The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program is a U.S. Department of Energy-Office of Nuclear Energy (DOE-NE) program developing advanced modeling and simulation tools and capabilities to accelerate the deployment of advanced nuclear energy technologies, including light-water reactors (LWRs), non-light-water reactors (non-LWRs), and advanced fuels. We leverage the nation’s scientific talent to deliver on our nuclear energy objectives across six technical areas: Fuel Performance, Reactor Physics, Structural Materials and Chemistry, Thermal Fluids, Multiphysics, and Application Drivers.

In 2020, the NEAMS program was allocated $35 million in federal appropriations. The NEAMS program currently manages over $15 million and $8 million in life-cycle funds allocated to industry life-cycle awards and the NEUP program, respectively.

The NEAMS program has sites at Argonne National Laboratory, Idaho National Laboratory, Lawrence Berkeley National Laboratory, Lawrence Livermore National Laboratory, Los Alamos National Laboratory, Oak Ridge National Laboratory, and Sandia National Laboratories.

Modeling and Simulation Tools

Fuel Performance


  • Capabilities: Built on top of MOOSE, which solves partial differential equations (PDEs) important to fuel performance via finite element (energy conservation, stress divergence, and species migration), BISON adds specific fuel behavior and material models designed to represent the response of fuel and cladding/structural layers in a variety of reactor types.
  • Applications: UO2, TRISO, metallic, UN/UC, and U3Si2
  • Integrations: BISON can be coupled to other codes that supply material models, boundary conditions (T/H), and power source term (neutronics); examples include TRACE and MPACT.
  • Coming soon: High burnup structure models for extending light-water reactor (LWR) operation, uranium mono nitride/uranium mono carbide (UN/UC), predictive strain-based cladding performance, uncertainty quantification, particle fuel verification and validation (V&V)
  • Learn more:

Lower Length Scale Methods

  • Capabilities: In addition to the Marmot code, a suite of lower length scale methods is used to obtain information for input to Marmot as well as to BISON directly. These codes are typically not exclusively developed or maintained by the NEAMS program and in some cases, they are community or off-the-shelf codes, but knowledge and experience of using them in the context of multiscale nuclear fuel performance is a unique capability maintained by NEAMS. The codes include but are not limited to VASP (density-functional theory) and LAMMPS (empirical potentials) for atomistic scale simulations, Centipede (MOOSE-based) for cluster dynamics simulations, VPSC for advanced mechanical models and LAROMANCE for coupling reduced order models of VPSC to BISON.
  • Applications: Fission gas evolution in UO2, doped UO2, U3Si2and metal fuel, swelling in U3Si2 and metal fuel, thermal conductivity of UO2 as function of burnup, fission gas bubble resolution in UO2 and U3Si2, mechanical response of Zr-alloys, FeCrAl and HT9.
  • Integrations: Tools like Centipede and VPSC are being coupled to BISON using data analytics techniques.


  • Capabilities: Marmot solves the phase-field equations that resolve evolving microstructure features (mesoscale) and is used to develop empirical-like material models used at the engineering scale (e.g., BISON).
  • Applications: Supports all fuel types in BISON
  • Integrations: Marmot can be coupled to the binary collision Monte Carlo radiation damage application, Magpie, and engineering scale codes such as BISON or Grizzly.
  • Coming soon: Coupling to commercial thermodynamic and kinetic databases for high-fidelity materials models
  • Learn more:

nuclear fuel performance, code, modeling and simulation, microstructure, engineering, binary, thermodynamic, kinetic, Marmot nuclear fuel performance, code, modeling and simulation, microstructure, engineering, binary, thermodynamic, kinetic, Marmot

Reactor Physics


  • Capabilities:
    • Developed in FY 2020, Griffin is a time-dependent finite-element based reactor physics code built using the MOOSE framework with weak form formulations for diffusion, PN, and 1stand 2nd-order SN transport and a variety of equivalence techniques
  • Applications:
    • Griffin is specially designed to support multi-physics applications and to natively couple to any MOOSE application. It supports both two-step and on-line cross sections preparation and includes state-of-the-art depletion solvers.
  • Integrations:
    • Griffin integrates two previously code suites, the INL-developed MAMMOTH/Rattlesnake code suite and the ANL-Developed MC2-3/PROTEUS code suite
  • Coming Soon:
    • Improved heterogeneous transport and on-line cross section preparation
  • Learn More:


  • Capabilities: MPACT is a transient one-step code that primarily relies on the 2D/1D method for whole core transport and has been applied to pressurized-water reactors (PWRs) (dozens of operational cycles), and boiling-water reactors (BWRs). With this approach, 3D geometries are separated into 2D-radial and pin-homogenized, 1D-axial problems.  For the 2D solvers, the method of characteristics is used and for the 1D solvers, P3 transport with a fourth-order Legendre-based spatial flux expansion is used.  These 2D and 1D solvers are coupled through axial and radial transverse leakages and accelerated with a 3D-coarse mesh finite difference solver. MPACT uses inline cross section generation based on the subgroup method and uses ORIGEN for depletion.
  • Applications: Nuclear reactor physics steady-state eigenvalue (core follow) and transient simulation (load follow and NRC Chapter 15 licensing events) with cross section feedback and material isotopic depletion
  • Integrations: Full integration with the thermal-hydraulics code CTF, fuel temperature codes CTFFuel and BISON, crud chemistry code MAMBA, and ex-core transport code Shift.
  • Coming Soon:
    • BWR Support
    • Extended transient analysis for series of operational transients
  • Learn More:


  • Capabilities: VERAShift is a hybrid deterministic-Monte Carlo ex-core radiation transport interface package serving as the application programming interface (API) for integrating in-core depletion and thermal-hydraulics calculations with ex-core calculations performed with Shift. Its initial development began in 2015 and it is mainly used to perform ex-core transport calculations for LWRs in the VERA code suite. These calculations include vessel fluence, detector response including during reactor startup, and concrete fluence. VERAShift is also used as a verification and/or validation tool by utilizing its Monte Carlo eigenvalue solution capability.
  • Applications: Vessel and reactor component fluence, detector response (including reactor startup), concrete fluence, and coupon fluence
  • Integrations: MPACT, CTF and Shift
  • Coming Soon: BWR geometry support for ex-core dose analysis. Efficient reference solutions for non-LWR calculations.
  • Learn More:


  • Capabilities: A parallel, Monte Carlo solver for radiation transport application development on high-performance computing (HPC) platforms with multiple physics and geometry options. Shift is integrated with the Denovo deterministic solvers for hybrid radiation transport calculations and can execute on CPU and GPU based architectures.
  • Applications: LWR and Non-LWR reactor analysis and design, radiation shielding, criticality safety, radiation dosimetry, and fusion system analysis. Shift supports two-step workflows to pre-generate cross sections for Griffin, reference reactor physics solutions for given temperature and density distributions, and ex-core detector or dose assessment.
  • Integrations: Shift is part of the SCALE system and is integrated with ORIGEN for depletion and Denovo for hybrid radiation transport calculations.
  • Coming Soon: Optimized pole data for continuous energy (CE) cross section data, on-the-fly doppler broadening, optimized CE MC full core depletion on GPUs. Shift-to-Griffin two-step workflow.
  • Learn more:
nuclear fuel performance, code, modeling and simulation, microstructure, engineering, binary, thermodynamic, kinetic, Shift
An axial slice at the core midplane of a Watts Bar Unit 1 excore model detector response adjoint flux, optimized for the north and south detectors.

Structural Materials and Chemistry


  • Capabilities: MAMBA (MPO Advanced Model for Boron Analysis) simulates three-dimensional crud growth along the surface of fuel rods within light-water reactors. For pressurized-water reactors (PWRs), MAMBA also simulates the boron uptake in the crud layer during cycle operation. MAMBA includes models for crud heat transfer and boiling, chemical kinetics and transport, chemistry thermodynamics (including boron solubility), and reactor primary system source term modeling for crud constituent release and tracking (e.g. Ni and Fe).
  • Applications: MAMBA calculations include crud induced power shift (CIPS), which is the adverse effect on nuclear power distribution due to boron uptake in the crud layer, and crud induced localized corrosion analysis which can have an adverse effect on fuel performance.
  • Integrations: MAMBA is tightly coupled to the neutron transport code, MPACT and the thermal-hydraulics code within the VERA code suite.
  • Coming Soon: Current development activities are focused on use of physics-based calibration methods utilizing operating reactor flux map data for determination of unknown crud model parameters.
  • Learn More:


  • Capabilities: Grizzly simulates progression of aging mechanisms in nuclear power plant components and the integrity of these components subjected to degradation during normal and off-normal loading conditions. BlackBear is an open-source code, with non-nuclear-specific models, that provides a subset of Grizzly’s capabilities.
  • Applications:
    • LWR reactor pressure vessel (RPV) probabilistic fracture mechanics (PFM): 1D, 2D, or 3D modeling of RPV thermomechanical response during pressurized thermal shock (PTS) transient events. Semi-empirical model currently used for embrittlement prediction. PFM model evaluates the probability of large sets of flaws using Monte Carlo analysis.
    • Concrete degradation: Coupled physics (thermal and moisture transport and mechanics) modeling of processes leading to expansive reactions in concrete due to alkali-silica reaction (ASR) and radiation induced volumetric expansion (RIVE)
    • High-temperature creep modeling: Models for high-temperature creep of advanced reactor structural materials using both traditional phenomenological approach and a reduced order model based on material microstructure
  • Integrations:
    • VERAShift: Provides 3D fluence maps used for radiation-induced damage in RPVs
    • NEML: Material constitutive model library that provides phenomenological high-temperature creep models
    • MASTODON: MOOSE-based code for simulation of seismic response of structures
  • Coming soon:
    • Coupling with 3D computational fluid dynamics (Nek5000) to provide thermal boundary conditions for RPV during PTS
    • Predictive models for LWR RPV steel embrittlement based on multiscale modeling are under development
  • Learn more:
light water reactor, lwr, nuclear power plant components, open source code, NEAMS, grizzly
Grizzly calculation of thermomechanical response of a light-water reactor (LWR) reactor pressure vessel (RPV) under transient loading conditions. Equivalent results are obtained for 1D axisymmetric, 2D planar, and 3D geometric representations in the beltline region.
light water reactor, lwr, nuclear power plant components, open source code, NEAMS, grizzly
Grizzly calculation of the locations of flaws with a nonzero probability of fracture (shown as white dots) in a large number of Monte Carlo iterations superimposed on a fluence map of the beltline region of a LWR RPV obtained from the VERAShift code. The clustered points along lines indicate that failures are more likely to occur along welds.


Yellowjacket Corrosion Suite

  • Capabilities: The Yellowjacket Corrosion Suite simulates leaching of species from a material in contact with a fluid in a flow loop and deposition from the fluid onto a solid surface. The current focus is fueled and unfueled molten salts, but the capability can be applied to any heat transfer fluid (HTF). Fundamental to corrosion and mass accountancy, in general, are thermophysical properties. These are generated using ab initio and classical molecular dynamics simulations for molten salts and HTFs. The code suite is composed of four fundamental parts:
      1. Yellowjacket, a mesoscale phase field (PF) code for corrosion
      2. Mole, an engineering scale diffusion and reaction kinetics code
      3. A Gibbs Energy Minimizer (GEM) with an accompanying thermodynamic database (MSTDB-TC). The GEM is part of Yellowjacket.
      4. A database for prediction of thermophysical properties (MSTDB-TP) derived from simulations and available experiments (performed by other programs).
  • Applications: 
    • Thermophysical properties, thermochemical properties, and phase equilibria of molten salt systems
    • Corrosion and fouling predictions in MSR flow loops
  • Integrations: 
    • Coupling within Yellowjacket suite (Yellowjacket GEM and PF models and Mole).
  • Coming soon: 
    • Coupling within Yellowjacket suite (Yellowjacket GEM and PF models and Mole)
    • Coupling with Grizzly for chemistry-induced degradation of structural components
    • Coupling with thermal hydraulic codes, e.g., SAM and Nek5000
  • Learn more:

Thermal Fluids


  • Capabilities: CTF is a modernized version of the thermal-hydraulics subchannel code COBRA-TF. CTF uses a two-fluid, two-phase flow modeling approach for modeling fluid flow and heat transfer and provides both subchannel and 3D solution options. Flow can be modeled as three independent fields: continuous liquid, vapor, and liquid droplets. Two-phase flow and heat transfer models consider both pre- and post-critical heat flux conditions. CTF also provides a set of solid thermal conduction modeling capabilities that can be used for tubes, cylinders, and nuclear fuel rod geometries. A nuclear fuel rod solver, CTFFuel, is also provided for modeling of burnup-dependent fuel temperatures.
  • Applications: CTF is used primarily for modeling in-core light-water reactor fluid flow and heat transfer in normal and off-normal operating conditions. It is used for performing steady-state and transient simulations of pressurized water reactor (PWR) and boiling water reactor (BWR) conditions including depletion conditions, prediction of departure from nuclear boiling (DNB), and reactivity-initiated accident (RIA) analysis.
  • Integrations: CTF is coupled to the VERA core simulator for calculation of thermal-hydraulic feedback needed by neutronics, fuel performance, and crud simulation tools. CTF and CTFFuel have also been integrated into the NEAMS Workbench.
  • Coming soon: Current development activities are focusing on improvements for BWR modeling and simulation, including development and implementation of more accurate closure models, addition of support for BWR-specific geometry, and addition of critical power modeling capabilities.
  • Learn more:


  • Capabilities: Nek5000 is an open-source high-fidelity computational fluid dynamics code based on the spectral element method. It can simulate fluid flows and heat transfer at high spatial discretization order using a variety of turbulence models including direct numerical simulation (DNS), large eddy simulation (LES), and Reynolds-averaged Navier-Stokes (RANS). It has demonstrated impressive computational scalability, routinely running on leadership class high-performance computing (HPC) facilities. Nek5000 has a long history of excellent performance in international verification and validation benchmarks and continues to be used for a wide variety of fluid flow and heat transfer problems.
    • High-fidelity simulation of fluid flow and heat transfer phenomena
    • Multiple turbulence models: DNS, LES, RANS
    • Variable fluid properties using a low-Mach model
    • Analysis of complex flow physics, i.e., flow-induced vibration (FIV)
  • Applications: Determination of friction factors, loss coefficients, and heat transfer coefficients
  • Integrations: Integrations with MOOSE, SAM, OpenMC, PROTEUS, and Diablo
  • Coming soon: Improved robustness for the RANS models, wall functions for high-y+ turbulence models
  • Learn more:


  • Capabilities: Pronghorn is a multidimensional, coarse-mesh, thermal-hydraulics (TH) code for advanced reactors. It serves the intermediate fidelity realm situated between detailed computational fluid dynamics (CFD) analysis and lumped system models.
  • Applications:
    • Pronghorn is particularly well-suited for gas-cooled pebble-bed (PBR) and prismatic reactors, fluoride-salt cooled high-temperature reactors (FHR), and open-pool molten salt reactors (MSRs).
    • The main area of application for Pronghorn is the analysis of accident scenarios of advanced reactors, e.g., loss of flow and overcooling (among others). Pronghorn’s coarse-mesh approach also enables rapid scoping studies and design iterations.
  • Integrations:
    • Pronghorn is built and tested with the newest version of MOOSE and follows the MOOSE development paradigm. Therefore, Pronghorn leverages the seamless multi-application data transfer and execution capabilities of MOOSE and enables thermal-hydraulic feedback in many other MOOSE applications.
    • Pronghorn readily integrates with the Thermal-Hydraulics Module (THM) for coupled core/primary loop/secondary loop TH analysis and prediction of bypass flows.
    • Pronghorn has been coupled to Griffin deterministic neutronics for a number of applications, such as coupled steady-state, 100-40-100 load follow, and LOFC transient for the PBMR-400 benchmark and precursor drift feedback for pool-type MSRs.
    • A multiscale coupling of macro, meso, and micro length scale models based on MOOSE’s multi-application system is used to predict pebble bed reactor TRISO transient thermal performance.
  • Coming soon:
    • Pronghorn currently uses a continuous finite element method (CFEM) discretization. For more robust solution, a finite volume discretization is being added with both compressible and Boussinesq-type incompressible formulations.
    • Turbulence modeling
    • Coupling with NEAMS system analysis tools such as THM and SAM to efficiently model 1D components outside of the multidimensional core region
    • Advanced plug-and-play interface for agnostic porous media calculations for importing custom closures from higher fidelity calculations. This capability will allow computationally efficient analysis of complex phenomena such reflector bypass flows and the near-wall physics in PBRs.
    • Improvement and validation of correlations for pebble bed reactors using detailed, spatiotemporally resolved data.
  • Learn more:
Molten Salt Reactor Modeling, NEAMS, pronghhorn, finite element method, turbulence modeling
Molten Salt Reactor Modeling


  • Capabilities: System Analysis Module (SAM) is a fast-running, whole-plant transient analysis code with improved-fidelity capability for fast turnaround design scoping and safety analyses of advanced non-light-water reactors.
    • One-D flow networks represent general fluid systems such as the reactor coolant loops
    • Multichannel representation of fuel assembly and interassembly heat transfer modeling
    • Point kinetics model with various reactivity feedback including thermal expansions
    • Specific component models for reactor transient simulations
    • A general mass transport capability for species transport
    • Control and trip system models
    • Reduced-order multidimensional flow models for thermal mixing and stratification modeling
  • Applications: System transient analysis of various advanced reactor types for design optimization and licensing support, under operational transients and accident conditions
  • Integrations: Coupling with Mammoth, Proteus, Nek5000, and BISON, STAR-CCM+, SAS4A/SASSYS-1, and TRACE
  • Coming soon:
    • Fluid freezing/thawing modeling capability for modeling of overcooling events in liquid-salt-cooled reactors
    • General mass transport modeling capability in the reduced-order multidimensional flow model, for application to molten salt fueled (fast spectrum) reactors
    • Tritium transport modeling in molten salt cooled reactors
  • Learn more:
nuclear reactor analysis, code, NEAMS, modeling and simulation, safety, fluid systems
Snapshots of Temperature Distribution in the coupled SAM-CFD simulations of ABTR Protected Loss-of-Flow Transient


  • Capabilities: Sockeye is a MOOSE-based heat pipe simulator and analysis tool. Heat pipe simulation offers the ability to accurately predict heat transfer for applications involving heat pipes, including heat-pipe-cooled microreactors. Additionally, it provides insight on heat pipe performance; operational heat pipe limits can be predicted in transient conditions and with greater flexibility than steady-state analyses can provide. Heat pipe simulation can be used by industry to inform heat pipe and microreactor design. Simulations can be performed to determine the range of operating conditions applied to a heat pipe and whether reliable heat pipe operation can be sustained in these conditions. If operational limits are encountered, heat pipe design can be altered to avoid these limits.The following is a summary of Sockeye’s capabilities:
    • Transient, 1D, two-phase flow coupled to 2D heat conduction
    • Transient, 2D, effective thermal conductivity model
    • Analytic tools for analyzing heat pipe operating envelope
  • Applications: Heat pipe transient simulation
  • Integrations: Sockeye couples via heat transfer to other physics applications in the DireWolf microreactor simulator application
  • Coming soon: Robust phase disappearance, startup/shutdown modeling
  • Learn more:

heat pipe simulator and anaysis, modeling and simulation, NEAMS, heat transfer



  • Capabilities:
    • The foundation of the multiphysics simulations is based upon the MOOSE framework. The use of MOOSE as the library for multiphysics applications standardizes code inputs, communication, and linking. The MOOSE framework is well-suited to solve a wide range of fully coupled physics problems. Additionally, MOOSE as a platform gives developers access to a wide range of tools and processes to build production quality applications including the pluggable, parallel regression test system and NQA-1 compliant processes.
    • MOOSE is a finite-element framework that aids in application development by harnessing state-of-the-art fully coupled, fully implicit multiphysics solvers while providing automatic parallelization, mesh adaptivity, and a growing set of physics modules. It is the framework upon which several of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program codes are based, including BISON, SAM and Pronghorn. It has also been used to successfully couple MOOSE-based codes (such as BISON) to external codes such as Nek5000 and the NRC’s FAST and TRACE applications.
  • Application: MOOSE is the underlying framework for most of the NEAMS physics tools as well as the capability for connecting multiple physics applications together. Individual codes employing MOOSE include Griffin, BISON, SAM, Pronghorn, and Grizzly.
  • Integrations: MOOSE serves as the integrating tool for all applications built upon the framework as well as those that have been “wrapped” by MOOSE including Nek5000, MPACT, OpenMC, TRACE, and FAST.
  • Coming soon: MOOSE recently gained an automatic differentiation capability as well as finite volume discretization that can be coupled with existing finite element simulations. Improvements in dynamic application loading are being planned that will enable more flexible use of different combinations of applications for complex multiphysics simulations in support of advanced reactor concepts.
  • Learn more:

NEAMS Workbench

  • Capabilities: The NEAMS Workbench is an initiative intended to facilitate the transition from conventional tools to high-fidelity tools. The vision of the NEAMS Workbench is to provide a cross-platform graphical user interface (GUI) designed to facilitate problem creation, modification, navigation, validation, and visualization, as well as output and data file interaction as needed by new and experienced users in order to lower learning curves, enable understanding, and shorten research and development iterations and subsequent nuclear energy science and technology deployment timelines.
    • Cross platform GUI supporting users on Windows, Mac, and Linux operating systems
    • Integrated input development environment providing users application input auto completion and checks, and quick navigation
    • Remote job launch to Portable Batch Scheduler (PBS)-type scheduled commodity compute cluster and scheduler-less compute boxes
  • Applications:
    • MOOSE applications (BISON, SAM, GRIFFIN, etc.) fuel performance and system analysis, etc.
    • Argonne Reactor Computation (ARC) PyARC fast reactor workflow
    • CTF-SubKit subchannel thermal hydraulics workflow
    • Nek5000 v17 scalable high-order solver for computational fluid dynamics
    • CTFFuel fuel analysis
    • SCALE Code System for nuclear safety and design
    • Monte Carlo N-Particle (MCNP) general purpose Monte Carlo particle transport
    • Dakota sensitivity and uncertainty quantification and problem optimization
    • VisIt data visualization and analysis tool kit
  • Integrations:
    • Integrated data plotting and mesh data visualization capabilities
    • Integrated constructive solid geometry (CSG) viewer providing users visual verification of CSG input
    • Open-source language processing capabilities
  • Coming soon:
    • Support for Virtual Environment for Reactor Analysis (VERA) light-water reactor (LWR) core simulator input editing, job launch, and results analysis
    • Integration of ParaView for data visualization and analysis
    • Support for Nek5000 v19 and 20
  • Learn more:


  • Capabilities:
    • Used to solve neutronics, thermal-hydraulics, fuel performance, and coupled physics problems.
    • Integrates physics components based on science-based models and state-of-the-art numerical methods.
    • Verified and validated using data from operating light-water reactors, single-effect experiments, integral tests, and critical experiments.
    • Optimized for efficient execution on multiple platforms, including leadership-class computers, advanced architecture platforms now under development, and industrial engineering workstation clusters.
  • Applications:
    • Core follow calculations for a range of pressurized-water reactor (PWR) designs
    • Integrated ex-core dose calculations
    • Integrated crud simulation capability
    • Integrated fuel performance
    • Integrated capability to perform reactivity insertion events
  • Integrations:
    • VERA integrates MPACT for neutronics, CTF for subchannel analysis, ORIGEN for depletion, MAMBA for crud chemistry, BISON for fuel performance, and Shift for ex-core modeling and reference solutions.
  • Coming Soon:
    • BWR Support
    • Extended transient analysis for series of operational transients
  • Learn More:
    • Contact: Ben Collins
    • Website:
    • VERA Users Group: Independently operated from the NEAMS Program, the VERA Users Group is a forum for information exchange and user support activities to encourage further use and adoption of the VERA products. It is managed by a Steering Committee selected from among its members.