1. NAME AND TITLE
MOCA: Monte Carlo Criticality Code System for Hexagonal Geometries.
Nuclear Research Institute, Rez, Czechoslovakia.
3. CODING LANGUAGE AND COMPUTER
FORTRAN 77; IBM PC/AT, IBM PS/2 Model 60.
4. NATURE OF PROBLEM SOLVED
This criticality code system for hexagonal geometries was used to calculate neutron flux in the Soviet VVER reactor assembly in steady state.
5. METHOD OF SOLUTION
The Monte Carlo Method of successive generations is used. MOCA solves the linear Boltzmann transport equation without external source. The time-independent neutron transport equation is solved in three-dimensional geometry for the hexagonal reactor assembly in an infinite hexagonal lattice. MOCA calculates the multiplication factor, flux distribution and other neutron quantities.
6. RESTRICTIONS OR LIMITATIONS
Only the region of the fuel assembly is considered. The following limits are set but can be adjusted by modifying input parameters: neutrons per generation 100, energy groups 4, material compositions 9. P10 expansion of the scattering cross sections to angular variable is allowed .
7. TYPICAL RUNNING TIME
The sample problem has a time cutoff of 15 minutes.
8. COMPUTER HARDWARE REQUIREMENTS
The code runs on IBM PC/AT and IBM PS/2 Model 60.
9. COMPUTER SOFTWARE REQUIREMENTS
The code is written in FORTRAN 77. The compiler is Microsoft FORTRAN Version 4.01 or 5.0 under the DOS 3.3 operating system.
Jan Kyncl, ``The Code Moca,'' Informal Document, Nuclear Research Institute, Rez, Czechoslovakia (March 1983).
11. CONTENTS OF CODE PACKAGE
Included are the referenced document and one DS/HD 5.25-inch diskette (1.2 MB), containing the source code, sample input and output.
12. DATE OF ABSTRACT
KEYWORDS: COMPLEX GEOMETRY; MICROCOMPUTER; MONTE CARLO; NEUTRON