RESULTS OF U~(0.55)Pu~(0.45)N AND U~(0.4)Pu~(0.6)N MIXED MONONITRIDE FUEL TESTS IN A BOR-60 REACTOR TO BURNUP 12% h.a.

摘要

The results of post-reactor studies of U~(0.55)Pu~(0.45)N and U~(0.4)Pu~(0.6)N mixed mononitride fuel elements (density 85% of the theoretical value) and a helium sublayer are presented. The fuel elements are irradiated in a BOR-60 reactor to burnup 9.4 and 12.1% h.a., respectively, with power density 430 and 540 W/cm. All fuel elements remained hermetic; the ChS-68 steel cladding (20% cold deformation) retained excess plasticity. The maximum zone of interaction between the cladding and the fuel and fission products did not exceed 15 μm. The swelling rate of U~(0.55)Pu~(0.45)N and U~(0.4)Pu~(0.6)N fuel was 1.1 and 0.68%/% burnup, respectively. The gas release did not exceed 19.3 and 19%. The steel damage dose was 43 dpa. The character of the porosity distribution in the fuel affects the swelling and gas release.