[en] The method and the algorithm ensuing therefrom are described for the determination of the optimum operating state of a reactor. The optimum operating state is considered to be the extreme of the selected functional of the radial power distribution. The functional extreme is determined numerically, using a method which is one of the possible variants of the maximum gradient method. The radial distribution of the neutron absorption in regulating rods and the fuel element burnup are considered to be the variable parameters used in the optimization. (author)

[en] A heterogeneous reactor system is considered which is not in the critical state. The critical state may be effected by changing absorption in the regulating rods. The shape of the matrix operator describing the given reactor system depends on the physical properties of the system and among others on the absorption constants of the regulating rods. The task is solved by the introduction of a perturbation which converts the initial operator into an operator describing the critical state. The multi-parameter formulation of the problem of the reactor criticality has one great advantage, namely that in the suitable introduction of more parameters of criticality they may be assigned a measurable physical meaning, in this case the absorption of neutrons by selected regulating rods. A solution is given for one regulating rod (one cluster of regulating rods), for more than one regulating rods with the pre-set ratio of absorption constants and fcr more than one regulating rods with the optimization condition (the selection of a certain critical regime or critical state for which ksub(eff)=1). (B.S.)

[en] A matrix operator is considered describing a subcritical reactor based on a two-dimensional heterogeneous model. From the reactor criticality condition ensues the condition for the positive eigenvalue of the operator to which corresponds the positive eigenvector. From it ensues the formulation of the inverse problem for the eigenvalue of the operator describing the preset reactor state. The problem is solved by introducing a perturbation which changes the initial operator to one that describes the critical state. Several techniques are considered for introducing the perturbation. In introducing the perturbation the different physical properties of different types of rods are used. The advantage is considered of introducing the perturbation into the operator member corresponding to the action of absorption rods. (author)

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[en] The flow chart is given of the physics calculation of a heavy water reactor and its respective blocks are described. The methods and programmes of calculation developed during the A-1 power plant construction are outlined. The results are assessed obtained in correlating the A-1 theory and experiment. The results interesting for practical applications in reactor operation are indicated, such as compensation optimization or the temporal distribution of burn-up during reactor operation. (F.M.)

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[en] A number of programmes in two-dimensional geometry were written for the calculations of A-1 reactor criticality and for the assessment of experiments carried out on the TR-0 reactor. The programmes have been applied in calculating the effective critical heights of the hot A-1 reactor and for a partially filled core, the hot reactor reactivity, the thermal neutron flux distribution for the cold and the hot reactors, the distribution of heat generation in hot reactor fuel rods, and the hot reactor regulating rod efficiency. For calculations, a heterogeneous method has been used based on a two-dimensional reactor description which, as against commonly used methods also considers the anisotropy of diffusion in the radial and the axial directions, the dependence of resonance escape probability in a fuel element on its position in the lattice, and the nonzero fuel element radius. Some results are shown in graphs. (Z.M.)

No abstract available

[en] A numerical analysis of differences between the diffusion and the transport solution was carried out for WWER-1000 fuel assemblies in the 3-year fuel cycle. Ten alternatives of the fuel assemblies were considered. The precision of the diffusion solution was tested by using a data set calculated by the MICROBE code in the diffusion approximation, the transport calculations were tested by the Monte Carlo method using the MOCA code. The differences between the diffusion and the transport solution are within tolerable limits and decrease appreciably with increasing fuel enrichment. 27 tabs., 9 figs., 5 refs