## GRUCON - evaluated data processing code package

### Purpose

The major objective of GRUCON code package is ENDF data processing for preparing data for various tasks and applications.

Corresponding Author:

Valentin Sinitsa (sinitsa_vv@nrcki.ru)

### Content

GRUCON code package (IPPE-NRCKI, 1980-2021) is a system of modules for evaluated nuclear data processing for production of detailed and multi-group working libraries for transport calculations in reactor physics and radiation shielding applications. The package has an original architecture and command language (see INDC-CCP-344). This page presents GRUCON-D: version 2021.12 (certificate of state registration No. 2014663246).

### Functions

The package GRUCON-D contains functional modules allowing to- read evaluated data files in the ENDF format;
- reconstruct point-wise cross sections from resonance parameters for given temperatures;
- reconstruct point-wise probability tables from average resonance parameters;
- compute cross sections moments in the unresolved resonance range;
- prepare subgroup parameters from cross section moments by Pade technique;
- prepare probability tables in group intervals from point-wise cross sections;
- compute correlation matrices between subgroup cross sections of different reactions, temperatures, materials, successive collisions in slow-down process (generalized subgroups);
- reconstruct Legendre coefficients from resonance parameters with smoothing for given temperatures;
- calculate energy-angular distributions of neutrons scattered on nuclei in free gas and resonance scattering models;
- calculate energy-angular distributions of neutrons scattered on bound nuclei in thermal energy range;
- prepare photon and particle production cross sections;
- prepare radioactive nyclide production cross sections;
- average integral cross sections with build-in or calculated weight function in groups and prepare shielded group cross sections;
- prepare neutron group transition, neutron production in fission, photon or particle production matrices in neutron reactions;
- prepare microscopic group vectors and matrices for mixtures from vectors amd matrices of separated isotopes;
- prepare group matrices for thermal and resonance neutron scattering with upscatter transitions;
- prepare group matrices of photon/particle production in neutron reactions;
- prepare group cross sections and intergroup transition matrices for photo-atomic interactions;
- perform visualized comparison of data files in the PENDF, GENDF, ENDF and GNDS formats, obtained by NJOY, PREPRO and FUDGE processing code systems, with verification purposes;
- output point-wise data in the ACE format for fast/thermal/photo-atomic&atomic-relaxation data for Monte-Carlo transport calculation codes;
- output group-wise data in the BNAB, MATXS, CCCC and TEMBR formats for neutron-photon transport calculation codes;