Search NASA⌕ Search

SEARCH · Search NASA

Results for “Nuclear Energy & Advanced Modeling & Simulation”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 37 records · Page 2
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

MOOSE: A Modular Platform for Fission and Fusion Multiphysics

The Multiphysics Object-Oriented Simulation Environment (MOOSE) Framework, as well as MOOSE-based simulation tools, have accelerated the development of fission energy and advanced reactor technologies through the United States Department of Energy, Office of Nuclear Science, Nuclear Energy Advanced Modeling & Simulation (NEAMS) Program. MOOSE contains a complete platform of multiphysics simulation capabilities, capable of running on massively parallel systems, and is developed in an open-source manner with great attention paid to high-quality software quality assurance practices. This overall approach could greatly benefit the fusion energy community, which requires rapid design iteration and improvement in order to facilitate the successful development of fusion as an alternative energy source to fossil fuels. In the first half of this talk, applications of MOOSE and MOOSE-based tools for advanced reactor designs will be showcased, as well as MOOSE ecosystem infrastructure (such as the NEAMS Virtual Test Bed) that enables and accelerates fission reactor design. In the second half, a discussion of how the MOOSE approach to modeling and simulation is currently being applied internationally in fusion energy research and development at the United Kingdom Atomic Energy Authority will be discussed, and ongoing/future domestic research efforts will be highlighted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Nek5000 developments in support of industry and the NRC

This year, the Nuclear Energy Advanced Modeling Simulation program (NEAMS) thermal-hydraulics verification and validation (V&V) work has focused in three areas of Nek5000 V&V-driven development. First, in a close collaborative effort with the U. S. Nuclear Regulatory Commission (NRC) staff, we have continued V&V efforts for the HYMERES-2 project using the OECD/NEA sponsored testing in the PSI PANDA facility. This year’s focus of ANL-NRC collaboration involves Nek5000 setups and validation for a range of problems relevant to and including the HYMERES-2 benchmark from PSI. The primary outcome of this year efforts is a more efficient geometry and inlet modeling simplification after a careful sensitivity study of the inlet profiles and pipe geometries. The resulting modeling choice of a short recycling/fully-developed turbulent inlet is within the experimental uncertainty estimate. This finding simplifies the next step of the cross-V&V HYMERES-2 project. In addition, the ANL team continue to provide assistance to the NRC staff in the form of Nek5000 application support in general and on the use of the HPC platforms of ALCF and INL in particular. This supports the NRC’s assessment of Nek5000 for use with the NRC Blue CRAB code suite. Second, we have implemented and tested more robust model of URANS, namely the k – τ model, a variant of the k-ω model, along with other improvements to RANS Nek5000 modeling in general. Because of its demonstrated robustness and stability, the k – τ model is the only RANS model that has been implemented in the new GPU version of the Nek5000 code, nekRS. Lastly, we report the initial implementation of Jacobian-free Newton Krylov approach to the direct Newton method for steady fluid solvers aimed at acceleration of RANS modeling and at IC improvement for LES campaigns. Also leveraging the Exascale Computing Project (ECP) ANL/CEED & SMR team’s software development effort to support NEAMS problems at large scale of the advanced computing architectures, NekRS, a GPU variant of Nek5000, built on top of kernels from libParanumal using OCCA for portability, has been successfully run on the full system of Summit (4608 nodes, 27648 GPUs).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Nek5000: improvements in the available RANS models, meshing, tutorials, and training

This year, the Nuclear Energy Advanced Modeling Simulation program (NEAMS) thermal-hydraulics report for Nek5000 NRC- and verification and validation (V&V)-driven development focuses on following areas of code application and improvement. First we have continued improvements of RANS modeling capabilities in Nek5000 including improved k-tau model focusing mostly on wallfunction initial implementation with spectral element method (SEM) and initiating investigation of an alternative approach XSEM that greatly reduces discretization errors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Warthog: Progress on Coupling BISON and PROTEUS

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Office of Nuclear Energy at the Department of Energy (DOE) provides a robust toolkit for modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines, with two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line (FPL) and the Reactor Product Line (RPL) provide advanced computational technologies that serve each respective field effectively. There is currently a lack of integration between the product lines, impeding future improvements of simulation solution fidelity. In order to mix and match tools across the product lines, a new application called Warthog was produced. Warthog is built on the Multi-physics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory (INL). This report details the continuing efforts to provide the Integration Product Line (IPL) with interoperability using the Warthog code. Currently, this application strives to couple the BISON fuel performance application from the FPL using the PROTEUS Core Neutronics application from the RPL. Warthog leverages as much prior work from the NEAMS program as possible, enabling interoperability between the independently developed MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. Previous work performed on Warthog allowed it to couple a pin cell between the two codes. However, as the temperature changed due to the BISON calculation, the cross sections were not recalculated, leading to errors as the temperature got further away from the initial conditions. XSProc from the SCALE code suite was used to calculate the cross sections as needed. The remainder of this report discusses the changes to Warthog to allow for the implementation of XSProc as an external code. It also discusses the changes made to Warthog to allow it to fit more cleanly into the MultiApp syntax of the MOOSE framework. The capabilities, design, and limitations of Warthog will be described, in addition to some of the test cases that were used to demonstrate the code. Future plans for Warthog will be discussed, including continuation of the modifications to the input and coupling to other SHARP codes such as Nek5000.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Assembling Multiphysics Nuclear Reactor Simulations Using the MOOSE Framework

The Multiphysics Object Oriented Simulation Environment (MOOSE) [1] is an open-source, parallel finite element framework which provides the foundation for many advanced modeling and simulation tools developed under the Department of Energy (DOE) Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program [2] for the analysis of advanced reactors. The MOOSE framework provides the common foundational capability on which many NEAMS codes for reactor analysis are built. The MOOSE framework also includes several systems to assemble unique workflows and couplingamong MOOSE-based applications. In particular, the MultiApp and Transfer Systems are widely used to assemble different MOOSE-based or MOOSE-wrapped physics applications together to perform loosely or tightly coupled multiphysics simulations. The National Reactor Innovation Center (NRIC) Virtual Test Bed (VTB) [3] hosts publicly available nuclear reactor multiphysics simulation examples which leverage MOOSE’s MultiApp System to meet the modeling needs of different reactor types. The flexibility and robustness of coupling provided by MOOSE permits rapid development of coupled physics models for a wide range of reactor types and events

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Molten salt reactor core simulation with PROTEUS

The Nuclear Energy Advance Modeling and Simulation Program (NEAMS) of the US Department of Energy (DOE) aims to develop advanced simulation tools to aid the development of advanced nuclear technologies. The high-fidelity PROTEUS code was developed in the NEAMS program for reactor physics analysis. This paper tests the applicability of the PROTEUS code to simulate a Molten Salt Reactor (MSR). The OpenMC code based on the Monte Carlo Method was employed to generate the cross sections and to provide stochastic solutions for verification. The PROTEUS code was successfully employed to simulate the steady-state performance of an MSR represented by the ThorCon core design. The 2D lattice, 2D core, and 3D full core calculations performed by PROTEUS show good agreement with that predicted by OpenMC. Furthermore, for the nodal option within PROTEUS, the precursor drift model was also tested, and this effect resulted in a reduction of the k-eff by 300 pcm.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

NEAMS Update: Quarterly Report for October - December 2011

The Advanced Modeling and Simulation Office within the DOE Office of Nuclear Energy (NE) has been charged with revolutionizing the design tools used to build nuclear power plants during the next 10 years. To accomplish this, the DOE has brought together the national laboratories, U.S. universities, and the nuclear energy industry to establish the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program. The mission of NEAMS is to modernize computer modeling of nuclear energy systems and improve the fidelity and validity of modeling results using contemporary software environments and high-performance computers. NEAMS will create a set of engineering-level codes aimed at designing and analyzing the performance and safety of nuclear power plants and reactor fuels. The truly predictive nature of these codes will be achieved by modeling the governing phenomena at the spatial and temporal scales that dominate the behavior. These codes will be executed within a simulation environment that orchestrates code integration with respect to spatial meshing, computational resources, and execution to give the user a common 'look and feel' for setting up problems and displaying results. NEAMS is building upon a suite of existing simulation tools, including those developed by the federal Scientific Discovery through Advanced Computing and Advanced Simulation and Computing programs. NEAMS also draws upon existing simulation tools for materials and nuclear systems, although many of these are limited in terms of scale, applicability, and portability (their ability to be integrated into contemporary software and hardware architectures). NEAMS investments have directly and indirectly supported additional NE research and development programs, including those devoted to waste repositories, safeguarded separations systems, and long-term storage of used nuclear fuel. NEAMS is organized into two broad efforts, each comprising four elements. The quarterly highlights October-December 2011 are: (1) Version 1.0 of AMP, the fuel assembly performance code, was tested on the JAGUAR supercomputer and released on November 1, 2011, a detailed discussion of this new simulation tool is given; (2) A coolant sub-channel model and a preliminary UO{sub 2} smeared-cracking model were implemented in BISON, the single-pin fuel code, more information on how these models were developed and benchmarked is given; (3) The Object Kinetic Monte Carlo model was implemented to account for nucleation events in meso-scale simulations and a discussion of the significance of this advance is given; (4) The SHARP neutronics module, PROTEUS, was expanded to be applicable to all types of reactors, and a discussion of the importance of PROTEUS is given; (5) A plan has been finalized for integrating the high-fidelity, three-dimensional reactor code SHARP with both the systems-level code RELAP7 and the fuel assembly code AMP. This is a new initiative; (6) Work began to evaluate the applicability of AMP to the problem of dry storage of used fuel and to define a relevant problem to test the applicability; (7) A code to obtain phonon spectra from the force-constant matrix for a crystalline lattice has been completed. This important bridge between subcontinuum and continuum phenomena is discussed; (8) Benchmarking was begun on the meso-scale, finite-element fuels code MARMOT to validate its new variable splitting algorithm; (9) A very computationally demanding simulation of diffusion-driven nucleation of new microstructural features has been completed. An explanation of the difficulty of this simulation is given; (10) Experiments were conducted with deformed steel to validate a crystal plasticity finite-element code for bodycentered cubic iron; (11) The Capability Transfer Roadmap was completed and published as an internal laboratory technical report; (12) The AMP fuel assembly code input generator was integrated into the NEAMS Integrated Computational Environment (NiCE). More details on the planned NEAMS computing environment is given; and (13) The NEAMS program website (neams.energy.gov) is nearly ready to launch.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Integrated Modeling and Simulation Capability For Full Scale Multi-Physics Simulation and Visualization of Microreactor Concept

Microreactors are small mobile transportable nuclear reactors with thermal power less than 20 MWt that are designed to meet the needs of remote areas, military installations, emergency operations, and disaster relief zones. These reactor are concepts that are intended to be factory-manufacturable, easily transportable, and will allow for semi- or full-autonomous operation. To assist with development of microreactors, modeling and experimentation need to be performed with verification and validation to gain confidence in the models and tools. The objective of this study is apply advanced modeling and simulation capabilities being developed by the DOE Office of Nuclear Energy Nuclear Energy Advanced Modeling and Simulation (NEAMS) program to perform integrated modeling and simulation capability for multiphysics simulation of the microreactor concepts. This activity will provide an opportunity to exercise the codes to provide both gain experience and to provide feedback to the developers.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Problem Specification for FY12 Modeling of UNF During Extended Storage

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program of the Advanced Modeling and Simulation Office (AMSO) of the US Department of Energy, Office of Nuclear Energy (DOE/NE) has invested in the initial extension and application of advanced nuclear simulation tools to address relevant needs in evaluating the performance of used nuclear fuel (UNF) during extended periods of dry storage. There are many significant challenges associated with the prediction of the behavior of used fuel during extended periods of dry storage and subsequent transportation. The initial activities are focused on integrating with the Used Fuel Disposition (UFD) Campaign of the DOE/NE and a demonstration that the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code (AMPFuel) for modeling the mechanical state of the cladding after decades of storage. This initial focus will model the long-term storage of the UNF and account for the effect, and generation, of radially and circumferentially oriented hydride precipitates within the cladding and predict the end of storage (EOS) mechanical state (stress, strain) of the cladding. Predicting the EOS state of the cladding is significant because it (1) provides an estimate of the margin to failure of the cladding during nominal storage operation and it (2) establishes the initial state of the fuel for post-storage transportation. Because there are significant uncertainties associated with the storage conditions, hydride precipitate formation, and the beginning of storage (BOS) condition of the UNF, this will also allow for the development of a rigorous capability to evaluate the relative sensitivities of the uncertainties and can help to guide the experimental and analysis efforts of the UFD Campaign. This document is focused on specifying the problem that will be solved with AMPFuel. An associated report, documents the specifics of the constitutive model that will be developed and implemented in AMPFuel to account for the presence and predict the generation of the hydride precipitates. This report satisfies the deliverable for the DOE Office of Nuclear Energy, Advanced Modeling and Simulation Office, milestone M3MS-12OR0605083, 'Definition of Problem Specification,' which defines the problem to be solved that will satisfy milestone M2MS-12OR0605081, 'Demonstration of the Advanced Multi-Physics (AMP) Nuclear Fuel Performance code for modeling UFD.' This document should provide sufficient detail to model a high burnup pressurized water reactor (PWR) fuel rod to provide an estimate of the end of storage (EOS) mechanical state of the cladding. The fuel rod and irradiation history are based on seven cycles of irradiation in the CP and L H.B. Robinson nuclear reactor, which achieved a discharge burnup of 66.682 MWd/kgU. The fuel has been experimentally examined for storage conditions by Argonne National Laboratory for the NRC. In addition, we have compiled a list of key factors that have been shown to strongly influence the EOS state of the fuel and have identified baseline values and ranges of uncertainties that will be considered. The simulations that will be performed have been described in detail and include the modeling assumptions and boundary conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Application of the SHARP Toolkit to Sodium-Cooled Fast Reactor Challenge Problems

The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. To better understand and exploit the benefits of advanced modeling simulations, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task, which include the assessment of hot channel factors (HCFs) and the demonstration of zooming capability using the SHARP toolkit. If both challenge problems are resolved through advanced modeling and simulation using the SHARP toolkit, the economic competitiveness of a SFR can be significantly improved. The efforts in the first year of this project focused on the development of computational models, meshes, and coupling procedures for multi-physics calculations using the neutronics (PROTEUS) and thermal-hydraulic (Nek5000) components of the SHARP toolkit, as well as demonstration of the HCF calculation capability for the 100 MWe Advanced Fast Reactor (AFR-100) design. Testing the feasibility of the SHARP zooming capability is planned in FY 2018. The HCFs developed for the earlier SFRs (FFTF, CRBR, and EBR-II) were reviewed, and a subset of these were identified as potential candidates for reduction or elimination through high-fidelity simulations. A one-way offline coupling method was used to evaluate the HCFs where the neutronics solver PROTEUS computes the power profile based on an assumed temperature, and the computational fluid dynamics solver Nek5000 evaluates the peak temperatures using the neutronics power profile. If the initial temperature profile used in the neutronics calculation is reasonably accurate, the one-way offline method is valid because the neutronics power profile has weak dependence on small temperature variation. In order to get more precise results, the proper temperature profile for initial neutronics calculations was obtained from the STAR-CCM+ calculations. The HCFs of the peak temperatures at cladding outer surface, cladding inner wall surface, and fuel centerline induced by cladding manufacturing tolerance and uncertainties on the cladding, coolant, and fuel properties were evaluated for the AFR-100. Some assessment on the effect of wire wrap configuration and size of the bundle shows that it is practical to use the 7-pin bare rod bundle to calculate the HCFs. The resulting HCFs obtained from the high-fidelity SHARP calculations are generally smaller than those developed for the earlier SFRs because the most uncertainties involved in the modeling and simulations were disappeared. For completeness, additional investigations are planned in FY 2018, which will use random sampling techniques.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Sockeye Heat Pipe Analysis Code Verification and Validation

Some of the most promising microreactor designs currently under development utilize heat pipe technology to transfer heat from the reactor core to the secondary side heat exchanger, due to the technology’s compactness, efficiency, passivity, and reliability. Sockeye is an engineering-scale heat pipe tool developed under the Nuclear Energy Advanced Modeling and Simulation Program to be used for the design and safety analysis of microreactors. Sockeye’s core capability lies in a 1D, two-phase, compressible flow model, used to simulate the working fluid inside the heat pipe. Sockeye is built on the Multiphysics Object-Oriented Simulation Environment framework, which allows for seamless multiphysics coupling with other Nuclear Energy Advanced Modeling and Simulation tools and can thus be used in a full-scale simulation of a microreactor assembly, which can include hundreds of heat pipes. This paper presents some initial verification and validation assessments performed for Sockeye. We demonstrate good agreement between Sockeye’s numerical results and steady-state analytic solutions for velocity and pressure drop, with differences attributable to the underlying assumptions made by the analytic solutions. We also show that Sockeye reproduces analytic predictions of key operational limits, such as the capillary limit and the sonic limit. Finally, we compare Sockeye results to experimental data for the SAFE-30 heat pipe module test.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

ANS MultiApps tutorial

The Multiphysics Object-Oriented Simulation Environment (MOOSE) [1] is an open-source, parallel finite element framework which provides the foundation formany advanced modeling and simulation tools developed under the Department of Energy’s (DOE’s) Nuclear Energy Advanced Modeling and Simulation (NEAMS) program [2] for the analysis of advanced reactors. The MOOSE framework provides the common foundational capability on which many NEAMS codes for reactor analysis are built. The MOOSE framework also includes several systems to assemble unique workflows and coupling amongMOOSE-based applications. In particular, the MultiApp and Transfer systems are widely used to assemble different MOOSE-based or MOOSE- wrapped physics applications together to perform loosely or tightly coupled multiphysics simulations. The National Reactor Innovation Center’s (NRIC’s) Virtual Test Bed (VTB) [3] hosts publicly available nuclear reactor multiphysics simulation examples which leverage MOOSE’s MultiApp system to meet the modeling needs of different reactor types. The flexibility and robustness of coupling provided by MOOSE permits rapid development of coupled physics models for a wide range of reactor types and events.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Seismic Soil-Structure Interaction Analyses of a Deeply Embedded Model Reactor – SASSI Analyses

This report summarizes the SASSI analyses of a deeply embedded reactor model performed by BNL and CJC and Associates, as part of the seismic soil-structure interaction (SSI) simulation capability project for the NEAMS (Nuclear Energy Advanced Modeling and Simulation) Program of the Department of Energy. The SASSI analyses included three cases: 0.2 g, 0.5 g, and 0.9g, all of which refer to nominal peak accelerations at the top of the bedrock. The analyses utilized the modified subtraction method (MSM) for performing the seismic SSI evaluations. Each case consisted of two analyses: input motion in one horizontal direction (X) and input motion in the vertical direction (Z), both of which utilized the same in-column input motion. Besides providing SASSI results for use in comparison with the time domain SSI results obtained using the DIABLO computer code, this study also leads to the recognition that the frequency-domain method should be modernized so that it can better serve its mission-critical role for analysis and design of nuclear power plants.

42 ENGINEERING↗
DOE OSTI2016

The Nuclear Energy Knowledge and Validation Center Summary of Activities Conducted in FY16

The Nuclear Energy Knowledge and Validation Center (NEKVaC) is a new initiative by the Department of Energy (DOE) and Idaho National Laboratory (INL) to coordinate and focus the resources and expertise that exist with the DOE toward solving issues in modern nuclear code validation and knowledge management. In time, code owners, users, and developers will view the NEKVaC as a partner and essential resource for acquiring the best practices and latest techniques for validating codes, providing guidance in planning and executing experiments, facilitating access to and maximizing the usefulness of existing data, and preserving knowledge for continual use by nuclear professionals and organizations for their own validation needs. The scope of the NEKVaC covers many interrelated activities that will need to be cultivated carefully in the near term and managed properly once the NEKVaC is fully functional. Three areas comprise the principal mission: (1) identify and prioritize projects that extend the field of validation science and its application to modern codes, (2) develop and disseminate best practices and guidelines for high-fidelity multiphysics/multiscale analysis code development and associated experiment design, and (3) define protocols for data acquisition and knowledge preservation and provide a portal for access to databases currently scattered among numerous organizations. These mission areas, while each having a unique focus, are interdependent and complementary. Likewise, all activities supported by the NEKVaC, both near term and long term, must possess elements supporting all three areas. This cross cutting nature is essential to ensuring that activities and supporting personnel do not become “stove piped” (i.e., focused a specific function that the activity itself becomes the objective rather than achieving the larger vision). This report begins with a description of the mission areas; specifically, the role played by each major committee and the types of activities for which they are responsible. It then lists and describes the proposed near term tasks upon which future efforts can build.

##CAI suggestions##↗
DOE OSTI2015

Warthog: A MOOSE-Based Application for the Direct Code Coupling of BISON and PROTEUS

The Nuclear Energy Advanced Modeling and Simulation (NEAMS) program from the Department of Energy's Office of Nuclear Energy provides a robust toolkit for the modeling and simulation of current and future advanced nuclear reactor designs. This toolkit provides these technologies organized across product lines: two divisions targeted at fuels and end-to-end reactor modeling, and a third for integration, coupling, and high-level workflow management. The Fuels Product Line and the Reactor Product line provide advanced computational technologies that serve each respective field well, however, their current lack of integration presents a major impediment to future improvements of simulation solution fidelity. There is a desire for the capability to mix and match tools across Product Lines in an effort to utilize the best from both to improve NEAMS modeling and simulation technologies. This report details a new effort to provide this Product Line interoperability through the development of a new application called Warthog. This application couples the BISON Fuel Performance application from the Fuels Product Line and the PROTEUS Core Neutronics application from the Reactors Product Line in an effort to utilize the best from all parts of the NEAMS toolkit and improve overall solution fidelity of nuclear fuel simulations. To achieve this, Warthog leverages as much prior work from the NEAMS program as possible, and in doing so, enables interoperability between the disparate MOOSE and SHARP frameworks, and the libMesh and MOAB mesh data formats. This report describes this work in full. We begin with a detailed look at the individual NEAMS framework technologies used and developed in the various Product Lines, and the current status of their interoperability. We then introduce the Warthog application: its overall architecture and the ways it leverages the best existing tools from across the NEAMS toolkit to enable BISON-PROTEUS integration. Furthermore, we show how Warthog leverages a tool known as DataTransferKit to seamlessly enable the transfer for solution data between disparate frameworks and mesh formats. To end, we demonstrate tests for the direct software coupling of BISON and PROTEUS using Warthog, and discuss current impediments and solutions to the construction of physically realistic input models for this coupled BISON-PROTEUS system.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Advanced Fuels Campaign 2019 Accomplishments

The mission of the Advanced Fuels Campaign (AFC) is to perform or support research, development, and demonstration (RD&D) activities to identify and mature innovative fuels and cladding materials and associated technologies with the potential to improve the performance and safety of current and future reactors; increase the efficient utilization of nuclear energy resources; contribute to enhancing proliferation resistance of the nuclear fuel cycle; and address challenges related to waste management and ultimate disposal. AFC pursues its mission objectives using a goal-oriented, science-based approach that seeks to establish a fundamental understanding of fuel and cladding behaviors under conditions that arise during fabrication, normal steady-state irradiation, off-normal transient scenarios, and storage/disposal. This approach includes advancing the theoretical understanding of fuel behavior, conducting fundamental and integral experiments, and supporting the mechanistic, multi-scale modeling of nuclear fuels to inform and guide fuel development projects, advance the technological readiness of promising fuel candidates, and ultimately support fuel qualification and licensing initiatives. In the area of advanced tools for the modeling and simulation of nuclear fuels, the AFC works in close partnership with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, participating with NEAMS in developing mechanistic fuel behavior models and providing experimental data to inform and validate its most advanced tools. Specifically, AFC objectives in thecoming five year horizon include: 1. support the industry-led development of Accident Tolerant Fuel (ATF) technologies with improved reliability and performance under normal operations and enhanced tolerance during hypothetical accident scenarios, with implementation of lead test rods/assemblies of one or more ATF concepts in commercial reactor(s) by or before 2022 and batch reloads of near-term concept(s) in the 2023-2026 timeframe; 2. lead research and development on innovative fuel and cladding technologies for applications to future advanced reactors, especially fast-spectrum reactors, including reactors that utilize both once-through and recycle scenarios; 3. continue the development and demonstration of a multi-scale,science-based approach to fuel development and testing, andcontribute to the establishment of a state-of-the-art R&D infrastructure necessary to accelerate the development of new fuel concepts; and collaborate with NEAMS on the development and validation of multi-scale, multi-physics, and increasingly predictive fuel performance models and codes. 4. collaborate with NEAMS on the development and validation of multi-scale, multi-physics, and increasingly predictive fuel performance models and codes.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Applications of SHARP Toolkit to SFR Challenging Problems: Evaluation of Hot Channel Factors and Demonstration of Zooming Capability

The Simulation-based High-efficiency Advanced Reactor Prototyping (SHARP) toolkit is under development by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Campaign of the U.S. Department of Energy, Office of Nuclear Energy. Beginning in FY17, the NEAMS Campaign initiated the “Sodium-Cooled Fast Reactor (SFR) Challenge Problems” task to apply its advanced modeling and simulation capabilities to important SFR applications. The scope of work includes two major items: the assessment of hot channel factors (HCFs) and the demonstration of a local refinement “zooming” capability, both using the SHARP toolkit. Resolving these challenge problems with the SHARP toolkit can significantly improve the economic competitiveness of an SFR as well as make the tools more accessible for full core confirmatory calculations.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

The Virtual Test Bed (VTB) repository: a library of multiphysics reference reactor models using NEAMS tools

With the next generation of nuclear reactors under development, modeling and simulation (MS) tools are being developed by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in order to support their design, licensing, and future operation. Mirroring the physical test beds currently under construction (i.e., EBR-II and ZPPR), the Virtual Test Bed (VTB) was launched by the National Reactor Innovation Center (NRIC) in collaboration with NEAMS to support the advanced reactor community. This collaborative effort, which involves multiple teams at both Idaho National Laboratory and Argonne National Laboratory aims to use NEAMS tools to model a wide range of reactor designs. Those models are automatically tested to ensure their continued functionality as the tools are further developed. Examples are extensively documented, each acting as a tutorial for applying the relevant NEAMS tools to that reactor design. Currently, five advanced reactor types (with a total of eight specific design variants) are simulated by a variety of different models. These models range from steady-state, core multiphysics simulations to integrated plant analysis during loss-of flow transients. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗