GGC4; multigroup cross sections fast thermal spectra. [UNIVAC1108; CDC6600; FORTRAN V (UNIVAC1108), FORTRAN IV (CDC6600)]

Technical Report ·

OSTI ID:6484840

Mathews, D R

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.

• Research Organization: General Atomic Co., San Diego, CA (USA)
• OSTI ID: 6484840
• Report Number(s): ANL/NESC-298; ON: DE83048298
• Country of Publication: United States
• Language: English

Related Subjects

654003* -- Radiation & Shielding Physics-- Neutron Interactions with Matter
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
ANGULAR DISTRIBUTION
BARYON REACTIONS
COMPUTER CODES
CROSS SECTIONS
DANCOFF CORRECTION
DISTRIBUTION
DOPPLER BROADENING
G CODES
HADRON REACTIONS
INTEGRALS
LINE BROADENING
MULTIGROUP THEORY
NEUTRON REACTIONS
NEUTRON SPECTRA
NEUTRON TRANSPORT THEORY
NUCLEAR REACTIONS
NUCLEON REACTIONS
PORTER-THOMAS DISTRIBUTION
RESONANCE INTEGRALS
SPECTRA
TRANSPORT THEORY