SCALE 6.3: 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; Intel Mac OS X; and Windows 10 (C860MNYCP02).
Executables only for Linux; Intel Mac OS X; and Windows 10 (C860MNYCP03).
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 system consists of easy-to-use analytical sequences, which are automated through control modules, to perform the necessary data processing and manipulation of well-established computer codes, referred to as functional modules. Computations with SCALE are typically characterized by the type of analysis to be performed (e.g., criticality, shielding, or lattice physics) and the geometric complexity of the system being analyzed. The user then prepares a single set of input descriptors in terms of easily visualized engineering parameters specified in a simplified, free-form format. The analytical sequence is defined by this single input specification. The SCALE control modules use this information to derive additional parameters and prepare input for each of the functional modules necessary to compute the quantities of interest that can utilize multiple radiation transport codes that employ discrete ordinates, Monte Carlo, or hybrid methods.
Modeling assumptions that limit or restrict the usefulness or accuracy of the individual module are discussed in SCALE 6.3 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 depletion/decay problem to several hours for a complex 3D shielding or sensitivity/uncertainty analysis or 3D Monte Carlo depletion case. Runtimes for sample problems vary from approximately 12 hours to 24 hours depending on the speed of the machine.
SCALE 6.3 includes binary executable files for Linux, Mac OS X, and Windows.
o Minimum requirements: 8 GB RAM per CPU, 80 GB of disk space + additional space to store output results
o Recommended requirements: 8 GB RAM per CPU, 80 GB of disk space + 100 GB of scratch space + additional space to store output results
o Production requirement for large models: 64 GB RAM, 80 GB of disk space + 500 GB of scratch space + additional space to store output results
Executables included in the SCALE 6.3 distributions:
· Linux executables created using the GNU GCC-4.8.3
· Darwin (MacOS X) executables were created using the GNU GCC-4.8.3
· Windows EXEs were created using the INTEL IFORT, ICC and ICPC 15.0
All executables included are 64-bit architecture.
Included in electronic (PDF) format on the distribution DVD:
SCALE 6.3.1 User Manual, ORNL/TM-SCALE6.3.1 (February 2023).
AMPX-6: A Modular Code System for Processing ENDF/B, ORNL/TM-2016/43.
Linux, Mac, and Windows executables, source, data libraries, sample problems are included in DVDs. The above referenced documents are distributed with the software. (Source/EXE version C860MNYCP02) &
(EXE only version without source C860MNYCP03)
February 2023
KEYWORDS: BURNUP; COMPLEX GEOMETRY; CONTINUOUS ENERGY; CRITICALITY CALCULATIONS; CROSS SECTION PROCESSING; DISCRETE ORDINATES; DOSE; GAMMA RAY SOURCE; ISOTOPE INVENTORY; DETERMINISTIC; MONTE CARLO; MULTIGROUP; NEUTRON; PLOTTING; SENSITIVITY ANALYSIS; SPENT FUEL CHARACTERIZATION; UNCERTAINTY ANALYSIS; WORKSTATION; CADIS; VALIDATION; REACTOR PHYSICS; DEPLETION.