Commonly requested NEA Data Bank Codes
Ongoing

The preceeding year's most commonly requested code packages from CPS are listed below by category. This list is not intended to be exhaustive, but rather to provide a shortcut for users wishing to request these popular packages. To submit a request, click the link nested in the name of the desired package.  

If you desire a package that is not listed, we invite you to search in the full catalogue to place your request.

1. Criticality and Safety Analysis

CRISTAL V2.0.2 - criticality calculations for nuclear fuel cycle facilities (fabrication, reprocessing, etc.), storage and transportation of fissile materials.

SCALE 6.2.3 - tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis.

MORET 5.D.1 - simulation tool that solves the Bolzmann's equation using the Monte Carlo method.

2. Monte Carlo Methods

ADVANTG 3.2.1 - automated tool for generating variance reduction parameters for fixed-source continuous-energy Monte Carlo simulations with MCNP.

INES-CLASS - assessment of accidents of significance that occurred at nuclear facilities.

SERPENT 1.1.7 - tool for continuous-energy Monte Carlo reactor physics burnup calculation, lattice physics applications.

SUPERMC 3.3.0 - transport calculation tool of neutron, gamma radiation process that can also be applied for criticality and shielding design of reactors as well as analysis in medical physics.

TRIPOLI-4 VERSION 9S - coupled neutron, photon, electron, positron 3-D, time dependent Monte-Carlo transport calculation software.

VISUAL EDITOR 61 - program that can run and plot MCNP geometries and calculations.

3. Multiphysics and Particle Transport

FISPACT-II 4.0 - inventory simulation platform for nuclear observables and materials science.

PENELOPE 2014 - a Monte Carlo code system for coupled electron photon transport at high energies.

PHITS-2.88 - transport suite covering almost all particles over wide energy ranges, using several nuclear reaction models and relevant nuclear data libraries.

4. Others

FINIX 1.19.1 - a fuel behavior module that calculates the thermal and mechanical behavior of a nuclear fuel rod during steady-state and transient conditions.

SOURCES-4C - code system that determines neutron production rates and spectra from (alpha,n) reactions, spontaneous fission, and delayed neutron emission due to radionuclide decay. 

VARSKIN 4 V4.0.0 - skin contamination dose calculation utility.