Boron carbide and boron steel are used for control rods in BBER reactors. Investigations have shown that their operation is limited to a burnup of up to 45-50% t~ This is due to the large amount of radiation damage to the absorbing materials as a result of the formation of gaseous products of nuclear reactions, which results in strong swelling of the materials and destruction of the control rods. The maximum burnup l~ hi boron ~arbide Dr the absorbing eleraeats of a BBER-1000 reactor is reached within three years of operation of the automatic tonal rod~..Control rods based on n, ~, absorbers are promising for thermal reactors. Gaseous pmdtm,'~ of-tmc~.m' ,re, actions are not formed in such rods, and the rods are characterized by a higher radiation resistance. Of,a, -~, ,absorbers, ~aly europium oxide and europium oxide in a metallic matrix are used in reactors built in this country. It has'been confu'med erperimentally that the operating time of control rods based on europium oxide absorbers can exceed 20 yr. :But, despite the high radiation resistance and neutron absorption efficiency with prolonged operation, europium based materials cannot be recommended for widespread use in reactor construction because of the high induced "r activity and the slow decrease :in this activity, which creates serious problems during storage and utilization of spent parts and can result in catastrophic con~quences in the ease of an accident. Therefore a search is being made for other radiation resistant n, 7r absorbers. These iinclude dysprosium titanate and hafaium.This paper reports the results of investigations of the radiation resistance of hafnium amd dyslmasium titanate a~ well as the main criteria which the absorbing materials must satisfy in order to ensure that the comical rode.can,operate for :~ long tinle.Choice of Criteria. Allowing for the increasing requirements on :safety ard x, eliabiI'm/,of almxation~ ~ arollo~ing criteria for the working capacity of the control rods were adopted:at least 8-10% reactivity margin; not more than 10% decrease in the efficiency of control rods at the .end of,~ration; shape and integrity preservation during operation; not more than 0.5 % increase in the diameter of the claddings of the control rods; strucatrally stable material for the absorbing core, not subjected to phase transitions with a large change in volume (not more than 10-15%); ultimate strength of carrying structural elements of at least 200 MPa, and relative elongation not exceeding 0.5%; rate of interaction of the absorbing and structural materials of not more than 0.001 mm/yr; and, rate of corrosion of absorbing and structural materials in the coolant medium of not less than 0.01 ram/yr. The first four criteria are the main criteria. If one of these criteria is not satisfied, then in order to show that the control rod is operable, specific features of the construction and characteristics of the reactor, in which it is used, must be specially substantiated. The next four are also extremely important. They make it possible...

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The present work reviews the primary benefits of radioactive isotope production in the fast neutron reactors the accumulated foreign and local experience in usage of this type of reactors for radioactive isotope production, and also the potential new trends in development of the existing technologies. The main advantage of isotope production in the fast neutron reactors is their high capacity and lower value in general as compared with the accelerating installations. One nuclear reactor may produce dozens of different isotopes including Cj-60, Sr-89 (targets on the base of yttrium), Ni-63 (targets on the base of copper), Eu-152, Eu-154, Sm-145, Sn-117n, Cu-64 and others. It should also be emphasized that a number of radioisotopes are most efficiently produced only in fast and intermediate neutron spectra, which is not possible in nuclear thermal reactors and, above all, in research reactors. Studies have shown that the organization of large-scale production of radioisotopes can significantly increase the economic attractiveness of fast-neutron reactors.

Two limiting regimes of gas release from particles are studied: diffusion and activation. The conditions under which either the diffusion or the activation desorption regime occurs are determined. A model is proposed for the activation release of helium atoms from centers of accumulation and diffusion of these atoms toward the surface of a spherical particle of boron carbide powder. A quantitative estimate is obtained for the release of helium from a material during annealing, and the influence of the size of the powder grains and the rate of change of the temperature of the material on the thermodesorption of helium from irradiated boron carbide is estimated.Information reported on the destruction of the cladding of absorbing elements caused by excess gas pressure [1, 2] dictates the need for developing a computational program, based on this information, for simulating the behavior of (n, α) absorbers and rods in the control and protection system under irradiation, specifically, the development of mathematical models of helium release from boron carbide powder. The amount of helium released from irradiated boron carbide by heating has been determined experimentally many times [3][4][5], but there is still no satisfactory model for describing this process adequately [6]. A method for estimating the amount of helium released from irradiated boron carbide during annealing under nonstationary conditions has been proposed [7]. The model was constructed on the basis of a theory of annealing of radiation defects [8] under the assumption that the limiting stage of helium release is overcoming an activation barrier when gas atoms detach from accumulation centers. However, this mechanism meglects the influence of helium diffusion toward the surface of a powder particle after the atoms detach from the accumulation center. It would be useful to develop a mathematical model that would make it possible to generalize the activation mechanism of helium release [7] so as to include diffusion transfer.Boron carbide powder is now used in absorbing kernels in the VVÉR-1000 control and protection system. The particle size is about 10 µm, which is close to the travel distance of a thermal neutron in boron carbide. Consequently, it can be assumed that accumulation centers capture helium uniformly over the entire volume of a particle.A mathematical model of helium release from a powder particle in the process of annealing was constructed under the following assumptions:• all of the helium in the irradiated boron carbide is located in stationary accumulation centers, and the number of helium atoms occupying interstices is negligibly small compared with the total number in a particle; • when the powder is being heated helium atoms are released from all capture centers by an activation mechanism with a low probability of helium being captured by an accumulation center;

Models of absorbing elements with a promising material for the control organs of nuclear reactors have been tested in the SM reactor -pelleted and powder kernels with different composition based on dysprosium hafnate in a mixture with boron carbide. The neutron fluence with energy >0.1 MeV averaged over a kernel volume was (0.9-1.3)·10 22 cm -2 at the moment the tests were completed for different samples. The temperature at the center of the kernels of the absorber element models during irradiation was 620-1100°C in channel No. 4 and 400-500°C in channel No. 9. The results of the materials science studies show that on the whole the serviceability of the absorbing elements based on pellets and powders of dysprosium hafnate is high.The search for materials for the control organs of nuclear reactors is proceeding in the direction of increasing the effectiveness, reliability, and safety of the materials which remain unchanged over a long period of service. In 2005-2007, a technology was perfected on the basis of the Federal Targeted Program "National Technological Resources," an experimental batch of high-density (7.7 -8.2 g/cm 3 ) pellets based on dysprosium hafnate with 23-75 mol.% Dy 2 O 3 and fluorite single-phase structure was prepared, and the batch was tested in an SM reactor. The permanently high effectiveness due to the presence of two absorbers with 13 daughter isotopes, relatively cheap raw material, high melting temperature (2700°C), and radiation resistance make Dy 2 O 3 ·HfO 2 a promising material (Fig. 1) [1].The absorbing materials (B 4 C, Dy 2 O 3 ·TiO 5 , Dy 2 O 3 ·HfO 2 , and others) in the form of VVER absorbing elements were tested and certified in the high-temperature water circuit VP-3 of the SM reactor with test parameters as close as possible to the operating parameters. This stage is a necessary condition for validating the adoption of a new material.For irradiation in SM, the Nos. 4 and 9 reflector channels with neutron flux (energy ≥0.1 MeV) density 0.32·10 15 and 0.66·10 14 sec -1 ·cm -2 , respectively, were connected to the VN-3 circuit facilities of high parameters [2]. An irradiation facility consisting of six perforated, steel, cylindrical containers 12 mm in diameter and 0.5 mm thick with models of the absorbing elements was placed into the channels. A 100-mm long model of an absorbing element contains an 8.2 mm in diameter and 0.45-mm thick 08Kh18N10T steel shell, sealed by welding on end pieces, and pellet or powder (vibrational compacted mixture of powders, sieve with cell size ≤0.0125 mm), and a 74 ± 1.5 and 40 mm high kernel, respectively, used in real parts. The kernel is compacted in helium-filled cladding by a gas-permeable wad made from a pressed nickel sieve (Fig. 2). A list of the absorbing element models tested and the main characteristics of the kernels are presented in Table 1.We shall examine the models and methods for calculating the neutron-physical conditions of irradiation of absorbing materials in the channels of an SM reactor.A numerical model was created on the ...