SCALE 6.3.3: A Comprehensive Modeling and Simulation Suite for Nuclear Safety Analysis and Design; Includes ORIGEN and AMPX.
Oak Ridge National Laboratory, Oak Ridge, Tennessee
Fortran 2003, C11, and C++11 source code and executables for Linux; MacOS; and Windows (C860MNYCP10).
Executables only for Linux; MacOS; and Windows (C860MNYCP11)
Federal regulations may restrict the distribution of SCALE 6.3 source code. If restrictions apply, RSICC will send the executable-only version. Please note that included executables run only on the machines listed below in section 9 of this abstract.
SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis and design that is developed and maintained by Oak Ridge National Laboratory under contract with the U.S. Nuclear Regulatory Commission, U.S. Department of Energy, and the National Nuclear Security Administration. The code suite is used to perform reactor physics, criticality safety, radiation shielding, and spent fuel characterization for nuclear facilities and transportation/storage package designs.
Visit the SCALE website for additional information.
The SCALE code system is a comprehensive, modular suite for nuclear safety analysis
and design, providing integrated sequences that automate data processing and physics
calculations for a wide range of applications. SCALE analyses are typically characterized
by the type of physics problem being addressed—such as criticality safety, radiation
shielding, reactor physics, sensitivity and uncertainty analysis, nuclide transmutation and
decay, or source terms evaluation—as well as the geometric and material complexity of
the system under consideration.
Users define a problem through a single, structured, free-form input that describes
materials, geometry, compositions, operating conditions, and desired responses using
engineering-oriented parameters. Based on this high-level problem specification, SCALE
automatically prepares and executes the necessary physics modules, including crosssection processing, neutron or gamma transport calculations (deterministic or Monte
Carlo), nuclear transmutations and decays, and response-function evaluation as required
by the selected sequence.
Modeling assumptions that limit or restrict the usefulness or accuracy of the individual module are discussed in SCALE 6.3.3 User Manual.
Running times are problem dependent and depend greatly on the sequence used and the cross-section library selected. They range from less than one minute for a simple 1D criticality or transmutation/decay problem to several hours for a complex 3D shielding or sensitivity/uncertainty analysis or 3D Monte Carlo-based fuel burnup simulation. Runtimes for the set of sample problems vary from approximately 12 hours to 24 hours depending on the speed of the computing system used.
SCALE 6.3 includes binary executable files for Linux, Mac OS X, and Windows. Minimum requirements: 8 GB RAM per CPU, 180 GB of disk space during installation, final disk usage after installation is ~110 GB.
Note: Once unpacked and installed, the SCALE data and code will be ~110 GB, which will increase the total disk space usage to ~171 GB.
After verifying the install was successful, you may delete the data pak and code installers to free your disk of the ~62 GB in setup files, bringing the final install size to ~110GB.
Executables included in the SCALE 6.3 distributions:
. Linux: RHEL 7 or newer; Java 1.18 or newer.
. MacOS: Darwin 13.0 or newer; Java 1.18 or newer.
. Windows 10 or newer; Java 1.18 or newer.
All executables included are 64-bit architecture.
Included in electronic (PDF) format on the distribution DVD:
SCALE 6.3.3 User Manual, ORNL/TM-SCALE6.3.3 (September 2025).
AMPX-6: A Modular Code System for Processing ENDF/B, ORNL/TM-2016/43.
Linux, MacOS, and Windows executables, source, data libraries, sample problems and documentation are included in the Source/Exe distribution (Source/EXE C860MNYCP10). Linux, MacOS, and Windows executables, data libraries, sample problems and documentation are included in the Exe Only distribution (EXE-Only C860MNYCP11) &
February 2026
KEYWORDS: BURNUP; COMPLEX GEOMETRY; CONTINUOUS ENERGY; MULTIGROUP; CRITICALITY CALCULATIONS; CROSS SECTION PROCESSING; DOSE RATES; RADIATION SHIELDING; NUCLIDE INVENTORY; DETERMINISTIC; MONTE CARLO; NEUTRON; VISUALIZATION; SENSITIVITY ANALYSIS; SPENT FUEL CHARACTERIZATION; UNCERTAINTY ANALYSIS; AUTOMATED VARIANCE REDUCTION; VALIDATION; REACTOR PHYSICS; NUCLIDES TRANSMUTATION AND DECAY.