The ENDF 6 format

Using SANDY requires a certain degree of knowledge of the standard format ENDF-6 used to store evaluated nuclear data into computer readable files. The full ENDF-6 documentation can be found at here.

The ENDF-6 format is a system developed for the storage and retrieval of evaluated nuclear data to be used for applications of nuclear technology. Evaluations are processed from the combination of experimentally measured physical parameters and the predictions of nuclear model calculations in the attempt to extract the true values of such parameters. Since they were constructed for nuclear data processing programs, the ENDF-6 files store collected evaluated data in computer readable format following constraining formatting rules, which render the data cumbersome to be processed without dedicated tools.

The ENDF-6 format provides representations for neutron-induced cross sections and distributions, neutron, photon and charged-particle production data from neutron reactions, photo-atomic interaction data, thermal neutron scattering data, radionuclide production, decay data and fission products yields. The rules to decrypt and process the ENDF-6 format are encoded in SANDY, which can read most of the sections of any ENDF-6 file for neutron-induced data.

An overview of the ENDF-6 structure is reported below.

The material number (MAT)

A ENDF-6 file is divided into material sections, each defined by a unique material number MAT (up to 4 digits). A single file can contain nuclear data evaluations for one or many MAT sections, which refer to different materials. Generally, a material correspond to a single nuclide, a natural element containing several isotopes, or a mixture of several elements such as compounds, alloys or molecules.

Most of the recent ENDF-6 files for neutron-induced data contain only one MAT number.

The data type number (MF)

For each material, the evaluated nuclear data are provided in sections, specified by the section number MF (up to 2 digits). Amongst the several MF numbers, SANDY can read and write the following sections:

MF	description
1	general information and fission multiplicities
3	cross sections
4	angular distributions of secondary particles
5	energy distributions of secondary particles
8	fission yields and radioactive decay data

The ENDF-6 format also allows dedicated MF sections for the storage of evaluated nuclear data uncertainties and covariance matrices. These evaluated data reflect the information coming from the measurements — e.g., systematic errors, machine resolution, ... — and are fine-tuned from the inference of practical reactor applications. The covariance MF sections processable (only parsing is implemented) by SANDY are:

MF	description
31	covariance data for the average fission neutron multiplicities
33	covariance data for cross sections
34	covariance data for angular distributions of secondary particles
35	covariance data for energy distributions of secondary particles

Covariance data for fission yield and radioactive decay data and stored in section MF=8.

The reaction number (MT)

Each MF section is divided in further subsections identified by the reaction number MT (up to 3 digits) that uniquely defines a reaction type. A list of the most common MT numbers for incident neutron reactions follows:

MT	description
1	total cross section
2	elastic scattering cross section
4	inelastic scattering cross section
18	fission cross section
102	radiative capture cross section
452	total average fission neutron multiplicity
455	delayed average fission neutron multiplicity
456	prompt average fission neutron multiplicity
454	independent fission product yields
459	cumulative fission product yields
457	radioactive decay data