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CCC-0431 MORSE-C.

MORSE-C, Neutron Transport, Gamma Transport for Criticality Calculation by Monte-Carlo Method

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1. NAME OR DESIGNATION OF PROGRAM:  MORSE-C.
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2. COMPUTERS
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Program name Package id Status Status date
MORSE-C CCC-0431/01 Arrived 08-FEB-2002

Machines used:

Package ID Orig. computer Test computer
CCC-0431/01 CDC 7600
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3. DESCRIPTION OF PROGRAM OR FUNCTION

MORSE-C is a Monte-Carlo code to solve the multiple energy group form of the Boltzmann transport equation in order to obtain the eigenvalue (multiplication) when fissionable materials are present. Cross sections for up to 100 energy groups may be employed. The angular scattering is treated by  the usual Legendre expansion as used in the discrete ordinates codes. Upscattering may be specified. The geometry is defined by relationships to general 1st or 2nd degree surfaces. Array units may be spcified. Output includes, besides the usual values of input quantities, plots of the geometry, calculated volumes and masses, and graphs of results to assist the user in determining the correctness of the problem's solution.
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4. METHOD OF SOLUTION:
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM:
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6. TYPICAL RUNNING TIME:
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7. UNUSUAL FEATURES OF THE PROGRAM

The code has been specially designed to solve neutron criticality problems.
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8. RELATED AND AUXILIARY PROGRAMS:
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9. STATUS
Package ID Status date Status
CCC-0431/01 08-FEB-2002 Masterfiled Arrived
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10. REFERENCES:
CCC-0431/01, included references:
- E.A. Straker, P.N. Stevens, D.C. Irving, and V. R. Cain:
  The MORSE Code - A Multigroup Neutron and Gamma-Ray Monte Carlo
  Transport Code
  ORNL-4585 (September 1970).
- M.B. Emmett:
  The MORSE Monte Carlo Radiation Transport Code System
  ORNL-4972/R1 (February 1983)
- T.P. Wilcox:
  MORSE-C, A CDC-7600 Program Designed to Solve Nuclear Criticality
  Problems by Using the Monte Carlo Neutron Method
  Informal Report UCID-18993 (January 1981)
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11. MACHINE REQUIREMENTS:
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
CCC-0431/01 FORTRAN-IV
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS:
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15. NAME AND ESTABLISHMENT OF AUTHORS

Contributed by: Radiation Safety Information Computational Center
                Oak Ridge National Laboratory
                Oak Ridge, Tennessee, U. S. A.
Developed by:   Lawrence Livermore National Laboratory, Livermore, CA, USA
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16. MATERIAL AVAILABLE
CCC-0431/01
source program   mag tapeSource Program (CHAT FORTRAN)              SRCTP
test-case data   mag tapeSample Problem Input                       DATTP
test-case output mag tapeSample Problem Output                      OUTTP
bin data lib     mag tapeN92GRP Library from LLNL ENDL Cmp.(11/1976)LBBTP
report                   ORNL-4972/R1  (February 1983)              REPPT
report                   UCID-18993  (January 1981)                 REPPT
report                   ORNL-4585 (September 1970)                 REPPT
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17. CATEGORIES
  • C. Static Design Studies
  • J. Gamma Heating and Shield Design

Keywords: Monte Carlo method, anisotropic scattering, criticality, cross sections, gamma radiation, multigroup, neutrons, one-dimensional, shielding, three-dimensional, time dependence, transport theory, two-dimensional.