Nuclear Engineering Division
Nuclear Engineering Division

Reactor Physics and Fuel Cycle Analysis



An extensive powerful suite of computer codes developed and validated by the NE Division and its predecessor divisions at Argonne supports the development of fast reactors; many of these codes are also applicable to other reactor types. A brief description of these codes follows. Contact information is available at the standard description pages for each code.

Fast Reactor Cross Section Processing Codes go to top


ETOE/MC2-2/SDX is a code system for generating broad-group, cell-average microscopic cross sections on the basis of ENDF/B basic nuclear data. The condensation process takes into account resonance self-shielding effects and the heterogeneity of the unit cell geometry and core arrangement. The resulting composition-, temperature- and region-dependent microscopic cross sections are suitable for use in diffusion or transport theory calculations for a fast reactor core.
Standard Code Descriptions: [ETOE code] | [MC2-2 code] | [SDX code]

Diffusion and Transport Theory Codes go to top


DIF3D is a versatile code system for carrying out multigroup, three-dimensional, whole core neutronic calculations. It is used to compute the reactor multiplication factor, flux and power distributions, and other functionals of the neutron flux. The system models Cartesian, curvilinear, and hexagonal core geometries either by means of the finite difference approach or by a highly efficient nodal option.
Standard Code Description: [DIF3D]


DIF3D-K is a code for executing three-dimensional reactor kinetics calculations using the DIF3D-Nodal diffusion formulation for Cartesian or hexagonal core geometries. The time-dependent nodal equations for the neutron flux and the delayed neutron precursors are solved using either the theta method or the space-time factorization method.
Standard Code Description: [DIF3D-K]


VIM is a Monte Carlo code for simulating neutral particle transport. It employs continuous-energy representations of cross sections and enables precise modeling of problem geometry and collision physics. Because long computing times are typically needed to obtain acceptably low statistical uncertainties in predicted quantities, VIM is used primarily for benchmarking less rigorous multigroup diffusion or transport models. To reduce its running time, a parallel computing option has been implemented for both networks of workstations and scalable multiprocessor computers.
Standard Code Description: [VIM]

Fuel Cycle / Depletion Codes go to top


REBUS-3 is a reactor burnup and fuel cycle analysis code that makes use of DIF3D as a flux calculation module. This code models nuclide transmutations on a three-dimensional, region-dependent basis and provides considerable flexibility for specifying operational constraints and fuel management strategies for both in-core and core portions of the fuel cycle. A unique and powerful technique for simulating equilibrium core characteristics can be invoked as an alternative to discrete, cycle-by-cycle core follow calculations.
Standard Code Description: [REBUS-3]


RCT is code that post-processes the depletion-dependent results of REBUS-3/DIF3D nodal calculations to reconstruct intra-assembly distributions of multigroup fluxes, power densities, burnup, and nuclide number densities.
Standard Code Description: [RCT]


ORIGEN-RA is a modified version of the ORIGEN code developed by Oak Ridge National Laboratory. This code is used to perform isotopically detailed nuclide transmutation calculations based on the flux history computed with REBUS-3 and RCT. In addition to nuclide inventories, this code is used to estimate radiation emission characteristics and decay power for irradiated reactor constituents.
Standard Code Description: [ORIGEN-RA]

Perturbation Theory Codes go to top


VARI3D is a generalized perturbation theory code that allows calculation of the effects on reactivity and reaction rate ratios of alterations in microscopic cross sections and/or material number densities. VARI3D is most frequently used to compute the reactivity coefficient distributions and kinetics parameters employed in reactor dynamics and safety analyses. The flux and adjoint distributions required to compute these quantities are provided by DIF3D.
Standard Code Description: [VARI3D]

Thermal-Hydraulic Codes go to top


SE2-ANL is a modified version of the SUPERENERGY-2 thermal-hydraulic code, which is a multi-assembly, steady-state sub-channel analysis code developed at MIT for application to fast reactor (wire-wrapped and ducted) rod bundles. At Argonne, the code was coupled to heating calculation methods based on the DIF3D code system, and models were added for hot spot analysis, fuel element temperature calculations, and allocation of coolant flow subject to thermal performance criteria.
Standard Code Description: [SE2-ANL]

Last Modified: Mon, October 14, 2013 7:36 PM



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