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Argonne’s Nuclear Science and Engineering Division researchers have developed a number of large-scale computer codes for scientific and engineering applications. These codes are validated and maintained for application in a variety of research programs.

Fast Reactor Cross-Section Processing Codes

  • ETOE-2: A code to process ENDF/B data to generate MC2-3 data libraries
  • MC2-3: A code to calculate fast neutron spectra and multigroup cross sections

Diffusion and Transport Theory Codes

  • DIF3D: A code to solve the multigroup steady-state neutron diffusion and transport equations in two- and three-dimensional hexagonal and Cartesian geometries; contains a finite difference diffusion theory option (DIF3D-FD), a transverse integrated nodal diffusion option (DIF3D-Nodal), and a even-parity transport option (DIF3D-Variant)
  • DIF3DK: A nodal diffusion kinetics code for solving the time-dependent diffusion equation; can carry out point kinetics calculations and spatial kinetics based upon the DIF3D-Nodal option
  • GAMSOR: A utility code that allows the coupled neutron and gamma transport calculation to be orchestrated with DIF3D; this utility code is primarily used to better define the spatial distribution of power in non-driver assemblies of the reactor such as reflector, control, and shield assemblies
  • PROTEUS: Simulation toolset for high-fidelity reactor physics analysis based on discrete ordinates, Method of Characteristics (MOC); PROTEUS also has a nodal methodology similar to DIF3D

Fuel Cycle / Depletion Codes

  • REBUS: A fuel cycle analysis code based upon the DIF3D solver. REBUS can solve the general non-equilibrium fuel cycle analysis problem with fuel assembly shuffling; REBUS can also provide a rapid estimate of the equilibrium state of a given fuel shuffling pattern that can consider fuel fabrication and recycling
  • RCT: A utility program that facilitates pin level power and depletion tracking. All usage is based upon an assembly homogenized non-equilibrium REBUS calculation; multi-cycle usage of specific fuel assembly pins or entire assemblies can be handled
  • ORIGEN-RA: A modified version of the ORIGEN code developed by Oak Ridge National Laboratory to perform nuclide transmutation calculations; ORIGEN-RA updates the actinide reaction rates in the ORIGEN library with values computed using either REBUS or RCT such that the ORIGEN calculation is consistent with the reactor system being analyzed; in addition to nuclide inventories, ORIGEN-RA is used to estimate radiation emission characteristics and decay power for irradiated reactor constituents
  • ADDER: Software to provide an interface between generic neutronics and depletion solvers to perform through-life reactor performance analyses, including fuel management, with a user-friendly interface

Perturbation Theory Codes

  • VARI3D: A code based upon DIF3D-FD that is used to generate kinetics parameters and perturbation theory based reactivity coefficients for use in a conventional point kinetics based safety analysis of a reactor; VARI3D can also generate R-Z based sensitivity coefficients for use in an uncertainty assessment
  • PERSENT: A code based upon DIF3D-VARIANT that is used to generate kinetics parameters and perturbation theory based reactivity coefficients for use in a conventional point kinetics based safety analysis of a reactor; PERSENT can also generate 3D transport based sensitivity coefficients and carry out the follow on uncertainty assessment given base uncertainty information on the cross section data

Thermal-Hydraulic Codes

  • SE2-ANL: A heavily modified version of the SUPERENERGY-2 thermal-hydraulic code; Interfaced with GAMSOR to define the assembly geometry and power distribution; includes pin level temperature and hot channel factor analysis capabilities; primarily used to define the steady state flow and temperature distribution in a proposed reactor design which is used by SAS and NUBOW as input
  • DASSH (Ducted Assembly Steady-State Heat transfer software): Calculates core-wide, steady-state coolant subchannel temperature and flow distributions in ducted, hexagonal assemblies

Reactor Dynamics and Safety Analysis Codes

  • SAS4A/SASSYS-1: Systems-level software for the deterministic analysis of design basis and beyond-design basis accidents in liquid metal cooled reactors (LMRs)
  • SAM: A modern system analysis tool for advanced non-LWR safety analysis

Surveillance and Diagnostics Codes

  • MSET: Software system for real-time process monitoring
  • PRO-AID (Parameter-Free Reasoning Operator for Automated Identification and Diagnosis): Software package to perform real-time monitoring and diagnostics for an engineering system using a form of automated reasoning

User Interface to ARC Codes

  • Workbench/PyARC: User interface to the ARC codes that facilitates input generation, workflow management, and output processing

Fuel Cycle Transition Analysis Codes

  • DYMOND: Nuclear fuel cycle systems code that simulates the time-dependent behavior and evolution of the entire nuclear fuel cycle

Economics Codes and Algorithms

  • ACCERT (Algorithm for the Capital Cost Estimation of Reactor Technologies): A general methodology along with a large set of algorithms developed for rough order of magnitude estimation of the overnight capital cost for reactors early in the design development
  • NE-COST: Combinations of algorithms that calculate levelized cost of nuclear fuel cycle options, developed for simple and efficient evaluation of multiple options and sampling cost distributions including correlated parameters between multiple fuel cycle options

Acquired Codes

  • KENO
  • WIMS
  • MCNP
  • RELAP5-Mod 3.2
  • OpenMC