COMPLETE DOCUMENT (1862 kb) - OECD Nuclear Energy Agency

On the basis of the study, it is not yet possible to select a single U-free material as the “best”
target. Interesting candidate materials are (Am, Zr)N solid solution, MgO-AmO 2-x , CERCER and
CERMET Am composites.
Comparative Dose Rate Values [72]
Storage of powders
Dose rates have been calculated for pure PuO 2 and for AmO 2 powders. The γ-dose rate due to
AmO 2 increases by a factor of 185 compared to PuO 2 , and the neutron dose by a factor of 3.
Compensating for these increases needs the addition of 7 cm lead to the shielding.
Handling and transportation of Pins
The γ-dose rate of AmO 2 is 2 780 times higher than that of conventional MOX fuel. For
neutrons, the ratio is about 7. Compensation needs typically the addition of about 4 cm lead and 4 cm
resin to the shielding is required, and compared with plutonium only half as many pins must be
transported at the same time.
2.2.2.2 Effect of curium on target fabrication
For one tonne of uranium loaded in a PWR, approximately 85 g of Cm are formed at the end
of a 45 GWd/t irradiation; the ratio Cm/Pu is about 0.8%. The isotopic composition of this curium
fraction (just after irradiation) is as follows:
242 Cm : 243 Cm : 244 Cm : 245 Cm : 246 Cm = 19 : 1 : 62 : 3 : 0.2.
The major effect of curium on the storage of nuclear wastes in the long-term is that the decay
of 244 Cm (T 1/2 = 18 years) adds some 2.5% to the 240 Pu quantities.
Both isotopes 242 Cm and 244 Cm are intense neutron sources. In case of a refabrication of fuel
or targets five years after core discharge, 242 Cm has mostly decayed while 244 Cm remains the
predominant neutron source.
The addition of Cm to the MOX fuel would increase the neutron dose rates around the
blending glove box by about a factor 100. This would require such thick protection layers (0.3 to 1 m
polyethylene shields) as to exclude its use in present MOX fuel production plants. Curium targets
containing gram quantities have been produced in laboratory conditions at ITU Karlsruhe.
2.2.2.3 Fabrication of targets for transmutation of fission products
The fabrication and irradiation of targets for transmutation of fission products has been
studied by the EFTTRA group in Europe [73-75]. The metallic form was selected for Tc. Rods of the
metal reduced from ammonium pertechnetate were prepared by arc melting, and casting in a
water-cooled copper mould. The targets have been irradiated in a thermal neutron flux in the High Flux
Reactor (HFR) in Petten, in a first test to a transmutation yield of about 6% and in a second test to a
yield of about 20%. Post-irradiation examinations of the first test have shown that technetium metal has
a good irradiation behaviour, showing negligible swelling and no microstructural changes [74]. The
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