MINX: a multigroup interpretation of nuclear X-sections from ENDF/B

Abstract

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Fast Reactor Physics and Computational Methods

“…Then, the cross sections for a particular composition at a given temperature are interpolated from the cross sections in the generalized library by calculating the background cross section for each material. Typical codes employing this approach are MINX [35], SPHINX [36] and NJOY [27]. An alternative approach was developed at Argonne National Laboratory (ANL) based on detailed spectrum calculations for individual compositions, and implemented in the ETOE-2/MC 2 -2/SDX code system.…”

Status report on multigroup cross section generation code development for high-fidelity deterministic neutronics simulation system.

“…Instead of the overlapping terms, the background cross section infers the approximate composition dependence of the multigroup cross sections so that this scheme provides an efficient means for survey-type calculations. Typical codes employing this approach are MINX/SPHINX [12,13] and NJOY [14].…”

The NJOY Nuclear Data Processing System: Volume 3, The GROUPR, GAMINR, and MODER modules

“…The basic energy-dependent cross sections and angle-energy distributions needed for Eqs. (lO), (11), and (12)…”