ECRAN PROGRAM FOR COMPUTATION OF EFFECTIVE FACTOR OF NEUTRON MULTIPLICATION IN REACTOR FACILITIES

A.N. Moskvin, A.V. Nikiforova, A.P. Trubitsyn, G.G. Farafontov, R.M. Shagaliev, V.A. Shumilin
VANT. Ser.: Mat. Mod. Fiz. Proc 1996. Вып.3. С. 35-39.

      This program is designed for computations of the effective neutron multiplication factor and neutron spectrum in multigroup two-dimensional kinetic approximation in cylindrical and plane geometries. The following iterative techniques of computation of effective neutron multiplication factor are implemented in the program: Kellog technique, direct iterative technique, method of fission source iteration and combined technique. This paper compares efficiency of these methods on test problems for various types of reactors (WWER, RBMK, SNR). To speed up simple iteration convergency within one group the consistent FCA-method is used. To speed up external iteration convergency in the fission source iteration method, the Chebyshev technique and successive upper relaxation techique are used.

      The database needed for numerical simulation of processes of binary fission fragment accumulation in power nuclear reactor fuel is discussed. The total amount of a great quantity of fragments is suggested to replace with the simplified model. Recommendations on using estimated characteristics of neutron interaction with separated fragments are given. For a great number of fission nuclear tables are presented showing yields of separated fragments, isotopes, total yields of light and heavy stable slug fragments and yields of effective light and heavy stable slug fragments. CONVERT and SUMMER programs are presented which are needed for extraction of the required values from estimated data and calculation of characteristics of neutron interaction with efficient light and heavy stable slug fragments. Algorithm implemented in SUMMER program for computation of characteristics of neutron interaction with effective fragments is given.

      Recommendations on neutron-nuclear data selection for numerical simulation of actinide nuclei accumulation in energy reactor fuel are given. ENDL-82, JENDL-3, ENDF/B-6 and BROND-2 libraries of estimated data on neutron interaction with nuclei of ^232,233Th, ^231-233Pa, ^232-239U, ^237-239Np, ^236,238,239Pu, ^241-244Am, ^242m,244mAm and ^242-249Cm are considered. RESINT program for processing data arrays corresponding to above listed libraries which was used to conduct attendant computations is presented.