sintering and grain growth in uranium dioxide, Thesis, 17 June 1992.
[3] PEREZ, V., Contribution to the study of second phases particles dispersion in polycristalline uranium dioxide, Thesis, 17 dec. 1993.
[4] DUGUAY, C., DEHAUDT, Ph., SLADKOFF, M., MOCELLIN, A., High temperature mechanical tests performed on doped fuels, TCM IAEA Tokyo, these Proceedings.
[5] ASOU, M., DEHAUDT, Ph., PORTA, J., Assessment of erbium as candidate burnable absorber for future PWR operating cycles : a neutronic and fabrication study, International KTG/ENS TOPical Meeting on Nuclear Fuel, Wurzburg, Germany, 12-15 march 1995, vol. II, p. 79-84.
[6] JOLICART, G., LEBLANC, M., MOREL, B., DEHAUDT, Ph., DUBOIS, S., Hydrothermal synthesis and structure determination of Cs2ZrSi6O15, Eur. J. Solid State Inorg. Chem., t33, n°7, p. 647-657, 1996.
[7] GAMAURY, S., MOREL, B., DEHAUDT, Ph., Advanced fuel with improved cesium retention, a study using simulated fuel, International KTG/ENS TOPical Meeting on Nuclear Fuel, Wurzburg, Germany, 12-15 march 1995, vol. II, p. 85-90.
[8] DEHAUDT, Ph., MOCELLIN, A., EMINET, G., CAILLOT, L., DELETTE, G., Composite fuels behaviour under and after irradiation, IAEA TCM on Research of Fuel Aimed at Low Fission Gas Release, Moscow, Russia, 1-4 October 1996.
DISCUSSION
(Questions are given in italics)
Do you have any plans to test the Cs retentive fuel in steam, which must be the proper condition?
Cs release is provoked by annealing. Today, experiments are in progress on as-irradiated pellet in inert gas or vacuum. Then, the same experiment will be performed after a pre-oxidation of the fuel.
The additives to Erfuel have 2 effects:
to increase sinterability. But are densities of 98-99% TD, to be compensated by pore-former additions, a desirable feature?
to enhance uniform dispersion ofEr within the fuel. But, since lower neutron cross-section of Er than Gd makes self-shielding less of a problem and since a low consumption rate of the burnable absorption is a reason to select Er rather than Gd, is homogeneous dispersion ofEr in the fuel of any benefit?
High densities are achieved with additives which are introduced first to enhance Er solubility and grain growth. Perhaps, homogeneous dispersion ofEr is not absolutely indispensable, it is useful for grain growth and in pile behavior. Such an improvement can easily be achieved.
NEXT PA«e(t>
ON THE SINTERING KINETICS IN UO2 XA9847843 A. MARAJOFSKY
Argentine Atomic Energy Commission, Buenos Aires, Argentina
Abstract
The fabrication process of UO2 pellets from powders involve pressing and a sintering anneal at high temperature ( 1650°C to 1750°C) during two or more hours in a hydrogen atmosphere. An alternative method is the oxidative sintering , made at lower temperature (1000°C to 1300°C) in a C02
or CO/C02 atmosphere. The sintering phenomena consist in the densification of the material by a thermal treatment below the fusion point. For a compact made by pressing a powder, sintering is the process of annulation of the porosity present in the compact or pellet. Several theories describe the sintering phenomena dividing it in three stages, initial, intermediate and final: in all of them the densification is a continuous growing function of time.
Nevertheless it has been experimentally reported that a reduction of the density occurs in the third step of the sintering. The phenomena has been called solarization.
Solarization has been attributed to the effect of the evolved gases from additives or to the CO2
atmosphere in oxidative sintering. Thus, it is convenient to distinguish between solarization in oxidative or reducing conditions.
Reducing solarization is a consequence of the tendency towards equilibrium of intergranular pores.
In oxidative sintering it occurs in the reducing anneal after the sintering and is due to the change in the lattice parameter.
This work shows examples of both types of solarization and qualitative interpretation of this phenomena. Both situations show the need of strict control of the sintering and powder production conditions.
1. Solarization
Solarization phenomena were observed and discussed by Amato et al. [1,2,3] and defined as a decrease in density of sinterized bodies, resulting from increasing time or temperature treatments. These authors attributed such phenomena to organic additives pressure within pores. Assmann et al.[4] attribute solarization to (i) volume transfer from closed fine to coarse pores through gas diffusion into the matrix according to concentration gradients which depend on the pore radius; (ii) transport from fine to coarse pores through grain-boundary diffusion, entailing both a decrease in density. In the case of oxidative sintering, these authors attribute solarization to the action of CO2.
2. Solarization during normal sintering
UO2 pellet manufacture is carried out through two-plateaus annealing, the first one between 400°C and 800°C, with a view to additive sublimation and/or stoichiometric uniformity; and the second one at approximately 1700°C for sintering itself (Fig 1).
Figure 2 shows the hydrostatic density of UO2 pellets sinterized at 1760, 1700, and 1650 °C for different periods, pressed from ex-AUC (uranyl ammonium tricarbonate) powders