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The results of studies of the physical and chemical properties of samples obtained during solidification of simulators of spent ion-exchange resin pulps with the use of polymer binders are presented. Investigations were conducted for pulps simulating heterogeneous radioactive wastes from the Kursk NPP. It is shown that the physical and chemical properties of the final compounds and the technological processes used to obtain them correspond to the requirements of solidified radioactive waste sent to storage or disposal. The surface of the samples is uniform and the components are distributed uniformly. Water resistance and radiation resistance tests showed that the mechanical strength of the tested samples is almost identical to the initial strength. It is shown that the mechanical load at which plastic deformation of the polymer compound occurs is a factor of two larger than and the leach rate does not exceed the established value.

The formation of tritium in reactors, its behavior in nitride fuel and possibility of bonding with fission products, and the characteristics of migration out of the fuel-element cladding are examined. It is shown by a computational method that owing to diffusion through the steel cladding the stationary content of tritium in a fuel element is established over several hours of operation and most of the formed tritium escapes through the cladding, some being retained in it. The presence of hydrogen (tritium) in structural reactor materials in the course of operation of the reactor and thermal vibrations of the structure can significantly degrade the strength of steel structural materials. In this connection, in the BREST facility, the emphasis on the problem of catching tritium is shifted from reprocessing of spent nuclear fuel to handling the lead coolant and the gas medium in the interior of the reactor.

The operation of a linear crystallizer, which performs crystallization extraction of uranium from a nitric acid solution of uranyl nitrate during nuclear reprocessing, is analyzed. Equations for the evaluation of the characteristics of the continuous operation of the apparatus are given, namely, the extraction efficiency of the crystal phase and the filling time of the storage container of the apparatus with the crystals of uranyl nitrate hexahydrate. The estimates are compared with the results of computer calculations and experimental data.

The USG-1 facility was developed to incinerate spent reactor graphite. It is equipped with a monitoring and control system; its maximum capacity is 10 kg/h. Trial combustion of a batch of unirradiated graphite was conducted. The tests showed that the units in the facility are serviceable and irradiated reactor graphite can be incinerated directly in an air flow. For a 1.5 kg load of graphite, the burn rate at 900°C was equal to 0.3 kg/h. The burn rate increases with increasing load. The incineration of irradiated graphite bushings of the EI-2 reactor in the USG-1 incinerator was studied at PDC UGR. The completeness of incineration was 95% and the 137Cs catching efficiency was 95%. With modification of the gas purification unit, the USG-1 incinerator can be used to burn irradiated graphite contaminated by traces of nuclear fuel and in monitoring the atmospheric emissions of 14CO2.

Salt cake radioactive waste is a remnant solid salt concentrate after deep evaporation of radioactive evaporator concentrate at WWER NPP’s. The traditional cementing of borate-containing liquid radioactive waste, to which the salt cake belongs, leads to a significant increase in the volume of the final product. This work describes borosilicate vitreous wasteforms developed to immobilize radioactive salt cake waste and comprises data on both glass synthesis and characterization. The composition of glass selected for the purpose of immobilisation of the salt cake radioactive waste allows to include up to 40 wt. % of the oxides contained in the salt cake and to reduce the volume of the final product by more than 2 times compared with the cement compound. The batches were melted in a cold crucible melter at 1200 °C. The normalized cesium leaching rate of the vitrified wasteform product was within range 3.0·10-5 – 3.7·10-6 g/(cm2·day).

A mathematical model of the crystallization of uranium in a continuous linear crystallizer, designed for the crystallization separation of desired products in the processing of an irradiated nuclear fuel, is proposed. This model defines the dynamics of growth/dissolution of uranyl nitrate hexahydrate crystals in a nitric acid solution of uranyl nitrate. Results of a numerical simulation of the indicated process, pointing to the existence of stationary conditions in the working space of the crystallizer, are presented. On the basis of these results, the characteristic time of establishment of the stationary regime at different parameters of the process was estimated. The mathematical model proposed was validated on the basis of a comparison of the results of calculations carried out within its framework with experimental data.

Crystallization refining is a technological operation that can separate and purity desired products (fissile materials) during the reprocessing of spent nuclear fuel. A mathematic model presented in this article describes the stationary operating regime of a linear crystallizer and makes it possible to determine the parameters of the process that ensure the most efficient separation of the desired products.

Russian Federation is developing the key technologies for closed nuclear fuel cycle with FR used mixed uranium-plutonium nitride or oxide (MOX) fuel within the framework of the project “Proryv”. This is a new integrated comprehensive product for nuclear power, which provides in the future - no accumulation of spent nuclear fuel; - radioactive waste management based on the principles of radiation-equivalent disposal; - technological support for the non-proliferation treaty; - competitiveness with other large-scale power technologies. The experimental and demonstration energy complex is under construction now on the site of Siberian Chemical Combine for the industrial development of the closed nuclear fuel cycle. ODEC will include a fast reactor BREST-OD-300 with lead coolant, facility for fabricating/refabrication of nitride fuel and the spent nuclear fuel processing facility. Up today, the key technology elements were demonstrated using macro quantities U, Pu, Np, irradiated mixed nitride fuel and real high-level waste. Within the framework of the project “Proryv” a prototype of the fuel cycle of fast reactors is developing.

The incineration of reactor graphite in an air stream is studied. Thermogravimetric analysis showed that in air disperse graphite starts to burn at 843 K. The burn rate reaches a maximum at 1023 K and remains constant above this temperature. The logarithm of the burn rate was found to be a linear function of the reciprocal of the temperature: logv = –(2.75 ± 0.35)·104/T. It was found that the burn rate at 1173 K per unit surface area for graphite samples with initial surface area from 600 to 1800 mm2 is almost the same for all samples and equals (0.74 ± 0.15)·10–3 kg/(m2·sec).

The precipitation of the dispersed phase on the bottom of a melter accompanying vitrification of model solutions of high-level wastes was analyzed and a procedure was proposed for determining the macroscopic parameter characterizing the diffusion of small inclusions of the dispersed phase. The maximum sizes of the conglomerates of the dispersed phase near the bottom of an EP-500 melter and the characteristic time over which the impurity is removed from the melt as a result of sedimentation of the heterogeneous phases were determined on the basis of the results. It was shown that these two characteristics of the dispersed phase have virtually no effect on the depth of the melt of the glass phase. For real furnaces used for vitrification with Joule heating, the results obtained could mean that in the hypothetical case when there is no convection and no temperature gradient in the melter, the precipitating dispersed phases based on fission products will settle almost entirely on the bottom of the melter in the form of a dense layer of precipitate.

There is particular importance in the storage of high-level wastes (for example, vitrified ones) in the stability of the solid matrix (glass) containing radionuclides. This paper examines the most rapid process in glass component redistribution near inhomogeneities (inclusions, pores, cracks, etc.), as these inevitably arise during vitrification. The authors consider two model problems, whose solutions enable one to estimate the changes in component concentrations near inhomogeneities.

In the mode—mode coupling approximation the critical exponents (CE) are calculated for the dampings and frequency shifts of soft optic modes and acoustic phonons (in uniaxial ferroelectrics) and for the dampings of soft acoustic phonons (in cubic crystals) in the structural transition point. The CE are traced back to the anharmonic interaction (of all orders) of soft phonons with the intrinsic dynamic central peak of any nature. [Russion Text Ignored]

A scratching cone can be considered as a knife edge with a negative leading angle; a conical indentor rises towards the surface of the metal on scratching with an unchanged normal load on account of work-hardening in the metal and of frictional forces, which have a component in opposition to the vertical load.

The work shows that the friction coefficient μ during slip of polycrystalline electrolytic corundum, single-crystal silicon carbide and sintered corundum over hardened ShKh15 steel, and the wear m in an electrolytic corundum test-piece according to slip speed have a low point at speeds close to 100 m/sec. A hypothetical explanation is given for this behavior of μ and m.

Dry friction of steel on steel at high pressures and sliding speeds takes place without any noticeable oxidation of the rubbing surfaces and is accompanied by intense seizing and wear. The products of wear are not oxidized and appear as shiny metal particles [1].