Computer Programs

NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROGRAM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHORS, MATERIAL, CATEGORIES

[ top ]

[ top ]

To submit a request, click below on the link of the version you wish to order. Rules for end-users are
available here.

Program name | Package id | Status | Status date |
---|---|---|---|

MOCA | IAEA1238/01 | Tested | 08-SEP-1994 |

Machines used:

Package ID | Orig. computer | Test computer |
---|---|---|

IAEA1238/01 | PC-80286 | PC-80486 |

[ top ]

3. DESCRIPTION OF PROGRAM OR FUNCTION

Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly.

Criticality problem in neutron transport for hexagonal fuel assembly in VVER nuclear reactor. The assembly is assumed to be either arranged in an infinite hexagonal array or placed in vacuum. The problem is solved in three- dimensional geometry, using standard energy group formalism and assuming that effective scattering cross sections are presented as Legendre polynomial expansions. The code evaluates ten different physical quantities, e.g. multiplication factor, neutron flux per energy group and spatial zone, integrated over angle and power in any zone of the assembly.

[ top ]

4. METHOD OF SOLUTION

Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks:

In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables.

The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated.

In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations.

Monte Carlo method of successive generations is applied. Computation proceeds according to an analog random process. The code is organized into three blocks:

In the first block, the input data are converted to quantities for use in the Monte Carlo calculation. An initial neutron distribution is calculated, which corresponds to a fission spectrum uniform in spatial and angular variables.

The main calculations are carried out in the second block (subroutine PROC2). This block is subdivided into geometrical and physical parts. Neutron tracks in individual zones and groups as well as probabilities for the formation of secondary neutrons are calculated.

In the third block (subroutine PROC3), the results are evaluated statistically. Effective multiplication coefficients, the neutron flux per group and zone, and respective errors are computed. These quantities serve as a basis for the evaluation of other quantities. The results are either printed or stored for future evaluations.

[ top ]

5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and array dimensions. (See pp 8-14 of th manual.)

In the PC version of the program, the maximum number of neutrons is 1000, the maximum number of energy groups is 4, and the maximum number of material compositions is 15. Angular expansion of scattering cross sections is allowed up to P10. These restrictions can easily be removed by increasing input parameters and array dimensions. (See pp 8-14 of th manual.)

[ top ]

6. TYPICAL RUNNING TIME

No general estimate was made. Running time for the sample problem on an 80386-based desktop computer with 80387 math coprocessor is 90 minutes.

No general estimate was made. Running time for the sample problem on an 80386-based desktop computer with 80387 math coprocessor is 90 minutes.

IAEA1238/01

Sample case DATA about 35 seconds with the Microsoft version and 16 the Lahey version. The sample case on the report UJV 6487 R took about 18 minutes on both versions.[ top ]

[ top ]

[ top ]

IAEA1238/01, included references:

- Jan Kyncl:The Code MOCA

UJV 6487 R (March 1983).

[ top ]

IAEA1238/01

NEA-DB tested the program on a PC DELL 40486 (33-MHz) with 8 MB of RAM.[ top ]

IAEA1238/01

MS-DOS 6.2 with compilers MS-FORTRAN 5.1, and Lahey F77L-EM/32 FORTRAN 77 VERSION 5.20.[ top ]

[ top ]

[ top ]

IAEA1238/01

File name | File description | Records |
---|---|---|

IAEA1238_01.001 | Information file | 169 |

IAEA1238_01.002 | Microsoft executable | 0 |

IAEA1238_01.003 | Lahey executable | 0 |

IAEA1238_01.004 | Source file AZIRN.FOR | 12 |

IAEA1238_01.005 | Source file CTI.FOR | 6 |

IAEA1238_01.006 | Source file EXPRNF.FOR | 15 |

IAEA1238_01.007 | Source file FLTRNF.FOR | 8 |

IAEA1238_01.008 | Source file GTISO.FOR | 15 |

IAEA1238_01.009 | Source file MAIN.FOR | 19 |

IAEA1238_01.010 | Source file OPFILE.FOR | 14 |

IAEA1238_01.011 | Source file PIS.FOR | 6 |

IAEA1238_01.012 | Source file PROC1.FOR (Microsoft version) | 600 |

IAEA1238_01.013 | Source file PROC11.FOR | 123 |

IAEA1238_01.014 | Source file PROC2.FOR (Microsoft version) | 890 |

IAEA1238_01.015 | Source file PROC3.FOR | 595 |

IAEA1238_01.016 | Source file RANDU.FOR | 10 |

IAEA1238_01.017 | Source file SFLRAF.FOR | 5 |

IAEA1238_01.018 | Source file PROC1.LAH (Lahey version) | 620 |

IAEA1238_01.019 | Source file PROC2.LAH (Lahey version) | 911 |

IAEA1238_01.020 | Input data (case on disk) | 50 |

IAEA1238_01.021 | Input data (case on report) | 50 |

IAEA1238_01.022 | Output data for input DATA | 711 |

IAEA1238_01.023 | Output data for input DATA2 (Ms. version) | 585 |

IAEA1238_01.024 | Output data for input DATA2 (Lh. version) | 696 |

IAEA1238_01.025 | DOS file-names | 24 |

Keywords: Monte Carlo method, criticality, fuel assemblies, hexagonal lattices, neutron transport equation, three-dimensional.