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

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Program name | Package id | Status | Status date |
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HEXANN-EVALU | NEA-1125/01 | Tested | 01-DEC-1993 |

Machines used:

Package ID | Orig. computer | Test computer |
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NEA-1125/01 | CDC CYBER 170 | CRAY Y-MP |

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3. DESCRIPTION OF PROGRAM OR FUNCTION

HEXANN-EVALU calculates the neutron irradiation of a pressure vessel surrounding a nuclear reactor core composed of hexagonal assemblies. The area outside the core may contain hexagonal shielding assemblies of non-multiplying materials, a core liner and annular material zones.

HEXANN-EVALU calculates the neutron irradiation of a pressure vessel surrounding a nuclear reactor core composed of hexagonal assemblies. The area outside the core may contain hexagonal shielding assemblies of non-multiplying materials, a core liner and annular material zones.

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4. METHOD OF SOLUTION

The Monte Carlo method is used. The neutrons start at the core boundary with a given distribution in space, angle and energy. The angular distribution is calculated by HEXANN itself from data describing the spacial distribution of the neutron source in the core. Survival biasing is used in all collisions. To increase efficiency the following options are included: regionwise importance, Russian roulette, low energy Russian roulette, splitting, path stretching with explicit exponential transform and automatic importance correction.

The Monte Carlo method is used. The neutrons start at the core boundary with a given distribution in space, angle and energy. The angular distribution is calculated by HEXANN itself from data describing the spacial distribution of the neutron source in the core. Survival biasing is used in all collisions. To increase efficiency the following options are included: regionwise importance, Russian roulette, low energy Russian roulette, splitting, path stretching with explicit exponential transform and automatic importance correction.

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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

30-degree symmetry assumed

Maximum number of energy groups = 30

Maximum group number change in downscattering = 19

No upscattering

Maximum number of materials = 10

Maximum number of cylindrical surfaces = 20

Maximum number of source faces (hexagon faces at edge of core in 30-degree sector)= 20

Vacuum boundary conditions are used at the top, bottom and outer boundary

30-degree symmetry assumed

Maximum number of energy groups = 30

Maximum group number change in downscattering = 19

No upscattering

Maximum number of materials = 10

Maximum number of cylindrical surfaces = 20

Maximum number of source faces (hexagon faces at edge of core in 30-degree sector)= 20

Vacuum boundary conditions are used at the top, bottom and outer boundary

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7. UNUSUAL FEATURES OF THE PROGRAM

The history tracing and estimation are handled by separate programs (HEXANN and EVALU, respectively). The user specifies those cylindrical surfaces or regions where estimation is to be performed, and every time an assigned surface is crossed, or a free flight or collision is played in an assigned region, a record containing all the information relevant to the event is written on the "History Tape", which passes information from HEXANN to EVALU. EVALU prints either the flux as such (in a number of groups, if desired) or some reaction rate obtained by weighting the flux with an energy-dependent function. Axial and/or azimuthal flux or reaction rate distributions can be obtained (but not distributions varying independently in both the azimuthal and axial coordinates; the dependence on these is assumed to be separable). To get different reaction rates, run EVALU repeatedly.

The history tracing and estimation are handled by separate programs (HEXANN and EVALU, respectively). The user specifies those cylindrical surfaces or regions where estimation is to be performed, and every time an assigned surface is crossed, or a free flight or collision is played in an assigned region, a record containing all the information relevant to the event is written on the "History Tape", which passes information from HEXANN to EVALU. EVALU prints either the flux as such (in a number of groups, if desired) or some reaction rate obtained by weighting the flux with an energy-dependent function. Axial and/or azimuthal flux or reaction rate distributions can be obtained (but not distributions varying independently in both the azimuthal and axial coordinates; the dependence on these is assumed to be separable). To get different reaction rates, run EVALU repeatedly.

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8. RELATED AND AUXILIARY PROGRAMS

The program TRIGON-PVND provides the neutron source distribution in the horizontal plane needed to provide the azimuthal position and angular distribution of starters. Cross section data from the BUGLE-80 library are prepared by the program BUGLER. The axial distribution and energy spectrum of neutrons leaving the core can be calculated by any appropriate program.

The program TRIGON-PVND provides the neutron source distribution in the horizontal plane needed to provide the azimuthal position and angular distribution of starters. Cross section data from the BUGLE-80 library are prepared by the program BUGLER. The axial distribution and energy spectrum of neutrons leaving the core can be calculated by any appropriate program.

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NEA-1125/01, included references:

- Ivan Lux:HEXANN-EVALU - A Monte Carlo Program System for Pressure Vessel

Neutron Irradiation Calculation

VTT-TUTK-210 (August 1983).

- Notes on the Use of HAXANN-EVALU.

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Package ID | Computer language |
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NEA-1125/01 | FORTRAN-IV EXTENDED |

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NEA-1125/01

File name | File description | Records |
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NEA1125_01.001 | Information file | 78 |

NEA1125_01.002 | JCL for first case (NEADB) | 15 |

NEA1125_01.003 | HEXANSC HEXANN in symbolic form | 3214 |

NEA1125_01.004 | EVALUSC EVALU in symbolic form | 1341 |

NEA1125_01.005 | XLIB26 Cross section lib. 26-group | 3121 |

NEA1125_01.006 | HEXTES1 JCL 26-group test case 1 (author) | 30 |

NEA1125_01.007 | HEXTES2 JCL 26-group test case 2 (author) | 34 |

NEA1125_01.008 | HEXTEST jcl 7-group test case (author) | 23 |

NEA1125_01.009 | HEXIN1 HEXANN input 26-group test case 1 | 103 |

NEA1125_01.010 | HEXIN2 HEXANN input 26-group test case 2 | 103 |

NEA1125_01.011 | ZHEXTES HEXANN input 7-group test case | 122 |

NEA1125_01.012 | EVALDP1 EVALU input dpa eval. test case 1 | 15 |

NEA1125_01.013 | EVALDP2 EVALU input dpa eval. test case 2 | 15 |

NEA1125_01.014 | ZEVATES EVALU input dpa eval 7-group t.c | 12 |

NEA1125_01.015 | EVALFM1 EVALU input 54Fe(n,p)54Mn case 1 | 15 |

NEA1125_01.016 | EVALFM2 EVALU input 54Fe(n,p)54Mn case 2 | 15 |

NEA1125_01.017 | EVALFX1 EVALU input flux eval. case 1 | 7 |

NEA1125_01.018 | EVALFX2 EVALU input flux eval. case 2 | 7 |

NEA1125_01.019 | HEXOUT1 Output of first run | 2982 |

NEA1125_01.020 | HEXOUT2 Output of second run | 378 |

NEA1125_01.021 | HEXOUT Output of 7-group test case | 857 |

Keywords: Monte Carlo method, neutron flux, pressure vessels, reaction rates.