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NESC0298 GGC4

GGC-4, MultiGroup Neutron Spectra and Broad Group Cross-Sections Calculation, P1, B1, B2, B3 Approximation

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1. NAME OR DESIGNATION OF PROGRAM:  GGC4
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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
GGC4-MAKE NESC0298/01 Tested 01-SEP-1976
GGC-4 NESC0298/02 Tested 01-NOV-1974
GGC-4 NESC0298/03 Tested 01-FEB-1970

Machines used:

Package ID Orig. computer Test computer
NESC0298/01 IBM 370 series IBM 370 series
NESC0298/02 CDC 6600 CDC 6600
NESC0298/03 IBM 360 series IBM 360 series
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3. DESCRIPTION OF PROBLEM OR FUNCTION

The GGC4 program solves the multigroup spectrum equations with spatial dependence represented by a single input buckling. Broad group cross sections (shielded or unshielded) are prepared for diffusion and transport codes by averaging with the calculated spectra over input-designated energy limits. The code is divided into three main parts. A fast (GAM) section which covers the energy range from 14.9 MeV to 0.414 eV, a thermal (GATHER) section which covers the energy range from 0.001 to 2.38 eV, and a combining (COMBO) section which combines fast and  thermal cross sections into single sets. Basic nuclear data for the  fast section which consists of fine group-averaged cross sections and resonance parameters is read from a data tape. The fine group absorption and fission cross sections may be adjusted by performing  a resonance integral calculation. Utilizing a fission source and an  input buckling, the code solves the P1, B1, B2, or B3 approximation  to obtain the energy-dependent fast spectrum. Two or six spatial moments of the spectrum (due to a plane source) may also be evaluated. Instead of performing a spectrum calculation, the user may enter the Legendre components of the angular flux directly. For  as many input-designated broad group structures as desired, the code calculates and saves (for the combining section) spectrum-weighted averages of microscopic and macroscopic cross sections and transfer  arrays. Slowing down sources are calculated and saved for use in the lower energy range. Given basic nuclear data, the thermal section of GGC4 determines a thermal spectrum by either reading it as input, by calculating a Maxwellian spectrum for a given temperature, or by an  iterative solution of the P0, B0, P1, or B1 equations for an input buckling. Time moments of the time and energy-dependent diffusion equations are calculated (as an option) using the input buckling to  represent leakage. Broad group cross sections are prepared by averaging fine group cross sections over the calculated spectra. Broad group structures are read as input. The combining section of GGC4 takes the broad group-averaged cross sections from the fast and thermal portions of GGC4 and forms multigroup cross section tables.  These tables are prepared in standard formats for transport or diffusion theory calculations. In addition, it is possible to use the combining section to produce mixtures not used in the spectrum calculation or to combine the results of different fast and thermal  section calculations and so on. These options are described in reference 4.
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4. METHOD OF SOLUTION

In the fast section either the P1 or the B1, B2, or B3 approximation is made to the transport equation using a single energy-independent buckling. In each approximation Legendre moments of the angular flux are computed by direct numerical integration of the slowing down equations. In the resonance calculations, Doppler broadened (at an input temperature) absorption and scattering cross sections are used. The resonance treatment allows up to two admixed moderators in an absorber lump imbedded in  a surrounding moderator of finite size. The absorber in the lump is  treated by using either the narrow resonance approximation, or a solution of the slowing down integral equations to determine the collision density through the resonance. The admixed moderators are  treated by using either an asymptotic form of, or an integral sol density. In the resonance calculation either standard geometry collision probabilities are used or tables of collision probabilities are entered. Dancoff corrections can also be made. In  the region of unresolved resonances, resonance absorption is calculated by using Porter-Thomas distributions, but only s-wave neutrons are considered. Slowing down sources into the thermal section may be computed for H, D, Be, C, and O according to the free gas scattering model and user-supplied effective temperatures. In the thermal section either the B0, B1, P0, or P1 approximation to the transport equation is made, and in all options Legendre moments  of the angular flux are computed. A trapezoidal energy integration mesh is used, and the resulting equations are solved iteratively by  using a source-normalized, overrelaxed, Gauss-Seidel technique. Averages over broad groups are performed by simple numerical integration. The results obtained in the fast and thermal sections are stored on special tapes. These tapes may contain results for a number of problems, each problem including fine group cross section  datafor a number of nuclides. If the problem number is specified on these tapes, and a desired list of nuclides is given, the combining  code will punch microscopic cross sections for the requested list of nuclides. The program also treats mixtures. Given the atomic densities of the nuclides in a mixture, the code will punch macroscopic cross sections.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

Maxima of -
    99 fast groups
   101 thermal fine groups
    99 fast broad groups
    50 thermal broad groups
    50 broad groups in the combining section
   250 resonances per nuclide
     2 moderators admixed with a resonance absorber
305 entries in the escape probability table for cylindrical        geometries
505 entries in the escape probability table for slab geometries The energy dependence of the input bucklings is restricted to separate energy-independent fast- and thermal-section values (positive, negative or zero values are allowed in either section).
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6. TYPICAL RUNNING TIME

A B1 calculation in the fast section for 4 nuclides and 5 broad groups takes approximately 1 CPU minute on the  UNIVAC1108 if a resonance calculation (1/2 minute) is performed for  one nuclide. The thermal calculation for 4 broad groups requires approximately 17 CPU seconds, which includes about 1.5 CPU seconds for the iterative procedure. To punch standard diffusion and standard transport cross sections for this problem requires 2 seconds.
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7. UNUSUAL FEATURES OF THE PROGRAM

There is no restriction on the number of problems that can be run consecutively in each section, nor is there a restriction on the number of nuclides per problem.
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8. RELATED AND AUXILIARY PROGRAMS

GGC4 is a revision of the earlier program, GGC3. To prepare, handle, and update the basic cross section tapes which are used as input for GGC4, the following codes  are utilized - MAKE, MST, PRINT, MIXER, WTFG, MGT3, SPRINT, and COMBIN. The "WTFG option" of the GAND2 code (NESC Abstract 596) is used to evaluate data for certain nuclides with resonances in the thermal neutron energy range.
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9. STATUS
Package ID Status date Status
NESC0298/01 01-SEP-1976 Tested at NEADB
NESC0298/02 01-NOV-1974 Tested at NEADB
NESC0298/03 01-FEB-1970 Tested at NEADB
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10. REFERENCES

- J. Adir and K. D. Lathrop:
  Theory of Methods Used in the GGC-4 Multigroup Cross Section
  Code, GA-9021 (October 1, 1968) and Errata (January 28, 1969 and
  March 21, 1975).
- D. Mathews and P. Koch:
  Revised  Input Instructions for the GGC-4 Code
  GGA Memorandum (October 16, 1972).
- D. Mathews and P. Koch:
  Slowing Down Sources for the Thermal Section of GGC
  GGA Memorandum (May 7, 1970).
- P. Koch and D. Mathews:
  Changes to  GGC-4 and GGC-5
  GGA Memorandum (July 8, 1970).
- D.R. Mathews:
  Non-1/E Denominator Tests
  GGA Memorandum (May 11, 1970).
  Simulation of Grain Effects in GGC-4, GGA Memorandum.
- R. J. Archibald  and D. R. Mathews:
  The GAF/GAR/GAND Fast Reactor Cross Section Preparation System
  Volume II, GAND2 and GFE2 - Computer Programs for Preparing Input
  Data for the GAFGAR, GGC and MICROX Codes from an ENDF/B Format
  Nuclear Data File, GA-7542, Vol. II (March 1973).
- Modifications to Implement the UNIVAC 1108 Version of the GGC-4
  Program to the CDC 6600 Computer, WANL Note.
NESC0298/01, included references:
- J. Adir and K. D. Lathrop:
  Theory of Methods Used in the GGC-4 Multigroup Cross Section Code
  GA-9021 (October 1, 1968)
- J. Adir, S.S. Clark, R. Froehlich and  L.J. Todt:
  Users' and Programmers' Manual for the GGC-3 Multigroup Cross
  Section Code, Parts 1 and 2
  GA-7157, (July 25, 1967)
- M. K. Drake, C.V. Smith and L.J. Todt:
  Description of Auxiliary Codes  Used in the Preparation of Data
  for the GGC-3 Code
  GA-7158 (August 7, 1967)
- O. Chiovato:
  Note on the CDC 6600 version of GGC4
  (May 1971) (in Italian)
NESC0298/02, included references:
- J. Adir and K. D. Lathrop:
  Theory of Methods Used in the GGC-4 Multigroup Cross Section Code
  GA-9021 (October 1, 1968)
- J. Adir, S.S. Clark, R. Froehlich and  L.J. Todt:
  Users' and Programmers' Manual for the GGC-3 Multigroup Cross
  Section Code, Parts 1 and 2
  GA-7157, (July 25, 1967)
- M. K. Drake, C.V. Smith and L.J. Todt:
  Description of Auxiliary Codes  Used in the Preparation of Data
  for the GGC-3 Code
  GA-7158 (August 7, 1967)
- O. Chiovato:
  Note on the CDC 6600 version of GGC4
  (May 1971) (in Italian)
- J. Adir and K.D. Lathrop:
Theory of Methods Used in the GGC-3 Multigroup Cross Sections
Code, GA-7156 (July 1967)
NESC0298/03, included references:
- J. Adir and K. D. Lathrop:
  Theory of Methods Used in the GGC-4 Multigroup Cross Section Code
  GA-9021 (October 1, 1968)
- J. Adir, S.S. Clark, R. Froehlich and  L.J. Todt:
  Users' and Programmers' Manual for the GGC-3 Multigroup Cross
  Section Code, Parts 1 and 2
  GA-7157, (July 25, 1967)
- M. K. Drake, C.V. Smith and L.J. Todt:
  Description of Auxiliary Codes  Used in the Preparation of Data
  for the GGC-3 Code
  GA-7158 (August 7, 1967)
- O. Chiovato:
  Note on the CDC 6600 version of GGC4
  (May 1971) (in Italian)
- J. Adir and K.D. Lathrop:
Theory of Methods Used in the GGC-3 Multrigroup Cross Section
Code, GA-7156 (July 1967)
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11. MACHINE REQUIREMENTS

64K memory with 11 tape units (some of which  may be drum areas) on the UNIVAC1108.
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NESC0298/01 FORTRAN-IV
NESC0298/02 FORTRAN-IV
NESC0298/03 FORTRAN-IV
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13. OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED:  EXEC8, Level 27 (UNIVAC1108), SCOPE (CDC6600).
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS

The latest UNIVAC1108 version of the GGC4 code is identified as the EXEC-8, Edition 8 version.
At least one nuclide with a thermal scattering kernel is required in the thermal section spectrum calculation. These nuclides are usually H, D, Be, C or O. Structural materials such as Fe, Cr and Ni as well as the actinides are usually treated as "absorber" nuclides  without detailed scattering kernels although the code is capable of  handling such nuclides with scattering kernels if desired.
There is an option in GGC-4 which makes it possible to shorten the punching process for large two-dimensional transfer arrays. This can be done by specifying a maximum number of desired upscattering and downscattering terms.
The CDC6600 version is organized into four overlays. With the buffer lengths set to 1030 (octal), the largest overlay requires 165,000 (octal) locations. It should be possible to reduce this requirement by a more efficient use of the COMMON area.
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15. NAME AND ESTABLISHMENT OF AUTHORS

   1108          D. R. Mathews
                 General Atomic Company
                 P. O. Box 81608
                 San Diego, California  92138
   6600          D. W. Drawbaugh
                 Astronuclear Laboratory
                 Westinghouse Electric Corporation
                 Box 10864
                 Pittsburgh, Pennsylvania  15236
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16. MATERIAL AVAILABLE
NESC0298/01
File name File description Records
NESC0298_01.001 INFORMATION 10
NESC0298_01.002 SOURCE PROGRAM (F4) 1056
NESC0298/02
File name File description Records
GGC4 INFORMATION 0
GGC4 SOURCE PROGRAM 0
GGC4 INPUT DATA FOR SAMPLE CASE 0
GGC4 PRINTED OUTPUT 0
NESC0298/03
File name File description Records
NESC0298_03.001 INFORMATION 7
NESC0298_03.002 GGC-4 SOURCE & OVERLAY CARDS & DATA BCD IBM 11764
NESC0298_03.003 GGC-4 OUTPUT 1521
NESC0298_03.004 MGT SOURCE & DATA 218
NESC0298_03.005 MST SOURCE & DATA 481
NESC0298_03.006 GATCVT SOURCE & DATA 127
NESC0298_03.007 GAMCVT SOURCE & DATA 341
NESC0298_03.008 PRINT SOURCE & DATA 774
NESC0298_03.009 PRINT OUTPUT 1556
NESC0298_03.010 SPRINT SOURCE & DATA 329
NESC0298_03.011 SPRINT OUTPUT 152
NESC0298_03.012 WTFG SOURCE 1643
NESC0298_03.013 COMBIN SOURCE 475
NESC0298_03.014 DOP SOURCE 924
NESC0298_03.015 MAKE SOURCE 1028
NESC0298_03.016 MIXER SOURCE 496
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17. CATEGORIES
  • B. Spectrum Calculations, Generation of Group Constants and Cell Problems

Keywords: Dancoff correction, Doppler broadening, angular distribution, averages, cross sections, multigroup, neutron spectra, porter-thomas distribution, resonance integrals.