Computer Programs

NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, CPU, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, OPERATING SYSTEM, ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES

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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 |
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ANISN-E | CCC-0082/008 | Tested | 01-FEB-1978 |

Machines used:

Package ID | Orig. computer | Test computer |
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CCC-0082/008 | IBM 370 series | IBM 370 series |

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3. NATURE OF PHYSICAL PROBLEM SOLVED

The ANISN system treats neutron and gamma transport in one- dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU.

ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point kernel) or for the preparation of multigroup libraries for two-dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors.

ANISN-E solves the same problems as the original ANISN code. Some modifications concern weighted cross sections output and fixed distributed sources input/output.

ANISN-E (CCC-0082/09): The CYBER 175 version of ANISN-E also contains the free-format input capability.

ANISN-JR extends the applicability of the original ANISN code for shielding analyses by adding options of calculating the reaction rates distributions from detector response, generating the volume- flux weighted cross sections in arbitrary regions or zones and plotting the neutron or gamma-ray spectra and the reaction rates distributions.

The ANISN system treats neutron and gamma transport in one- dimensional plane, spherical and cylinder geometry. The multigroup cross sections prepared by the programs LIANE and SUPERTOG are processed by the program RETTOG, which produces a binary library with Legendre expansions. The binary library can be updated and edited with the program LGR/B. The photon multigroup cross sections are created with the program GAMLEG/A. If the bulk of the data is too large, the program TAPEMA produces a special group-by-group library. The volume sources are calculated from a reduced set of input data and punched in a format suitable for input to ANISN, using the program PRESOU.

ANISN calculates fluxes by groups, space intervals, angle and any number of reaction rates. The energy and space dependent fluxes are stored on tape and can be reprocessed, edited and plotted with the program ANISEX, which also permits to calculate supplementary reaction rates. The program ANISN can condense cross sections into a reduced number of groups. The ANISN system is used as a reference system for the evaluation of approximation methods (space-diffusion or point kernel) or for the preparation of multigroup libraries for two-dimensional transport codes (DOT). In particular it is used for shielding problems with high attenuation in water reactors and fast reactors.

ANISN-E solves the same problems as the original ANISN code. Some modifications concern weighted cross sections output and fixed distributed sources input/output.

ANISN-E (CCC-0082/09): The CYBER 175 version of ANISN-E also contains the free-format input capability.

ANISN-JR extends the applicability of the original ANISN code for shielding analyses by adding options of calculating the reaction rates distributions from detector response, generating the volume- flux weighted cross sections in arbitrary regions or zones and plotting the neutron or gamma-ray spectra and the reaction rates distributions.

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4. METHOD OF SOLUTION

ANISN solves the one-dimensional Boltzmann transport equation for neutrons or gamma-rays in slab, sphere, or cylinder geometry. The source may be fixed, fission or a subcritical combination of the two. Criticality search may be performed on any one of several parameters. Cross sections may be weighted using the space and energy dependent flux generated in solving the transport equation.

ANISN-E : Besides diamond and weighted difference supplementary equations, exponential supplementary equations are available.

The new model:

(1) always gives positive solutions, without using any 'fixup' technique provided that the source is non-negative;

(2) allows, especially in deep penetration problems, the use of larger spatial meshes, hence requires shorter computer times than the ones requested by the diamond model combined with various types of fixup techniques or by weighted difference schemes to get the same accuracy;

(3) supplies solutions that are always reasonable overestimates of the exact solution.

In ANISN-JR, some optional functions are added to increase the utility of the code:

(1) print the total fluxes at boundary points of all mesh intervals. (The original ANISN prints the total fluxes at midpoint only.) (2) calculate, print and plot the lethargy width spectra. (3) print the angular fluxes at only required mesh boundaries or midpoints (maximum 10 points). The original ANISN prints at mid- point of all meshes, and therefore the number of print pages becomes vast according to the number of spatial and angular meshes.

(4) use the asymmetric quadrature set.

(5) calculate and plot the reaction rates for neutron and gamma-ray detectors, and collapse the response functions of detectors.

(6) generate volume-flux weighted cross sections for arbitrary zone or region. In the original ANISN, the cross sections can be collapsed only for a homogeneous zone or region.

(7) collapse into few group cross sections in ANISN, DOT, or TWOTRAN format. (In TWOTRAN format, the l-th Legendre coefficient of the scattering cross section is divided by (2l + 1) and the cross section of (n,2n) reactions is added for use of the coarse-mesh rebalancing technique.)

(8) multiply the average cross section by the density factor, when an option of density factors is used (IDFM=1).

ANISN solves the one-dimensional Boltzmann transport equation for neutrons or gamma-rays in slab, sphere, or cylinder geometry. The source may be fixed, fission or a subcritical combination of the two. Criticality search may be performed on any one of several parameters. Cross sections may be weighted using the space and energy dependent flux generated in solving the transport equation.

ANISN-E : Besides diamond and weighted difference supplementary equations, exponential supplementary equations are available.

The new model:

(1) always gives positive solutions, without using any 'fixup' technique provided that the source is non-negative;

(2) allows, especially in deep penetration problems, the use of larger spatial meshes, hence requires shorter computer times than the ones requested by the diamond model combined with various types of fixup techniques or by weighted difference schemes to get the same accuracy;

(3) supplies solutions that are always reasonable overestimates of the exact solution.

In ANISN-JR, some optional functions are added to increase the utility of the code:

(1) print the total fluxes at boundary points of all mesh intervals. (The original ANISN prints the total fluxes at midpoint only.) (2) calculate, print and plot the lethargy width spectra. (3) print the angular fluxes at only required mesh boundaries or midpoints (maximum 10 points). The original ANISN prints at mid- point of all meshes, and therefore the number of print pages becomes vast according to the number of spatial and angular meshes.

(4) use the asymmetric quadrature set.

(5) calculate and plot the reaction rates for neutron and gamma-ray detectors, and collapse the response functions of detectors.

(6) generate volume-flux weighted cross sections for arbitrary zone or region. In the original ANISN, the cross sections can be collapsed only for a homogeneous zone or region.

(7) collapse into few group cross sections in ANISN, DOT, or TWOTRAN format. (In TWOTRAN format, the l-th Legendre coefficient of the scattering cross section is divided by (2l + 1) and the cross section of (n,2n) reactions is added for use of the coarse-mesh rebalancing technique.)

(8) multiply the average cross section by the density factor, when an option of density factors is used (IDFM=1).

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8. RELATED AND AUXILIARY PROGRAMS

ANISN Library generators.

ANISN-JR uses, by input option, either group independent cross section sets produced by the code RADHEAT-V3, or those written in the original ANISN format. In the original ANISN and the present version, the adjoint calculations can not be performed with the group independent cross section tape (ID2=1) but with the tape generated from the step 3 of RADHEAT-V3. The data for the additional options are given before the ANISN original input data. If the reaction rates are required, the response functions of detectors follow after the ANISN data. A utility code of ANISN will be used for plotting the energy spectrum and flux or reaction rate distributions calculated by ANISN-JR.

ANISN Library generators.

ANISN-JR uses, by input option, either group independent cross section sets produced by the code RADHEAT-V3, or those written in the original ANISN format. In the original ANISN and the present version, the adjoint calculations can not be performed with the group independent cross section tape (ID2=1) but with the tape generated from the step 3 of RADHEAT-V3. The data for the additional options are given before the ANISN original input data. If the reaction rates are required, the response functions of detectors follow after the ANISN data. A utility code of ANISN will be used for plotting the energy spectrum and flux or reaction rate distributions calculated by ANISN-JR.

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10. REFERENCES

- R. Douglas O'Dell and Raymond E. Alcouffe:

Transport Calculations for Nuclear Analyses: Theory and Guidelines for Effective Use of Transport Codes

LA-10983-MS and UC-32 (September 1987).

- R. Douglas O'Dell and Raymond E. Alcouffe:

Transport Calculations for Nuclear Analyses: Theory and Guidelines for Effective Use of Transport Codes

LA-10983-MS and UC-32 (September 1987).

CCC-0082/008, included references:

- W.W. Engle, Jr. :A Users Manual for ANISN - A One-Dimensional Discrete Ordinates

Transport Code with Anisotropic Scattering

K-1693 (March 1967)

- R.W. Roussin:

Using ANISN to Reduce the DLC-2 100 Group Cross-Section Data to a

Smaller Number of Groups

ORNL-TM-3049 (May 7, 1969)

- W.W. Engle, M.A. Boling and B.W. Colston :

DTF II, A One-Dimensional Multigroup Neutron Transport Program

NAA-SR-10951 (March 1966)

- E. Sartori :

Lecture Notes on the Discrete Ordinates Transport

Codes ANISN & DOT.

"Course on Radiation Shielding Methods" Ispra (Nov. 20-24, 1978)

- P. Barbucci and F. Di Pasquantonio :

Exponential Supplementary Equations for SN Methods:

The One-Dimensional Case.

Reprint from "Nuclear Science and Engineering":

63, pp. 179-187 (1977)

- Enrico Sartori:

Note to all recipients of various versions of ANISN

NDB/93/0931 (27 August, 1993)

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CCC-0082/008

File name | File description | Records |
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CCC0082_08.001 | SOURCE PROGRAM (F4,EBCDIC) | 4178 |

CCC0082_08.002 | PROG. TO GENERATE DISTRI. SOURCE(F4,EBCDIC) | 16 |

CCC0082_08.003 | PROG. TO LIST GAMMA SOURCES (F4,EBCDIC) | 9 |

CCC0082_08.004 | SAMPLE PROBLEM 1 INPUT DATA | 41 |

CCC0082_08.005 | SAMPLE PROBLEM 2 INPUT DATA | 44 |

CCC0082_08.006 | JCL & INFORMATION | 64 |

CCC0082_08.007 | SAMPLE PROBLEM 1 PRINTED OUTPUT | 359 |

CCC0082_08.008 | LIST OF DISTRIBUTED SOURCE(SAMPLE PROB. 2) | 6 |

CCC0082_08.009 | SAMPLE PROBLEM 2 PRINTED OUTPUT | 441 |

CCC0082_08.010 | LIST OF GAMMA SOURCES | 16 |

Keywords: absorption, anisotropic scattering, buckling, cross sections, cylinders, discrete ordinate method, fission, gamma radiation, neutron transport theory, one-dimensional, shielding, slabs, spheres, transport theory.