NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM 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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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.

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.

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).

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.

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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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.

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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Package ID | Status date | Status |
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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.

- 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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Package ID | Computer language |
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NESC0298/01 | FORTRAN-IV |

NESC0298/02 | FORTRAN-IV |

NESC0298/03 | FORTRAN-IV |

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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.

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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NESC0298/01

File name | File description | Records |
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NESC0298_01.001 | INFORMATION | 10 |

NESC0298_01.002 | SOURCE PROGRAM (F4) | 1056 |

NESC0298/02

File name | File description | Records |
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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 |
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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 |

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