The ENDF-6 format

To better understand how SANDY works it is important to have some knowledge of the standard format ENDF-6 used to store evaluated data into computer readable files. The full ENDF-6 documentation can be found at https://www.oecd-nea.org/dbdata/data/manual-endf/endf102.pdf .

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 in the following sections.

The material number (MAT)

A ENDF-6 file is divided into material sections, each defined by a unique material number MAT. 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.

Note

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. Amongst the several MF numbers, SANDY can process the following:

• MF1:

  : general information and fission multiplicities;

• MF3:

  : neutron cross sections;

• MF4:

  : angular distributions of secondary particles;

• MF5:

  : energy distributions of secondary particles;

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 by SANDY are

• MF31:

  : covariances for the average fission neutron multiplicities;

• MF33:

  : covariances for neutron cross sections;

• MF34:

  : covariances for angular distributions of secondary particles;

• MF35:

  : covariances for energy distributions of secondary particles.

Note

Each MF covariance section contains covariances for the data given in section (MF - 30), e.g. MF33 contains covariances for section MF3.

The reaction number (MT)

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

• MT1:

  : total cross section;

• MT2:

  : elastic scattering cross section;

• MT4:

  : inelastic scattering cross section;

• MT18:

  : fission cross section;

• MT102:

  : radiative capture crross section;

• MT452:

  : total average fission neutron multiplicity;

• MT455:

  : delayed average fission neutron multiplicity;

• MT456:

  : prompt average fission neutron multiplicity.