BIGAMON: a program to process the GAMMON activation file. [In FORTRAN]

The BIGAMON program, designed to process the GAMMON activation file, is described. The program retrieves 100-group neutron reaction cross sections and 25-group gamma-ray decay spectra from the GAMMON file and bebins these data into broader multigroup sets. The photon broad group boundaries must be a subset of the Gammon group structure. 5 tables

GAMMON activation library

The GAMMON activation library is specifically designed for fusion reaction application, but is also adequate for many other design studies. The library contains multigroup cross sections (in 100 energy groups) for 420 neutron-induced reactions, multigroup gamma-ray spectra (in 25 energy groups) for 107 unique daughter products, maximum permissible concentrations (MPC's) for 200 reaction products, and absorbable decay energy (sensible heat) for 85 products. 3 tables

Shielding calculations for the Intense Neutron Source Facility. Final report

Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10/sup 15/ n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

Status of CINDER and ENDF/B-V based libraries for transmutation calculations

The CINDER codes and their data libraries are described, and their range of calculational capabilities are described using documented applications. The importance of ENDF/B data and the features of the ENDF/B-IV and ENDF/B-V fission-product and actinide data files are emphasized. The actinide decay data of ENDF/B-V, augmented by additional data from available sources, are used to produce average decay energy values and neutron source values from sponteneous fission, (..cap alpha..,n) and delayed neutron emission for 144 actinide nuclides that are formed in reactor fuel. The status and characteristics of the CINDER-2 code is described, along with a brief description of more well known code versions; a review of the status of new ENDF/B-V based libraries for all versions is presented