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Experimental results of investigations of pyrochemical conversion of weapons plutonium into plutonium oxide for fabricating fast-reactor fuel are presented. Weapons plutonium was hydrogenized by hydrogen at 220°C, after which the plutonium hydride obtained was acidified at 550-560°C with the formation of PuO 2 . To increase fire and explosion safety of the process, a mixture of oxygen with nitrogen, helium, or argon was used or nitriding with nitrogen and oxidation of plutonium mononitride were introduced. The particle size of the plutonium oxide powders obtained was less than 15 µm. The powders showed poor flowability, but after granulation they were suitable for fabricating kernels with mixed fuel. The gallium was removed by reduction of Ga 2 O 3 by hydrogen to Ga 2 O, which was sublimated. The mixed-fuel kernels sintered at 1600-1700°C in a hydrogen atmosphere contained <0.001 wt.% gallium, and their density was 94-97% of the theoretical value.Two variants of the technology were studied to obtain ceramic type plutonium dioxide: pyrochemical with only gallium removed from the plutonium in the process and combining the pyrochemical process of converting weapons plutonium into a disperse powder of plutonium hydride or plutonium mononitride by dissolving them in acids and extraction removal of gallium and americium. The simplest, most cost-effective and efficient method of converting weapons plutonium into plutonium oxide is the pyrochemical method with pre-hydrogenation followed by oxidation of plutonium hydride [1].In connection with the decision to use mixed fuel in BN-800 and other reactors, the conversion of weapons plutonium into a disperse plutonium oxide powder, which is suitable for fabricating mixed fuel, becomes urgent [2]. The initial plutonium and uranium oxide powders, from which after mixing, granulation, and pressing blanks kernels were sintered in a hydrogen-containing atmosphere at 1600-1750°C. Gallium can be removed in this variant of fuel fabrication [3,4].Americium is removed from plutonium oxide in order to bring excess plutonium into the fuel cycle and use it to fabricate mixed fuel for power reactors, first and foremost, thermal reactors, provided that the process is radiologically safe [2,5].

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:

The main results of a study of regimes for dissolution of plutonium oxide and mononitride powders, obtained by the pyrochemical method from weapons plutonium, in acids for subsequent extraction removal of gallium, americium, and ballast impurities and obtaining ceramic-type plutonium oxide powder suitable for fabricating mixing oxide fuel are presented. It is established experimentally that plutonium oxide and mononitride obtained by the pyrochemical method dissolve rapidly in the acid mixture HNO 3 12 moles/liter and HF 0.1 moles/liter. Plutonium extraction into solution reaches >99%.Americium must be removed from oxide fuel which is to be used in thermal reactors. This operation can be combined with the removal of gallium and ballast impurities by the Purex process, which by now has been well-perfected. The properties of plutonium dioxide which is obtained by oxidizing plutonium hydride or nitride have been little studied. Specifically, there are no data on the behavior of plutonium oxide and plutonium mononitride, obtained by the pyrochemical method, in solutions of nitric acid.For this reason the following were done: 1) the dissolution of plutonium dioxide and mononitride in solutions of nitric acid and mixtures of nitric and hydrofluoric acids was studied to determine the optimal technological parameters (the concentration of the nitric acid, the volume/mass ratio, the solution temperature, the composition of the gas phase, undissolved residue);2) the possibilities were determined of the formation of pyrophoric compounds in the undissolved residues after plutonium dioxide and nitride have dissolved; and3) the regimes of plutonium leaching from undissolved residues were perfected. The present work is devoted to studying the combined process which incorporates the pyrochemical synthesis of plutonium oxide or mononitride powder from weapons plutonium, their dissolution in acids, obtaining plutonium solutions for the standard extraction purification and preparation of plutonium dioxide powders (Fig. 1). A previous study of the dissolution of mixed uranium and plutonium mononitride and monocarbide pertained to compact fuel kernels with density 85-95% of the theoretical value [1,2].

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