Results of U0.55Pu0.45N and U0.4Pu0.6N mixed mononitride fuel tests in a bor-60 reactor to burnup 12% h.a.

Abstract

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 0.4 Pu 0.6 N and U 0.55 Pu 0.45 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.Fast reactors with dense uranium-plutonium fuel are required to create safe nuclear power on a large scale. Mixed mononitride fuel best meets the requirements for the fissile element content per unit volume and the thermal conductivity as well as the mechanical, radiological, technological, and thermochemical properties [1][2][3][4][5]. It has been successfully tested in our country, France, USA, Japan, and Great Britain at linear power density 350-1300 W/cm. The fuel density in the experiments was 83-95% of the theoretical value. Pellet fuel was used in the main. Burnup 16% h.a. in a thermal reactor was achieved [6][7][8][9][10][11]. Mononitride fuel was prepared from initial oxides and metals. Reactor tests established satisfactory compatibility of the mixed mononitride fuel with cladding made of austenitic and ferritic-martensitic steels. The carbon and oxygen content recommended on the basis of a generalization of the research results for such fuel is no more than 0.15% each [8][9][10][11][12][13][14][15][16][17]. As a result of accumulation, there arises the problem of efficient and cost-effective validated recovery of plutonium in fast reactors.The objective of the present work was to investigate the possibility of reaching maximum burnup and retaining seal tightness under irradiation in a BOR-60 reactor of fuel elements with a helium sublayer and high-purity U 0.55 Pu 0.45 N and U 0.4 Pu 0.6 N mononitride fuel with linear power density 450 and 540 W/cm, respectively. It was necessary to determine the following:

Incorporation of Kr and Xe in Uranium Mononitride: A Density Functional Theory Study

“…For the safe and efficient use of nitride fuels, it is therefore important to have a comprehensive understanding of their properties, including the generation and migration of fission products in the fuel matrix, especially as post-reactor studies using (U, Pu)N mixed fuel found relative yields of 19 and 19.3% of Kr and Xe, respectively, in the fuel-clad gap. 16 However, compared with UO 2 , gas release models for UN are rudimentary. There are several theoretical works on the behavior of Kr and Xe in UN, 17−20 but they are based on different methods and the results are not consistent.…”