|Program name||Package id||Status||Status date|
|Package ID||Orig. computer||Test computer|
|NEA-1896/01||Linux-based PC,UNIX W.S.|
MOSRA-SRAC is a lattice calculation module of the Modular code System for nuclear Reactor Analyses (MOSRA). This module performs the neutron transport calculation for various types of fuel elements including existing light water reactors, research reactors, etc. based on the collision probability method with a set of the 200-group cross-sections generated from the Japanese Evaluated Nuclear Data Library JENDL-4.0. It has also a function of the isotope generation and depletion calculation for up to 234 nuclides in each fuel material in the lattice. In these ways, MOSRA-SRAC prepares the burn-up dependent effective microscopic and macroscopic cross-section data to be used in core calculations.
A neutron cross-section library based on the latest and well validated Japanese Evaluated Nuclear Data Library JENDL-4.0 is available for more than 400 nuclides.
The routine of the collision probability method applicable to 16 types of geometries in covers lattice calculations for most of existing reactors.
The effective cross-sections by the conventional table look-up method based on the narrow resonance (NR) approximation can be replaced with those by an optional routine PEACO which solves a multi-region lattice problem by the collision probability method using an almost continuous (hyper-fine) energy group structure for the resonance energy region. The interaction of resonances can be accurately treated by the PEACO routine.
Arbitrary temperature of composite materials is allowed by the interpolation of resonance shielding factors and thermal scattering matrices. For the PEACO routine, Doppler broaden cross-sections in hyper-fine energy structure are internally calculated from the point-wise cross-sections at room temperature.
The Dancoff correction factor required in the interpolation of the self-shielding factors of resonant nuclides is automatically calculated by the installed collision probability routines. The factor is given not for an absorber lump but for each constituent nuclide in the lattice which contains a resonant nuclide in two or more materials with different compositions.
A doubly heterogeneous system can be solved by successive lattice calculations since homogenizing and collapsing of macroscopic cross-sections is carried out separately. Especially, the resonance absorption of which double heterogeneity effect should be solved simultaneously, can be treated as far as the microscopic lattice can be approximated by one of 1-dimensional lattices. This method is effective, for example, in a fuel assembly lattice of the high-temperature gas cooled reactor including many coated-particle fuels.
The option for the lattice burn-up calculation provides burn-up dependent microscopic/ macroscopic cross-sections and the change of nuclide composition during burn-up by a series of procedures to get a neutron spectrum, to get one-group effective cross-sections and to calculate generation and incineration of nuclides at each burn-up step. It can treat a detailed burn-up chain model which includes 28 heavy nuclides and 201 fission products or burnable poisons.
Spatial homogenizations and energy-group collapsing (from 200-groups until one-group) can be carried out any number of times, although some functions (e.g. use of PEACO, burn-up calculation) may be suppressed.
There are several restrictions on maximum number of regions, materials, nuclides. The maximum array dimension sizes are defined by parameter statements. The parameter values can be easily changed and a new executable can be generated by using equipped “Makefile”.
|Package ID||Status date||Status|
|Package ID||Computer language|
All source programs of MOSRA are written by FORTRAN77 except for some minor subroutines to get time or date. MOSRA-SRAC can be installed on the machines whose OS type is UNIX or similar ones (Linux, Cygwin) with FORTRAN and C compilers.
MOSRA-SRAC has been developed on 32 bit machines and compilers. There is no experience on 64 bit computers.
The execution tests have been done on the following compilers:
1) GNU FORTRAN77 and C compilers (g77 + gcc) :
OS Ubuntu 10.04.4 LTS 32bit (Linux) + gcc version 3.4.6
OS x86_64 GNU/Linux(64bit) + gcc version 3.4.6
OS Cygwin-6.1 (on Windows 7) i686 Cygwin + gcc version 3.4.4
2) Fujitsu FORTRAN77 and C compilers (frt + fcc):
OS Ubuntu 10.04.4 LTS 32bit (Linux) + Fujitsu Fortran Compiler
Driver Version 5.0
3) PGI FORTRAN77 and C compilers (pgf77 + gcc)
OS Ubuntu 10.04.4 LTS 32bit (Linux) + PGI compiler pgf77
11.10-0 32-bit target on x86 Linux
4) Intel FORTRAN and GNU compilers (ifort + gcc):
OS x86_64 GNU/Linux (64bit) + Intel compiler ifort version 13.1.3
and GNU compiler gcc version 3.4.6
Note: "gfortran" have not worked well.
Keywords: JENDL-4.0, burnup calculation, collision probability, effective cross section, lattice calculations, neutron transport.