Vol 127, No 1 (2019)

When fuel rods melt during a severe accident, the movement of the structural and fuel materials along fuel assemblies can be spatially non-uniform, so that a three-dimensional thermohydraulic model is implemented in the EVKLID/V2 code as part of the HYDRA-IBRAE/LM module. The transport of the same sets of components and their mixtures as in the one-dimensional version of the HYDRA-IBRAE/LM module can be calculated in the model: liquid sodium, sodium vapor or vapor-gas mixture of sodium vapor and non-condensable gases, solid lead, liquid lead, solid uranium dioxide, liquid uranium dioxide, hard stainless steel, liquid stainless steel, steel vapor. The three-dimensional model is implemented in a cylindrical coordinate system, which makes it easier to include the geometric dimensions and parameters of a particular fuel assembly (number and diameter of fuel rods, lattice pitch, and others) and the core. A description is given of the basic system of equations describing the motion of the components of the destroyed core in the three-dimensional r–z–φ geometry, and its numerical realization. Examples of test calculations showing the serviceability of the model are presented.

The main results of experimental validation of the operability of passive core cooling systems of nextgeneration VVER are presented. It is shown that the residual heat release can be removed in 24 h on account of the operation of the second-stage water-storage system and the passive heat-removal system. As part of the validation of VVER-TOI passive safety systems the thermophysical properties of boric acid (density and viscosity) were measured experimentally with parameters characteristic for a reactor emergency. Empirical relations satisfactorily describing the experimental data are obtained.

The behavior of mixed uranium-plutonium nitride fuel and uranium nitride was studied by means of thermogravimetry at high temperatures (to 2173 K) in a helium flow. The mass loss of the nitride samples was found in the low temperature range (<1773 K), which is not associated with decomposition of uranium or plutonium mononitrides. The mass loss occurs in two stages and is accompanied by the release of nitrogen. It is shown that nitride fuel can contain up to several percent of uranium sesquinitride U₂N₃, which decomposes in the indicated temperature range and can strongly affect pellet integrity and fuel-rod life during operation.

The purpose of this work is to investigate the sorption of americium from model solutions of liquid radwastes using modified solid-phase extracting agents based on TODGA as an alternative method of additional purification of liquid radwaste by removing transuranium elements by means of alkali precipitation. The kinetic parameters of uranium and americium sorption from highly saline weakly acidic (pH ~3.4) model solutions of liquid radwaste with americium content ~0.04 mg/liter were determined for three modified samples synthesized at VNIIKhT and one analog of a TODGA sample from the AXION Company. The time to equilibrium of all experimental samples in terms of americium is almost the same and equals ~240 min. The equilibrium concentration of uranium is reached in approximately 800 min. The salting-out effect in the extraction of uranium and americium from salt solutions is revealed, though this effect is much weaker for uranium than for americium. The most promising sorbent was determined – sample 7. Its distribution factor is 1.4 times higher than that of the AXION sample and the pair distribution factor Am/U is a factor of 2 higher.

The purpose of this work is to analyze the use of detecting materials in radiation monitors as well as the replacement of widely used ³H-based neutron counters by neutron-detection scintillation technology. The replacement of helium counters is a consequence of two factors: the lack of ³He and widespread use of ³He-based counters in safety equipment, such as volumetric neutron detectors. Selection criteria for evaluating promising technologies are used in this work, specifically, high absolute neutron detection efficiency – efficiency at least 1.5 counts in 1 sec in detecting 1 ng ²⁵²Cf at distance of 2 m in a 20 mm thick moderator and low sensitivity to γ-ray detection – γ-ray detection efficiency not exceeding 10^–6 with irradiation by a 0.1 μSv/h γ-ray source. Since they can have a large sensitive area and high resolution, scintillation detectors are now being proposed as alternatives to helium counters. But it is necessary to find an optimal scintillator possessing simultaneously low sensitivity to γ-radiation and to choose an optimal method of measuring information. Promising neutron detection technologies based on the glasses Li₂OSiO₂:Ce³⁺, LiF/ZnS(Ag⁺), Li₆Gd(BO₃)₃:Ce, Cs₂LiYCl₆(Ce) (CLYC) as well as EJ-254 boron-doped plastic are examined from the standpoint of the posed problems.

As part of the work on eliminating the nuclear legacy temporary storage facilities for high-level solid wastes were opened, radiometric inspection of the wastes was performed, wastes were sorted, control rods of ship reactors were identified in the wastes and removed, and spectrometric determination of the induced activity in each rod was made. Forty-four rods with activity totaling 1.35·10¹³ Bq were packaged and shipped to a long-term storage site. The activity distribution was determined from samples and will be taken into account in planning subsequent work on the handling of such wastes and picking the most appropriate type of packaging. The work was performed using remote-control machines in compliance with the standards and regulations for the radiation safety of personnel.

Neutronics measurements are regularly conducted during operation in order to evaluate the physical characteristics of the core. A nonstandard reactimeter connected to a backup neutron detector is used to perform the measurements. A calculation of the reactivity on the basis of the indications of a single detector results in distortion of the true value of the reactivity because of spatial effects. A method of increasing the accuracy of the reactivity calculations by taking account of the changes in the efficiency of the ionization chamber is proposed. Software for performing the required calculations has been developed. The correctness of the proposed solution was checked and confirmed by experiments performed on a critical stand at OKBM Afrikantov using the core of an icebreaker reactor.