MINX; multigroup crosssections from ENDF/B-IV data. [CDC7600; IBM370; FORTRAN IV (H compiler) and BAL (IBM370), FORTRAN IV (RUN or FTN compiler) and COMPASS (CDC7600)]
MINX calculates fine-group averaged infinitely-dilute cross sections and self-shielding factors from ENDF/B-IV (reference 2) data. Its primary purpose is to generate a pseudo-composition independent multigroup library for input to the SPHINX (reference 3) space-energy collapse program using CCCC-III (reference 4) standard interfaces. MINX incorporates and improves upon the resonance capabilities of existing codes such as ETOX (reference 5) and ENDRUN (reference 6) and the high-order group-to-group transfer matrices of SUPERTOG (reference 7) and ETOG (reference 8). Fine-group energy boundaries, Legendre expansion order, gross spectral shape component in the Bondarenko (reference 9) flux model, temperatures, and dilutions can all be user-specified.CDC7600;IBM370; FORTRAN IV (H compiler) and BAL (IBM370), FORTRAN IV (RUN or FTN compiler) and COMPASS (CDC7600); OS/370 (IBM370), SCOPE (CDC7600); Storage requirements depend on characteristics of the problem. The 50-group library problem with overlay on the IBM370 requires approximately 330K bytes of memory. NESC tested this problem without overlay using 610K bytes on an IBM370/195. NESC execution of the CDC7600 sample problem required about 127,000 (octal) words of storage.
- Research Organization:
- Los Alamos National Lab., NM (USA)
- OSTI ID:
- 6423151
- Report Number(s):
- ANL/NESC-851; ON: DE83048851
- Country of Publication:
- United States
- Language:
- English
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Related Subjects
73 NUCLEAR PHYSICS AND RADIATION PHYSICS
BARYON REACTIONS
COMPUTER CODES
CROSS SECTIONS
HADRON REACTIONS
LIBRARIES
M CODES
MULTIGROUP THEORY
NEUTRON REACTIONS
NEUTRON TRANSPORT THEORY
NUCLEAR REACTIONS
NUCLEON REACTIONS
SELF-SHIELDING
SHIELDING
TRANSPORT THEORY