[en] The PRIM integrated system providing for unification of the simulating model design, assembling and study as well as the results analysis using graphics is considered. It is designed for automated similation of continuous-discrete dynamic systems, includes the library of mathematical models for technological circuit elements, interactive subsystems for mathematical model construction and simulation initial condition formulation, subsystem for dynamic variable graphic visualization and monitor for simulating model composition. 3 refs

No abstract available

[en] Methodological problems on analysis of FBR efficiency with metal fuel using uranium and thorium in uranium-plutonium and mixed fuel cycle are considered. The investigations carried out permit to confirm that transition to application of LMFBR with metal fuel when accumulating in thorium screens of uranium-233 designed for application in WWER thermal power reactors permits to realize fuel self-providing condition for multi-component general purpose nuclear power in the framework of mixed fuel cycle with combined application of uranium and thorium. 10 refs.; 7 tabs

[en] An approach to calculation of thermohydraulic and strength characteristics of a fast reactor core is considered. The algorithms for calculating temperature profile for a reactor operating in a stationary regime with cyclic overloadings are witten. Criteria for reliability evaluation of fuel elements and assembly shrouds are discussed. The problem of determination of optimal coolant flow rate distribution over throttling zones is formulated. The analysis of the above model of the core thermohydraulic and strength characteristics shows that this problem can be solved without using general methods of nonlinear programming. The algorithms considered are assumed as a basis for the package of applied programs for calculation and optimization of fast reactors

[en] Short communication. 7 refs

[en] The mathematical model of fast power reactor core as a base of a packet of applied programs elaborated for the solution of problems of automation of serial design calculations carried out at the NPP designing stage and optimization of some reactor components of plant and the whole power plant is considered. Physical principles constituting a base for complex calculation of the fast reactor core are formulated. The algorithm of calculation of basic neutron physical performances of the reactor operating in a stationary mode of cyclic fueld over loadings is suggested. The formulas for calculating fuel component of given specific costs, doubling time and fuel reproduction coefficient are presented. The conclusion is drawn that the described algorithm gives a possibility to determine all the physical values being necessary for subsequent thermohydraulic and strength calculation (energy release depending on time in packets with various burnup depth neutron flux with more than 100 keV energy determining radiation swelling of construction materials), as well as values which can serve as a total function or boundary conditions for an external optimization problem (isotope mass balance, burnup depth, doubling time, fuel cost component)

[en] In this report the authors give the results of test model calculation for fast neutrons power reactor type BN-1600. The calculations are made in two dimensional geometry with an hexagonal array before and after refuelling

[en] A 3D model of large-power fast breeder power reactor is described in hexagonal geometry. A number of problems is formulated for carrying out test calculations: determination of 'wave-up' enrichments under steady-state refuelling conditions; temperature and power variations; determinations of reactivity effects depending on fuel burnup; determination of control rod efficiency; investigation of energy release field distribution at non-design position of burnup compensators. For the purpose of solving this problems, a 2D model of the reactor is presented in (R,Z) geometry which has been derived from the 3D model through homogenization of certain regions. The results of calculating for this model are presented. The calculations were performed in 2D (R,Z) geometry and in 2D and 3D geometry with hexagonal grid in few-group approximation. 8 refs.; 5 figs.; 18 tabs

[en] The report deals with application of calculation procedures which use a two-stage method for optimization of fast power reactor core parameters. The program package RBR-80 used here is designed for optimization of breeding in a fast power reactor. In particular, calculation cost was a major consideration in selecting the most desirable programme structure and procedures for optimization calculation. The first part of the report addresses the two-stage method, which is effective especially for reducing the calculation cost. The theoretical models for RBR-80 are described in the second part. The basic model and optimization model for a BN-type fast power reactor operating in an equilibrium state are given as various forms of approximated equations for distribution of neutrons and isotopes. The third part shows some algorithms designed for non-linear programming, centering on the application of an iterative linearization algorithm and another similar approximation algorithm. Practical calculation procedures are described in the fourth part. Two groups of data are used for the mathematical model of a fast neutron power reactor. One of them contains data on discrete range, physical constants, etc., while the other covers internal control vectors. Some results of actual calculations are presented in the final part of the report. (Nogami, K.)

[en] The report contains description of the method of construction of computer programs complexes for computation purposes for M-220 computers using the ALGOL-60 code for programming. The complex is organised on the modulus system principle and can include substantial number of modulus programs. The information exchange between separate moduli is done by means of special interpreting program and the information unit exchanged is a specially arranged file of data. For addressing to the interpreting program in the ALGOL-60 frameworks small number of specially created procedure-codes is used. The method proposed gives possibilities to program separate moduli of the complex independently and to expand the complex if necessary. In this case separate moduli or groups of moduli depending on the method of segmentation of the general problem solved by the complex will be of the independent interest and could be used out of the complex as traditional programs. (author)