TREND IN PLUTONIUM CONTENT OF MOX IN THERMAL REACTOR USE AND IRRADIATION BEHAVIOR OF MOX WITH HIGH PLUTONIUM CONTENT

摘要

The uranium enrichment of UO2 fuel for the current power reactors, both PWR and BWR, tends to increase because of increasing burn - up target. The plutonium content of MOXfuel used in thermal reactors shall be determined in order to have reactivity worth equivalent to enriched UO2 fuel based on physical accounting method for adjusting fissile enrichment, thus the plutonium content tends to increase according to the increment of the uranium enrichment of UO_2 fuel and this trend shall further be accentuated due to the fact that Pu recovered from reprocessing of the spent high burnup U2 fuel contains less fissile isotopes. The plutonium content is calculated by use of the physical accounting method with the plutonium having several kinds of isotope ratios and the calculation results indicate that the plutonium content in MOX will evolve to ratios in excess of 10%. It shall be, therefore, important to know the irradiation behavior of MOX with high plutonium content of more than 10 wt%. MOXfuel rods having a plutonium content of about 14 wt% and fabricated by use of MIMAS process have been irradiated under PWR conditions in the Belgian test reactors BR - 3 and BR - 2. The peak fuel rod burn - up of the fuel rods studied in this paper ranges from 31 to 37 GWd/t - HM and their average burnup is about 22 - 26 GWd/t - HM with the rod averaged linear heat generation rate of about 15 - 21 kW/m. The MOX rods are investigated by destructive and non - destructive post irradiation examinations and some of them are now continued to be irradiated in BR - 2. Mixed Oxide (U, Pu) O2 fuel produced by the MIMAS process results in a fine dispersion ofPu enriched particles in a UO2 matrix and effectively gives three enrichment classes: low, medium and high enriched. The high enriched particles (often called "Pu spots"), have an enrichment of around 25 wt% Pu, the low enriched phase is the UO2 matrix and contains only trace amounts of Pu. An intermediately enriched phase is formed by interdiffusion of U and Pu. The extent of the intermediately enriched phase depends on the fabrication process (powder type, mixing method, sintering conditions etc.), and continues to further develop in - pile. One of the objectives of the present study is to investigate the fission gas release and to improve the thermo - mechanical modeling of these high enriched fuels. The MOX rods investigated here show higher fission gas release as compared with fission gas release meared from fuels with lower enrichment and reported in the open literature, but since the latter data stem from different fuels, irradiated under different conditions, it is not possible yet to draw conclusions about underlying mechanisms or systematic trends of increased fission gas release when enrichment goes up.