Griffin is a general reactor physics code for advanced reactor applications being jointly developed by Argonne National Laboratory (ANL) and Idaho National Laboratory (INL).  It is based on the multiphysics object oriented simulation environment (MOOSE), which leverages existing applications for solving complex reactor multi-physics problems. Griffin is built on the radiation transport capabilities implemented in the INL Rattlesnake transport solver and MAMMOTH reactor physics tool, as well as capabilities from the PROTEUS, cross section API and MC2-3 tools developed at ANL. Griffin solves the steady-state (source & eigenvalue) and transient radiation transport equations using the finite element method in space and SN/PN/diffusion in angle, and implements micro- and macro-depletion and fuel cycle capabilities, homogenization equivalence (SPH & discontinuity factors) as needed, and cross section generation algorithms.

This workshop will include a tutorial of Griffin analysis and will discuss the strong (implicit) and tight (Picard iteration) coupling of MOOSE applications. A variety of examples will be covered including benchmark problems for: eigenvalue, fixed source, depletion, coupling with other physics tools, transient simulation, etc.

It is recommended that participants bring a Linux laptop with Griffin installed to be able to participate in hands-on exercises. Under the US Department of Energy and Department of Commerce rules, acquiring the code requires a license agreement. Please contact the Software Agreements Administrator at to initiate the licensing process.  It is recommended that this process be started at least two months before the workshop to ensure that the licensing process is completed in time.  More information on the licensing process is available at



RAPID (Real-time Analysis for Particle-transport In-situ Detection) is developed based on the MRT (Multi-stage Response-function particle Transport) methodology that enables its real-time simulation capability. The current version of RAPID is capable of simulating nuclear systems such as spent fuel pools, spent fuel casks, and reactor cores. RAPID solves for pin-wise, axially-dependent fission density, critical/subcritical multiplication, and detector response. Recently, new algorithm for 3-D fuel burnup (bRAPID) calculation and reactor kinetics (tRAPID) have been developed and benchmarked for test problems. Experimental benchmarking for these latter algorithms are underway using the Jozef Stefan Institute’s TRIGA research reactor.

 Further, a multi-user virtual reality system (VRS) has been developed that provides a web application for input preparation, real-time simulation, and output processing and visualization in a virtual environment. For an introduction, please view the following demo

 Topics to be covered:

  • RAPID’s MRT methodology and formulation
  • RAPID code system and benchmarking
  • VRS-RAPID demonstration
  • Hands-on use by participants 


There will be access to wireless internet so that the participants can have remote access to VRS- RAPID. The current version of VRS-RAPID is optimized for a Personal Computer using the Google Chrome browser, but it can be accessed through iPad, Tablet, etc. using any other browser. 

To facilitate establishing individual accounts, participants are encouraged to contact Prof. Haghighat prior to the workshop.



SCALE is a comprehensive modeling and simulation suite for nuclear safety analysis and design 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 to perform reactor physics, criticality safety, radiation shielding, and spent fuel characterization for nuclear facilities and transportation/storage package designs. This workshop will present recent light water reactor (LWR) and non-LWR applications using SCALE. In LWR space, this includes High Assay Low-enriched Uranium (HALEU) up to 8% enrichment, High Burnup (HBU) up to 80 GWd/MTU, and Accident Tolerant Fuel (ATF) concepts. A hands-on tutorials will be given on estimating isotopic bias and bias uncertainty for 80 GWd/MTU spent fuel. In non-LWR space, this includes neutronics and inventory calculations for the following reactor types: Sodium Fast Reactor (SFR), Fluoride High Temperature Reactor (FHR), High Temperature Gas Reactor (HTGR), and Molten Salt Reactor (MSR). An open-source SCALE model for each reactor type will be provided to participants and a hands-on tutorial will demonstrate a recommended analysis workflow using the SCALE GUI, Fulcrum. The tutorials will be taught with the latest 6.3 beta (or 6.3.0 release if available at time of conference), although the majority of the analyses may be performed with the latest release in the previous 6.2 series.