Upcoming Training Courses
Training is provided by developers and expert users from the ORNL team. Courses provide a review of theory, description of capabilities and limitations of the software, and hands-on experience running problems of varying levels of complexity.
All attendees for the SCALE training courses must be licensed users of SCALE 6.3 which is available from ORNL/RSICC and the RIST/NUCIS in Japan.
Please request the software at least two months prior to the start of the training course.
Course Dates and Descriptions
Fall 2026 Training
The fall 2026 training block will be held at Oak Ridge National Laboratory October 12-30, 2026. There will be no virtual or hybrid option for the courses. Tentative schedules are available below.
Sensitivity and uncertainty analysis methods provide advanced techniques for code and data validation including the identification of appropriate experiments, detailed quantification of bias and bias uncertainty, identification of gaps in available experiments, and the design of new experiments. The Sampler sequence within SCALE provides a flexible tool for quantifying uncertainties due to manufacturing tolerances as well as composition and dimensional uncertainties in criticality safety assessments. This 5-day training class provides a foundation on sensitivity and uncertainty analysis and applies these methods to criticality safety validation applications, as well as instruction on the use of Sampler for uncertainty quantification. Topics covered include:
- The TSUNAMI sensitivity and uncertainty analysis techniques for determining the sensitivity of the k-eff eigenvalue to cross section uncertainties using both multigroup and continuous-energy physics.
- SCALE’s comprehensive cross section covariance data library, which is applied to these sensitivity coefficients to estimate the data-induced uncertainty in k-eff.
- The TSUNAMI-IP code, which determines the correlation between benchmark and application systems in terms of their shared sources of data-induced uncertainty.
- The VADER trending analysis tool, which uses similarity coefficients from TSUNAMI-IP (among other parameters) to estimate the computational bias and bias uncertainty for design and licensing applications.
- The TSURFER data adjustment tool, which uses generalized linear least squares to adjust nuclear data parameters to minimize discrepancies between computed predictions and the results of integral experiments; these adjustments can then be used to estimate bias and bias uncertainty in design and licensing applications.
- The SAMPLER code for uncertainty assessment, which randomly samples nuclear data and/or system compositions and dimensions to quantify the uncertainty in system k-eff.
This course will cover the theoretical basis for these analysis techniques and will also conduct exercises for attendees to familiarize themselves with these tools. Previous experience with SCALE CSAS5 or CSAS6 is required.
- Monday to Thursday: 8 am-5 pm and Fri: 8 am-12 pm EDT
- An example of the agenda can be viewed here.
This course provides instruction on the use of the CSAS6 sequence for criticality safety calculations and the MAVRIC shielding sequence with 3-D automated variance reduction for deep-penetration problems. CSAS6 uses either the KENO-VI or Shift 3-D Monte Carlo (MC) code to solve the k-effective eigenvalue problem for criticality safety. MAVRIC uses either the Monaco or Shift 3-D MC code for fixed-source shielding analysis. Both sequences use the SCALE Standard Composition Library and the SCALE Generalized Geometry Package (SGGP), which allows for versatile modeling of complex geometries and provides convenient, efficient methods for modeling repeated and nested geometry configurations such as lattices.
The MAVRIC sequence is based on the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. For a given tally to be optimized in a MC calculation, the CADIS method uses the result of an adjoint calculation from the Denovo 3-D deterministic code to create both an importance map for weight windows and a biased source distribution. MAVRIC is completely automated in that from a single user input, it creates the cross sections (forward and adjoint), computes the adjoint fluxes, creates the importance map and biased source, and then executes MC calculation. An extension to the CADIS method using both forward and adjoint discrete ordinates calculations (FW-CADIS) is included in MAVRIC so that multiple point tallies or mesh tallies over large areas can be optimized (calculated with approximately the same relative uncertainty).
Both CSAS6 and MAVRIC use ENDF/B-VII.1 or ENDF/B-VIII.0 nuclear data distributed with SCALE to perform continuous energy (CE) or multigroup (MG) calculations. Both sequences can also be used with the Fulcrum consolidated SCALE user interface for interactive model setup, computation, output review, and 3-D visualization. Instruction is also provided on the SCALE material input and resonance self-shielding capabilities and the data visualization capabilities within Fulcrum for visualizing fluxes, reaction rates, and nuclear data as well as mesh tallies. CSAS6 and MAVRIC/Monaco can be applied together to perform an integrated criticality accident alarm system (CAAS) analysis.
No prior knowledge of SCALE is required
- Monday to Thursday: 8 am-5 pm and Fri: 8 am-12 pm EDT
- An example of the agenda can be viewed here.
This is a hands-on training course that covers the use of ORIGEN for depletion and decay analysis, nuclide inventories, decay heat, and radiation emissions and spectra calculations.
Instruction is provided to enable all participants to:
- Understand what ORIGEN is: ORIGEN’s capabilities for depletion and decay analysis, type and role of nuclear data and nuclear data libraries, capabilities of other modules in the ORIGEN family (ORIGAMI, OBIWAN, OPUS, ARP).
- Gain confidence in creating, validating, and running input with SCALE’s graphical user interface, Fulcrum.
- Learn best practices for determining nuclide inventories, decay heat, and radiation emissions and spectra in spent nuclear fuel and activated materials.
- Gain hands-on experience through interactive tutorials and challenging real-world workshop problems: perform typical calculations with ORIGEN and ORIGAMI; determine nuclide inventories, decay heat, radiation emissions; generate ORIGEN reactor libraries using TRITON; generate and apply ORIGEN activation libraries; post-process, view, and analyze results.
The training will be presented using the latest maintenance release in the 6.3 series. No prior knowledge of SCALE is required.
- Monday to Thursday: 8 am-5 pm and Fri: 8 am-12 pm EDT
Spring 2027 Training
The Spring 2027 training block will be held at Oak Ridge National Laboratory March 8-19, 2027.
There will be no virtual or hybrid option for the courses. The schedule is shown below. Registration fee information is available here.
Registration will open by end of December 2026.
This training course, which is appropriate for novice through advanced users, provides instruction on the use of the CSAS criticality safety sequences in SCALE, which use KENO and Shift Monte Carlo codes for criticality safety calculations.
KENO V is a fast code that allows users to build complex geometry models using basic geometrical bodies such as cuboids, spheres, cylinders, hemispheres, and hemicylinders. KENO-VI is a 3D generalized geometry Monte Carlo code that allows for versatile modeling of complex geometries. Shift is available with both KENO V and KENO-VI geometry capabilities. Both versions of KENO and Shift provide convenient, efficient methods for modeling repeated and nested geometry configurations such as lattices. Both versions of KENO and Shift can use ENDF/B-VIII.0 or ENDF/B-VII.1 cross-section data distributed with SCALE to perform either continuous energy (CE) or multigroup (MG) calculations.
This training course uses the Fulcrum user interface for interactive model setup, 2D and 3D visualization, computation, and output review. Instruction is also provided on the SCALE material input and resonance self-shielding capabilities and Fulcrum’s capabilities for visualizing fluxes, reaction rates, and cross-section data.
The training is presented using the latest maintenance release in the 6.3 series. No prior knowledge of SCALE is required.
- Monday to Thursday: 8 am-5 pm and Fri: 8 am-12 pm EDT
- An example of the agenda can be viewed here.
This is a hands-on training course that covers the use of ORIGEN for depletion and decay analysis, nuclide inventories, decay heat, and radiation emissions and spectra calculations.
Instruction is provided to enable all participants to:
- Understand what ORIGEN is: ORIGEN’s capabilities for depletion and decay analysis, type and role of nuclear data and nuclear data libraries, capabilities of other modules in the ORIGEN family (ORIGAMI, OBIWAN, OPUS, ARP).
- Gain confidence in creating, validating, and running input with SCALE’s graphical user interface, Fulcrum.
- Learn best practices for determining nuclide inventories, decay heat, and radiation emissions and spectra in spent nuclear fuel and activated materials. Gain hands-on experience through interactive tutorials and challenging real-world workshop problems: perform typical calculations with ORIGEN and ORIGAMI; determine nuclide inventories, decay heat, radiation emissions; generate ORIGEN reactor libraries using TRITON; generate and apply ORIGEN activation libraries; post-process, view, and analyze results.
The training will be presented using the latest maintenance release in the 6.3 series. No prior knowledge of SCALE is required.
- Monday to Thursday: 8 am-5 pm and Fri: 8 am-12 pm EDT
- An example of the agenda can be viewed here.
Training Courses: Extended Descriptions
This course takes the participants through the fundamentals of the nuclear data pipeline, from the creation of Evaluated Nuclear Data File (ENDF) libraries, through processing with the AMPX code suite, to end use in SCALE. In addition to their use in SCALE, AMPX libraries are used in the CASL VERA high-fidelity multi-physics code suite; provide depletion, activation, and decay data for ORIGEN, which is integrated in a wide range of tools; and generate covariance data used in sensitivity uncertainty (S/U) calculations. This course is relevant for all users interested in understanding the sources, approximations and important differences in nuclear data libraries as well as for the advanced practitioners wishing to learn to process nuclear data libraries on their own. Those interested in generating custom libraries, whether from international sources such as the Joint European Fission Fusion (JEFF), the Japanese Evaluated Nuclear Data Library (JENDL), among others, or generating special purpose libraries with customized group structures and weighting spectra will find this course particularly useful. The course is focused on the practical use of the AMPX nuclear data processing code distributed with SCALE and includes demonstrations and in-class exercises, in addition to theoretical lectures. Participants will learn how nuclear data, along with associated uncertainty information, is generated before it enters the ENDF library. They will then be guided through the building of processed libraries for neutron transport in continuous-energy and multi-group formats. Nuclear data uncertainty information and its propagation to quantities of interest through S/U methods will be discussed and the associated AMPX processing capabilities will be demonstrated. Detailed discussion of nuclear data validation will be presented. The course will conclude with lectures on fission product yield and decay data and associated uncertainties, along with demonstrations of their use in SCALE.
No prior knowledge of SCALE or AMPX is required.
This training course focuses on using the SCALE/Polaris lattice physics code for generating nuclear data for the PARCS nodal core simulator. Application of the Polaris-generated data for typical core analyses with PARCS will be demonstrated for PWR conventional cores. Participants will learn how to model different lattice types with Polaris, how to process Polaris-generated nuclear data using GenPMAXS, and how to apply these data in PARCS to model LWR cores and perform typical core analyses such as reactivity balance, power distributions, control rod worth, reactivity coefficients and related parameters, throughout a cycle of operation.
Previous experience with SCALE is recommended.
This course provides instruction on the use of the CSAS6 sequence for criticality safety calculations and the MAVRIC shielding sequence with 3-D automated variance reduction for deep-penetration problems. CSAS6 uses either the KENO-VI or Shift 3-D Monte Carlo (MC) code to solve the k-effective eigenvalue problem for criticality safety. MAVRIC uses either the Monaco or Shift 3-D MC code for fixed-source shielding analysis. Both sequences use the SCALE Standard Composition Library and the SCALE Generalized Geometry Package (SGGP), which allows for versatile modeling of complex geometries and provides convenient, efficient methods for modeling repeated and nested geometry configurations such as lattices.
The MAVRIC sequence is based on the CADIS (Consistent Adjoint Driven Importance Sampling) methodology. For a given tally to be optimized in a MC calculation, the CADIS method uses the result of an adjoint calculation from the Denovo 3-D deterministic code to create both an importance map for weight windows and a biased source distribution. MAVRIC is completely automated in that from a single user input, it creates the cross sections (forward and adjoint), computes the adjoint fluxes, creates the importance map and biased source, and then executes MC calculation. An extension to the CADIS method using both forward and adjoint discrete ordinates calculations (FW-CADIS) is included in MAVRIC so that multiple point tallies or mesh tallies over large areas can be optimized (calculated with approximately the same relative uncertainty).
Both CSAS6 and MAVRIC use ENDF/B-VII.1 or ENDF/B-VIII.0 nuclear data distributed with SCALE to perform continuous energy (CE) or multigroup (MG) calculations. Both sequences can also be used with the Fulcrum consolidated SCALE user interface for interactive model setup, computation, output review, and 3-D visualization. Instruction is also provided on the SCALE material input and resonance self-shielding capabilities and the data visualization capabilities within Fulcrum for visualizing fluxes, reaction rates, and nuclear data as well as mesh tallies.
CSAS6 and MAVRIC/Monaco can be applied together to perform an integrated criticality accident alarm system (CAAS) analysis.
No prior knowledge of SCALE is required.
Polaris is a new 2-dimensional (2-D) lattice physics capability in the SCALE code system for LWR analysis. Polaris provides an easy-to-use input for defining lattice geometries, material compositions, and reactor state conditions. Other features of Polaris include a new resonance self-shielding implementation based on the novel embedded self-shielding method (ESSM), a new 2-D method of characteristics (MOC) neutron transport solver, and the integration of the ORIGEN depletion and decay solver for depleting material compositions. For the first three days of this five days course, attendees will learn how to model typical PWR and BWR assemblies (VVER currently not supported): develop geometry models, perform depletion simulations, setup branch and history calculations to generate few-group cross sections for full-core nodal diffusion analysis (.t16 file), and perform reflector calculations.
Sampler is a new uncertainty analysis capability in SCALE that propagates uncertainties in nuclear data and input parameters to estimate the resulting uncertainty in calculated responses for most codes and sequences within the SCALE code system. Using stochastic sampling to generate perturbed calculation models, Sampler can automate multiple runs (i.e. samples) of a calculation model and then post-process the outputs to quantify the uncertainty in user-selected quantities of interest. In the final two days of this course, attendees will learn how to use Sampler with Polaris to quantity the uncertainty in lattice physics quantities of interest (reactivity, nodal cross sections, isotopic inventories) from a broad range of input uncertainty sources (nuclear data, geometry, composition, and reactor condition).
No prior knowledge of SCALE is required.
Sensitivity and uncertainty analysis methods provide advanced techniques for code and data validation including the identification of appropriate experiments, detailed quantification of bias and bias uncertainty, identification of gaps in available experiments, and the design of new experiments. The Sampler sequence within SCALE provides a flexible tool for quantifying uncertainties due to manufacturing tolerances as well as composition and dimensional uncertainties in criticality safety assessments. This 5-day training class provides a foundation on sensitivity and uncertainty analysis and applies these methods to criticality safety validation applications, as well as instruction on the use of Sampler for uncertainty quantification. Topics covered include:
- The TSUNAMI sensitivity and uncertainty analysis techniques for determining the sensitivity of the k-eff eigenvalue to cross section uncertainties using both multigroup and continuous-energy physics.
- SCALE’s comprehensive cross section covariance data library, which is applied to these sensitivity coefficients to estimate the data-induced uncertainty in k-eff.
- The TSUNAMI-IP code, which determines the correlation between benchmark and application systems in terms of their shared sources of data-induced uncertainty.
- The VADER trending analysis tool, which uses similarity coefficients from TSUNAMI-IP (among other parameters) to estimate the computational bias and bias uncertainty for design and licensing applications.
- The TSURFER data adjustment tool, which uses generalized linear least squares to adjust nuclear data parameters to minimize discrepancies between computed predictions and the results of integral experiments; these adjustments can then be used to estimate bias and bias uncertainty in design and licensing applications.
- The SAMPLER code for uncertainty assessment, which randomly samples nuclear data and/or system compositions and dimensions to quantify the uncertainty in system k-eff.
This course will cover the theoretical basis for these analysis techniques and will also conduct exercises for attendees to familiarize themselves with these tools. It is recommended that attendees are familiar with one or both of the SCALE CSAS5 and CSAS6 sequences for criticality safety analysis, although these are not necessary prerequisites.
One of the unique features of the SCALE code system is the flexibility of assembling different SCALE codes or sequences to solve complex problems. Transportation and storage of spent fuel require a computational tool set to characterize both the spent fuel source terms and the doses for containers used to transport or store the fuel. Spent fuel is a complex neutron and photon source that can be well characterized using the ORIGEN code in SCALE. Additionally, ORIGEN can be used to characterize the radioactive sources resulting from activation of non-fissile materials and components in a nuclear reactor, such as the pressure vessel. The variety of source terms generated with ORIGEN can be used for shielding analyses with the MAVRIC sequence. MAVRIC can estimate particle fluxes and dose rates outside of containers, to ensure that the safety requirements for transportation, storage and ultimate disposal of spent fuel or activated materials are met.
This one-week course will first cover the use of ORIGEN for isotopic depletion, decay and radiation source-terms calculation, generation of ORIGEN activation libraries, and the use of the ORIGAMI tool for quick calculation of spent fuel sources. The next part of the course will focus on MAVRIC, including: building complex 3D models (materials and compositions); using a connection to the ORIGEN libraries to model simple radioactive sources; importing complex ORIGEN sources; and, calculating neutron fluxes to create ORIGEN activation libraries. Additionally, the advanced variance reduction tools for deep penetration problems, CADIS and FW-CADIS, that are the foundation of MAVRIC will be covered. This class uses the Fulcrum user interface for interactive model setup, visualization, computation, and output review.
Previous experience with the SCALE/KENO-VI geometry is required.
The TRITON reactor physics sequence in SCALE enables time-dependent simulation of nuclide transmutation in models of varying complexity by coupling SCALE’s neutron transport solvers with the depletion and decay code ORIGEN. TRITON automates the entire workflow––nuclear data processing, neutron transport calculations, nuclide depletion and decay––to estimate key quantities such as nuclide compositions, mixture-specific powers, neutron flux, and other important quantities as a function of burnup.
Attendees of this course will learn how to perform 2D depletion analyses using TRITON with the NEWT deterministic transport solver, with focus on applications relevant to light-water reactors (LWRs). The analysis of results will focus on mixture-specific power, burnup, and neutron flux, and the material inventory stored in ORIGEN nuclide concentration files (f71).
Participants will learn how to:
- Create the NEWT neutron transport model (material composition and the 2D geometry)
- Determine the input for power and use the power normalization options for depletion
- Perform branching calculations for few-group cross section generation
- Perform time-dependent perturbations of densities and temperatures
- Generate ORIGEN reactor libraries in support of rapid spent fuel characterization with SCALE’s
ORIGAMI sequence
Application of SCALE’s Sampler sequence to perform uncertainty analysis for model parameters and nuclear data in TRITON simulations will be demonstrated.
No prior knowledge of SCALE is required.
This is a hands-on training course that will provide instruction on best practices for modeling and simulations with SCALE of nuclide inventories and decay heat in LWR spent nuclear fuel. The course will summarize the underlying modules and sequences in SCALE that enable these types of simulations, including ORIGEN, ORIGAMI, TRITON. SCALE’s graphical user interface, Fulcrum, designed to create, edit, validate, and visualize SCALE input, output, and data files, will be used throughout the course.
Participants will learn how to apply different approaches for nuclide inventory and decay heat calculations for representative LWR fuel assemblies. That will include: how to generate ORIGEN reactor libraries with TRITON, how to perform rapid depletion and decay calculations with ORIGAMI for LWR fuel assemblies with assembly average or axially-varying burnup, how to perform ORIGEN decay simulations over a cooling time of interest. Participants will also learn how to use the Sampler uncertainty quantification tool in SCALE to estimate uncertainties in calculated nuclide inventories and decay heat that are due to uncertainties in input modeling data (geometry, materials, operating parameters).
No prior knowledge of SCALE is required.
In this training course, participants learn how to generate nuclide inventory for several 3D advanced reactor configurations, and how to analyze and post-process these data for use as input in other computational tools.
Participants will learn how to:
- Use SCALE’s TRITON reactor physics sequence to generate core inventory data in the form of ORIGEN nuclide concentration files (f71)
- Perform decay simulations for the determined inventory with ORIGEN
- Interrogate the f71 files with the OBIWAN utility to ensure a thorough understanding of the available cases, the data at each position, and the applied normalizations.
The TRITON portion of this training focuses on the TRITON 3D sequence with the KENO-VI and Shift Monte Carlo neutron transport codes. The data post-processing includes the generation of an easy-to-use inventory interface file (II.JSON) and relevant examples for further post-processing and analysis. This training course provides best practices for generating f71 files and teaches multiple approaches to interrogate and modify results.
Previous experience with TRITON or ORIGEN is required.