Behaviour of CS and I In Zirconia Based Fuel

AbstractA zirconia based ceramic is foreseen as an inert matrix fuel for burning excess plutonium in light water nuclear reactors. For reactor safety reasons the behaviour of volatile fission products such as cesium and iodine must be studied since a retention of fission products is favourable for licensing the studied inert matrix fuel. In this study, implantation of Cs and I was performed into polycrystalline (Zr0.85, Y0.15)O1.925 samples. The implantation depth was selected on the basis of the ability to observe by Rutherford backscattering spectroscopy (RBS) the behaviour of Cs and I after treatment. With a 1 MeV incident energy, the ions are implanted at a depth of 200 nm as predicted by TRIM. After implantations full quantification of I and Cs concentration profiles was performed by RBS. The implantation profiles are measured as a function of sample temperature during stepwise heating programs. It is interesting to observe retention of Cs and I at relatively high temperature (e.g. for 2 h, below 900 K for Cs and below 1400 K for I). This behaviour is likely to be due to the size and interactions of these species in the zirconia solid solution.

Download Full-text

Zirconia Inert Matrix Fuel for Plutonium and Minor Actinides Management in Reactors and as an Ultimate Waste Form

MRS Proceedings ◽

10.1557/proc-1104-nn03-01 ◽

2008 ◽

Vol 1104 ◽

Cited By ~ 3

Author(s):

Claude Degueldre ◽

Wolfgang Wiesenack

Keyword(s):

Near Field ◽

Fission Products ◽

Minor Actinides ◽

Spent Fuel ◽

Geological Disposal ◽

Inert Matrix Fuel ◽

Inert Matrix ◽

Fuel Pellets ◽

Reducing Conditions ◽

Deep Geological Disposal

AbstractA plutonia stabilised zirconia doped with yttria and erbia has been selected as inert matrix fuel (IMF) at PSI. The results of experimental irradiation tests on yttria-stabilised zirconia doped with plutonia and erbia pellets in the Halden research reactor as well as a study of zirconia solubility are presented. Zirconia must be stabilised by yttria to form a solid solution such as MAz(Y,Er)yPuxZr1-yO2-ζ where minor actinides (MA) oxides are also soluble. (Er,Y,Pu,Zr)O2-ζ (with Pu containing 5% Am) was successfully prepared at PSI and irradiated in the Halden reactor. Emphasis is given on the zirconia-IMF properties under in-pile irradiation, on the fuel material centre temperatures and on the fission gas release. The retention of fission products in zirconia may be stronger at similar temperature, compared to UO2. The outstanding behaviour of plutonia-zirconia inert matrix fuel is compared to the classical (U,Pu)O2 fuels. The properties of the spent fuel pellets are presented focusing on the once through strategy. For this strategy, low solubility of the inert matrix is required for geological disposal. This parameter was studied in detail for a range of solutions corresponding to groundwater under near field conditions. Under these conditions the IMF solubility is about 109 times smaller than glass, several orders of magnitude lower than UO2 in oxidising conditions (Yucca Mountain) and comparable in reducing conditions, which makes the zirconia material very attractive for deep geological disposal. The behaviour of plutonia-zirconia inert matrix fuel is discussed within a burn and bury strategy.

Download Full-text