INTRODUCTION

Discrete-ordinates (SN) transport codes, which solve the Boltzmann equation for the distribution of neutrons and photons in nuclear systems, have reached a high level of development. The early one-dimensional codes DTF-IV (Ref. 1) and ANISN (Ref. 2) are very widely used. The development of diffusion acceleration (Ref. 3) as well as increasing computer speed and capacity has made detailed transport calculations more economical; as a result, codes such as ONEDANT (Ref. 4) and TWODANT (Ref. 5) are seeing increasing use. ` The toroidal geometry capability of TRIDENT-CTR (Ref. 6) and its successor TRISM (Ref. 7)are very useful for fusion reactor analysis. The DIF-3D (Ref. 8) diffusion code and Monte-Carlo codes with multigroup capability like MCNP (Ref. 9) are also used frequently.

However, many of the users of transport codes have the same complaint: it is hard to get good, up-to-date, documented cross-section data and prepare them for input into these codes. The difficulties are multiplied if there is anything unusual about the problem, such as fine groups, self-shielding, transport cross sections, or sophisticated response edits (e.g., heating, damage, or gas production). This report describes a utility code called TRANSX (for transport cross-section code) that works together with a generalized cross-section library called MATXS (for material cross-section library) to give the transport code user easier access to appropriate nuclear data and some capabilities difficult or impossible to get with any other system.

The version of TRANSX described in this report can be used to construct data for fusion reactors, fast fission reactors, thermal fission reactors, and shielding problems. Its main weakness is in computing resonance effects in thermal reactors.

TRANSX was originally developed in the late seventies to handle cross sections for fission, fusion, and shielding applications at Los Alamos. In the early eighties, extensions to handle heterogeneous self-shielding problems for fast reactors were added. In 1984, a version without the fast reactor features was released as TRANSX-CTR (Ref. 10). It was especially well-suited for problems in controlled thermonuclear research (CTR) because of its ability to prepare coupled tables and response edits for heating, damage, gas production, and delayed activity. Various later versions have been made available without formal release or complete documentation; for example, Version 1.11 has been available as S11 at Los Alamos, on the Magnetic Fusion Energy (MFE) computer system at Livermore, and on the San Diego Super Computer Center system since late 1987. Version 1.13, which includes charged-particle capabilities, has been used extensively at Los Alamos. This gradual accumulation of changes and new capabilities, plus the need to react to a new version of the MATXS format developed for NJOY 91, plus the need for a fully documented code more consistent with modern Quality Assurance (QA) requirements, has led to the creation of this new version.

Section II of this report discusses the data needs of the typical SN codes and describes the theoretical background; Section III gives detailed examples of a wide variety of problems that can be solved using TRANSX; Section IV contains descriptions of the input and output files used by TRANSX, including the MATXS cross-section library; Section V summarizes several available MATXS libraries; and Section VI gives a detailed discussion of the code aimed at programmers who have to convert, modify, or maintain the code. Several appendices are attached that give the format specifications for the input and output interface files and partial listings of the outputs from the sample problems.

TRANSX HyperText Manual
TRANSX HyperText Help Package
T-2 Nuclear Information Service