Calculations of spatial and energy distribution of neutrons in reactor core and shielding. Part 2

Special procedures have been developed for fast computations with modest demands on computer memory. The principle underlying the so-called removal method is the multigroup approach, separation of primary fluxes and their calculation using the integration of point kernels. Included in the primary fluxes are non-scattered neutrons and small-angle scattered neutrons. Primary flux neutrons are the source for solving scattered radiation fluxes which are calculated in the diffusion approximation or in the P1-approximation. A number of one- and two-dimensional programs have been written based on these methods. The determination is described of cross sections and group constants, and problems of numerical calculations and the method of kernels are dealt with. Macrocomputations of the reactor core are performed by few-group diffusion programs (the BIPR, MOBY-DICK codes), computations of spectra in cells by MICROS, MICROBE-02 program packages. Sources of errors in the calculation of shielding are given. (J.P.)