P. A. Fomichenko

Results of Neutronic Benchmark Analysis for a High Temperature Reactor of the GT-MHR Type

The paper presents a description of benchmark cases, achieved results, analysis of possible reasons of differences of calculation results obtained by various neutronic codes. The comparative analysis is presented showing the benchmark–results obtained with reference and design codes by Russian specialists (WIMS-D, JAR-HTGR, UNK, MCU, MCNP5-MONTEBURNS1.0-ORIGEN2.0), by French specialists (APOLLO2, TRIPOLI4 codes), and by Korean specialists (HELIOS, MASTER, MCNP5 codes). The analysis of possible reasons for deviations was carried out, which was aimed at the decrease of uncertainties in calculated characteristics. This additional investigation was conducted with the use of 2D models of a fuel assembly cell and a reactor plane section.Copyright

The scheme for evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of power. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. To solve the closed nuclear fuel modeling tasks REPRORYV code was developed. It simulates the mass flow for nuclides in the closed fuel cycle. This paper presents the results of modeling of a closed nuclear fuel cycle, nuclide flows considering the influence of the uncertainty on the outcome of neutron-physical characteristics of the reactor.

A method for development of efficient 3D models for neutronic calculations of ASTRA critical facility using experimental information

The application of experimental information on measured axial distributions of fission reaction rates for development of 3D numerical models of the ASTRA critical facility taking into account azimuthal asymmetry of the assembly simulating a HTGR with annular core is substantiated. Owing to the presence of the bottom reflector and the absence of the top reflector, the application of 2D models based on experimentally determined buckling is impossible for calculation of critical assemblies of the ASTRA facility; therefore, an alternative approach based on the application of the extrapolated assembly height is proposed. This approach is exemplified by the numerical analysis of experiments on measurement of efficiency of control rods mockups and protection system (CPS).

Two-dimensional equations of the surface harmonics method for solving problems of spatial neutron kinetics in square-lattice reactors

Two-dimensional time-dependent finite-difference equations of the surface harmonics method (SHM) for the description of the neutron transport are derived for square-lattice reactors. These equations are implemented in the SUHAM-TD code. Verification of the derived equations and the developed code are performed by the example of known test problems, and the potential and efficiency of the SHM as applied to the solution of the time-dependent neutron transport equation in the diffusion approximation in two-dimensional geometry are demonstrated. These results show the substantial advantage of SHM over direct finite-difference modeling in computational costs.

Equations of the surface harmonics method for solving time-dependent neutron transport problems and their verification

Time-dependent equations of the surface harmonics method (SHM) are obtained for planar one-dimensional geometry. The equations are verified by calculations of test problems from Benchmark Problem Book ANL-7416, and the capabilities and efficiency of applying the SHM for solving the time-dependent neutron transport equation in the diffusion approximation are demonstrated. The results of the work show that the implementation of the SHG for full-scale computations will make possible substantial progress in the efficient solution of time-dependent problems of neutron transport in nuclear reactors.

An investigation of some models and approximations used in calculating fuel assemblies of a high-temperature gas-cooled reactor of the GT-MHR type

The paper describes basic neutron-physics models developed in the Division of Advanced Nuclear Power Systems of the Institute of Nuclear Reactors, Russian Research Center Kurchatov Institute, as design models intended for calculating the characteristics of block fuel assemblies of a high-temperature gas-cooled reactor GT-MHR, namely, models for calculating burnup of fuel and isotopes of burnable neutron absorbers and calculating fuel assemblies at fixed points with respect to burnup with preparation of the neutron constants in a preassigned number of energy groups for full-scale design of a reactor. A model problem for investigation of calculated approximations is proposed. The outcome of this investigation is a developed stage-by-stage procedure of preparing group homogeneous cross sections of a fuel assembly and its parts that has been introduced into the practice of design calculations of a GT-MHR reactor.

Profiling of energy deposition fields in a modular HTHR with annular core: Computational/experimental studies at the ASTRA critical facility

The paper presents the results obtained from the computational/experimental studies of the spatial distribution of the 235U fission reaction rate in a critical assembly with an annular core and poison profiling elements inserted into the inner graphite reflector. The computational analysis was carried out with the codes intended for design computation of an HTHR-type reactor.