**1. NAME AND TITLE**

MOCA: Monte Carlo Criticality Code System for Hexagonal Geometries.

**2. CONTRIBUTOR**

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.

**10. REFERENCE**

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.

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12. DATE OF ABSTRACT**

August 1991.

**KEYWORDS: ** COMPLEX GEOMETRY; MICROCOMPUTER; MONTE CARLO;
NEUTRON