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CCC-0082 ANISN.

ANISN-E, 1-D Transport Program ANISN with Exponential Model
ANISN-JR, 1-D Transport Program ANISN with ZZ JSD Data and Flux Plot

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1. NAME OR DESIGNATION OF PROGRAM:  ANISN.
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2. COMPUTERS
To submit a request, click below on the link of the version you wish to order. Only liaison officers are authorised to submit online requests. Rules for requesters are available here.
Program name Package id Status Status date
ANISN-E CCC-0082/08 Tested 01-FEB-1978

Machines used:

Package ID Orig. computer Test computer
CCC-0082/08 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.
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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).
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

In ANISN, the complexity of the problem is limited by storage size.
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6. TYPICAL RUNNING TIME
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7. UNUSUAL FEATURES OF THE PROGRAM

ANISN-E: Exponential equations to compute mesh centre fluxes.
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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.
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9. STATUS
Package ID Status date Status
CCC-0082/08 01-FEB-1978 Tested at NEADB
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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).
CCC-0082/08, 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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11. HARDWARE REQUIREMENTS: MACHINE REQUIREMENTS
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
CCC-0082/08 FORTRAN-IV
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13. SOFTWARE REQUIREMENTS: OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED
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14. ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

ANISN supersedes DTF-II (NAA-SR-10951, March 25, 1966) which followed a series of developmental efforts over a period of years.
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15. NAME AND ESTABLISHMENT OF AUTHOR

Contributed by: Oak Ridge National Laboratory
                    Oak Ridge, Tennessee, U.S.A.

   ANISN-E :  P. Barbucci and F. Di Pasquantonio:
              ENEL Centro di Ricerca termica e nucleare
              Bastioni di Porta Volta 10
              20121 Milan, Italy

   ANISN-JR :  Japanese Atomic Energy Research Institute
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16. MATERIAL AVAILABLE
CCC-0082/08
File name File description Records
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
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17. CATEGORIES
  • C. Static Design Studies
  • J. Gamma Heating and Shield Design

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