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Lanthanum orthophosphate with the monazite structure was proposed on examinations as a suitable matrix for immobilization of future americium-containing liquid wastes, which could be formed in conversion of metallic plutonium into oxide at PA “Mayak.” Specimens of monazite non-active ceramics were fabricated from LaPO4 powders obtained using a thin-film evaporator by either hot-pressing or cold-pressing and sintering at 900–1300 °C. According to electron microprobe analysis (EMPA), scanning electron microscopy (SEM), and X-ray diffraction (XRD), which were used for characterization of produced samples, all specimens did not contain any phase other than the monoclinic monazite phase. Ceramics having the specific activity of Am-241 2.13×107 Bq/g were prepared by only cold-pressing with subsequent sintering at 1300°C during 1 hour. The normalized leach rates of lanthanum and americium in distilled water at 90°C were less than 1.2×10−4 and 2.3 10−4 g/m2×day, respectively.

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For Abstract see ChemInform Abstract in Full Text.

A graphic comparison is presented for results of analysis of stresses determined from approximateanalytical solution and the finite-element method using the CAN program. The validity of results derived from analytical calculations is characterized by the magnitude of the residual of the equilibrium equations of an infinitely small region of the median surface of structural elements, which solve the problem of the differential and boundary equations. The trial example makes it possible to confirm the performance of programs employed for calculation of the SSS of geometrically complex shell structures.Vessels having a complex annular geometry in the form of conjoined shells of revolution of cylindrical, conical, plane, or convex configuration are used in radiochemical productions. Requirements relative to the preservation of integrity and a nuclear-safe geometry under thermomechanical service loads are established for these vessels.In ensuring safety indicators for the equipment in question, as determined by strength calculations, it is necessary to evaluate the validity of results of stress determinations, which are approximate in nature during analysis of geometrically complex designs. In this connection, a trial example of the solution of the static edge problem of the mechanics of the axisymmetric deformation of an annular vessel subject to excess internal pressure and temperature variation of the body was formulated within the framework of the theory of thin plates and shells. The example consists of an approximate-analytical solution, and makes it possible to substantiate the performance of programs used for calculation of the stress-strain state (SSS) of shell structures by the finite-element method. Figure 1 shows a schematic diagram of an annular vessel with semi-toroidal bottoms. Let us adopt the following subscripts for the structural elements (see Fig. 1a): AB and DE for the shells i = 1 and i = 2, respectively; CB and CD for the segments of the upper bottom j = 1 and j = 2, respectively; and FA and FE for the segments of the lower bottom j = 3 and j = 4, respectively. To describe the geometry of the structure and the mechanics of its deformation, let us introduce designations (see Fig. 1b) for the median surfaces of an arbitrarily selected ith cylindrical shell (CS) and jth segment of a bottom. The angle ϕ j (the determining position of a point on the meridian of the bottom) within the interval from 0 to π/2 is referenced from the vertical axis counterclockwise around the center of curvature of the meridian for bottom segments j = 1, 4, and clockwise for bottom segments j = 2, 3.

The potential harm of long-lived radionuclides estimated as the product of the activity of a radionuclide and the dose coefficient is examined. The potential harm of actinides in high-level wastes is calculated taking account of the harm due to their decay products. At the same time, an analogous calculation is performed for the uranium isotopes 238 U, 235 U, and 234 U consumed in the reactor. The value obtained, which depends on the time, can be regarded as the averted harm. The time for establishing radiation equivalence between the high-level wastes and the consumed uranium is determined as the time in which the potential harm from actinides becomes equal to the averted harm. It depends on the holding time of the spent fuel before radiochemical reprocessing. For a 5 yr holding period, it is ~49000 yr.In [1], it was proposed that the potential harm due to long-lived radionuclides in high-level wastes be estimated as the product of the activity of a radionuclide by its dose coefficient, showing the dose which an adult person would obtain if 1 Bq of this radionuclide entered the stomach. Six radionuclides were separated from high-level wastes which remain dangerous to humans for periods longer than 100,000 yr: 99 Tc and 129 I among fission products and 239 Pu, 240 Pu, 241 Am, and 243 Am among transuranium elements [2]. However, this conclusion was drawn neglecting the products of decay of long-lived radionuclides and the holding time of spent nuclear fuel before radiochemical reprocessing. This deficiency is eliminated in the present paper.In a reactor, not only are dangerous radionuclides produced but uranium isotopes 238 U, 235 U, and 234 U are consumed, which results in a decrease of the overall radiation hazard as a result of these radionuclides and the products of their decay. Consequently, for the radioisotopes 238 U, 235 U, and 234 U consumed in a reactor it is desirable to perform a similar calculation of the radiation hazard, which can be regarded as an averted harm, which depends on the time. Then the time for establishing radiation equivalence can be determined as the time when the potential hazard from radionuclides in high-level wastes becomes equal to the potential hazard of the consumed uranium radionuclides, i.e., to the averted harm.Actinides in High-Level Wastes. Table 1 gives the composition of the spent nuclear fuel from a VVÉR-440 reactor with initial enrichment 3.6% and burnup 33.4 kg/ton [3]. Although other degrees of enrichment and burnup are used at the present time, we shall consider these conditions to be model conditions which make it possible to develop a scheme for calculating the potential hazard of long-lived radionuclides from high-level wastes and radionuclides which have been burned up in a nuclear reactor. In so doing, we shall take account of the fact that in radiochemical reprocessing of spent nuclear fuel 0.01% of the uranium, 0.025% of the plutonium, and 0.5% of the neptunium [4] and all isotopes of americium and curium

Interest in the transmutation of long-lived radioactive elements into stable or short-lived elements by means of nuclear reactions has increased sharply in the last few years. Specialists who support the practical application of methods for burning up radionuclides agree that efforts must first be focused on transuranium elements (237Np, 244Cm, 241Am, 241pu) as well as long-lived fission products, such as t29I and 99Tc [l, 2]. Radiation burnup of Np, Am and Cm and fission products with long half-lives is possible only after they are separated from the main mass of the spent fuel, including uranium and plutonium. Besides the main problem of eliminating an ecologically dangerous, long-lived radionuclide, transmutation of technetium will make it possible to obtain more stable ruthenium, which can then be used for industrial purposes.In studying methods for separating technetium from spent nuclear fuel, it becomes obvious that calcium pertechnetate obtained on an industrial scale (only at the Industrial Association NMayak') requires additional purification (even though it is of high radiochemical purity) in order to meet the requirements imposed on transmuted target preparations.According to the technical conditions, the content of radioactive impurities in 99Tc preparations should not exceed 0.024% of its own radioactivity. Radioactive impurities in technetium compounds were monitored on an ionizing-radiation spectrometer. In the course of this work new methods of preparation and analysis of especially pure technetium preparations were developed and tested.Data on the radiocbemical purity of industrial preparations of technetium are given in Table 1. The data were obtained by using methods which were developed for analytical monitoring.One can see from Table 1 that the main impurities in the 99Tc preparations are 98"1"c, l~ 137Cs, 238pu, 241Am, 239pu, 9~ Their content is at the level of 3.10 -4 of the radioactivity of 99Tc (radiochemical purity 99.9997%). Under production conditions, technetium preparations of higher purity cannot be obtained by technological schemes.To increase the radiochemical purity of technetium preparations, the initial potassium pertechnetate must be additionally purified by means of an optimal technological scheme for converting KTcO 4 into NH4TcO 4, which is the most acceptable form of the material for subsequent production of technetitun metal. When ammonium pertechnetate is reduced by hydrogen, only gaseous products and technetium metal are formed: NH4TcO 4 + 2H 2 ---, I/2N 2 + 4H20 + To. The conversion of KTcO 4 into NH4TcO 4 is possible on a KU-2 cation exchanger in an H + form followed by neutralization of the technetium acid by ammonia. Ammonium pertechnetate easily crystallizes from water solutions, which makes it possible to obtain this compound with a high purity.Extraction, ion-exchange, and precipitation methods for concentrating and removing from technetium the accompanying elements are well known and are described in detail in the literature. These methods can also be combined....