Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “NEAMS”

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 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Inl Open Ondemand Applications

Open OnDemand is a software tool that is used to access HPC resources. It provides a framework for organizations to create apps and other additional functionality that may be useful to the organization. This code expands upon the pre-existing INL applications, including the NEAMS Workbench application, MOOSE Herd applications, and others. These changes significantly expand upon the functionality originally provided. Due to the extensive functionality that we added, these changes would not be added to the original application but would function as a standalone application that other organizations would be able to utilize on their own systems.

Biggs, BrandonS.↗
DOE OSTI2022

Marmot V2

MARMOT is a robust numerical tool for mesoscale modeling of fuel performance developed under the NEAMS Fuels technical area to predict the coevolution of microstructure and properties in fuel and cladding materials. MARMOT accomplishes this using the phase field method coupled with finite strain mechanics and heat conduction. MARMOT is based on the open source Multiphysics Object-Oriented Simulation Environment (MOOSE) and solves the coupled partial differential equations defining the physics using the finite element method. MARMOT is being developed in order to facilitate the development of improved materials models for fuel performance, but it is also being developed as a powerful tool in and of itself for the simulation of mesoscale fuel performance.

Aagesen, LarryK.↗
DOE OSTI2025

SEFOR Core I-E

This repository contains the VTB model developed for the Southwest Experimental Fast Oxide Reactor (SEFOR) core configuration I-E using the NEAMS tools. The MOOSE Reactor Module was employed to set up the geometry mesh, MC2-3 was applied to generate multigroup cross sections at various temperatures, and Griffin was used to calculate the k-eff values of Core I-E across these temperature conditions. To validate the deterministic results from Griffin, reference solutions were obtained with the Monte Carlo code Shift. The reactivity feedback was then derived and compared against SEFOR experimental data from isothermal tests in order to assess the performance of Griffin. Reproducing these results requires licensed access to several code systems, including Shift, Griffin, and MC2-3. Instructions for obtaining and accessing these codes can be found at the following locations: https://www.ornl.gov/scale/releases for SCALE/Shift, https://mooseframework.inl.gov/ncrc/applications/ncrc_root_griffin.html for Griffin, and https://www.anl.gov/nse/mc23 for MC2-3.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

A Hierarchical Upscaling Method for Predicting Strength of Materials under Thermal, Radiation and Mechanical loading - Irradiation Strengthening Mechanisms in Stainless Steels

Stainless steels based on Fe-Cr-Ni alloys are the most popular structural materials used in reactors. High energy particle irradiation of in this kind of polycrystalline structural materials usually produces irradiation hardening and embrittlement. The development of predictive capability for the influence of irradiation on mechanical behavior is very important in materials design for next-generation reactors. Irradiation hardening is related to structural information crossing different length scale, such as composition, dislocation, crystal orientation distribution and so on. To predict the effective hardening, the influence factors along different length scales should be considered. A multiscale approach was implemented in this work to predict irradiation hardening of iron based structural materials. Three length scales are involved in this multiscale model: nanometer, micrometer and millimeter. In the microscale, molecular dynamics (MD) was utilized to predict on the edge dislocation mobility in body centered cubic (bcc) Fe and its Ni and Cr alloys. On the mesoscale, dislocation dynamics (DD) models were used to predict the critical resolved shear stress from the evolution of local dislocation and defects. In the macroscale, a viscoplastic self-consistent (VPSC) model was applied to predict the irradiation hardening in samples with changes in texture. The effects of defect density and texture were investigated. Simulated evolution of yield strength with irradiation agrees well with the experimental data of irradiation strengthening of stainless steel 304L, 316L and T91. This multiscale model we developed in this project can provide a guidance tool in performance evaluation of structural materials for next-generation nuclear reactors. Combining with other tools developed in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the models developed will have more impact in improving the reliability of current reactors and affordability of new reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Progress toward bridging from atomistic to continuum modeling to predict nuclear waste glass dissolution.

This report summarizes research performed for the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Subcontinuum and Upscaling Task. The work conducted focused on developing a roadmap to include molecular scale, mechanistic information in continuum-scale models of nuclear waste glass dissolution. This information is derived from molecular-scale modeling efforts that are validated through comparison with experimental data. In addition to developing a master plan to incorporate a subcontinuum mechanistic understanding of glass dissolution into continuum models, methods were developed to generate constitutive dissolution rate expressions from quantum calculations, force field models were selected to generate multicomponent glass structures and gel layers, classical molecular modeling was used to study diffusion through nanopores analogous to those in the interfacial gel layer, and a micro-continuum model (K{mu}C) was developed to study coupled diffusion and reaction at the glass-gel-solution interface.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2011

Strategic Plan for Nuclear Energy -- Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

The Nuclear Energy Computational Fluid Dynamics Advanced Modeling and Simulation (NE-CAMS) system is being developed at the Idaho National Laboratory (INL) in collaboration with Bettis Laboratory, Sandia National Laboratory (SNL), Argonne National Laboratory (ANL), Utah State University (USU), and other interested parties with the objective of developing and implementing a comprehensive and readily accessible data and information management system for computational fluid dynamics (CFD) verification and validation (V&V) in support of nuclear energy systems design and safety analysis. The two key objectives of the NE-CAMS effort are to identify, collect, assess, store and maintain high resolution and high quality experimental data and related expert knowledge (metadata) for use in CFD V&V assessments specific to the nuclear energy field and to establish a working relationship with the U.S. Nuclear Regulatory Commission (NRC) to develop a CFD V&V database, including benchmark cases, that addresses and supports the associated NRC regulations and policies on the use of CFD analysis. In particular, the NE-CAMS system will support the Department of Energy Office of Nuclear Energy Advanced Modeling and Simulation (NEAMS) Program, which aims to develop and deploy advanced modeling and simulation methods and computational tools for reliable numerical simulation of nuclear reactor systems for design and safety analysis. Primary NE-CAMS Elements There are four primary elements of the NE-CAMS knowledge base designed to support computer modeling and simulation in the nuclear energy arena as listed below. Element 1. The database will contain experimental data that can be used for CFD validation that is relevant to nuclear reactor and plant processes, particularly those important to the nuclear industry and the NRC. Element 2. Qualification standards for data evaluation and classification will be incorporated and applied such that validation data sets will result in well-defined, well-characterized data. Element 3. Standards will be established for the design and operation of experiments for the generation of new validation data sets that are to be submitted to NE-CAMS that addresses the completeness and characterization of the dataset. Element 4. Standards will be developed for performing verification and validation (V&V) to establish confidence levels in CFD analyses of nuclear reactor processes; such processes will be acceptable and recognized by both CFD experts and the NRC.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2011

Strategic Plan for Nuclear Energy – Knowledge Base for Advanced Modeling and Simulation (NE-KAMS)

NE-KAMS knowledge base will assist computational analysts, physics model developers, experimentalists, nuclear reactor designers, and federal regulators by: (1) Establishing accepted standards, requirements and best practices for V&V and UQ of computational models and simulations, (2) Establishing accepted standards and procedures for qualifying and classifying experimental and numerical benchmark data, (3) Providing readily accessible databases for nuclear energy related experimental and numerical benchmark data that can be used in V&V assessments and computational methods development, (4) Providing a searchable knowledge base of information, documents and data on V&V and UQ, and (5) Providing web-enabled applications, tools and utilities for V&V and UQ activities, data assessment and processing, and information and data searches. From its inception, NE-KAMS will directly support nuclear energy research, development and demonstration programs within the U.S. Department of Energy (DOE), including the Consortium for Advanced Simulation of Light Water Reactors (CASL), the Nuclear Energy Advanced Modeling and Simulation (NEAMS), the Light Water Reactor Sustainability (LWRS), the Small Modular Reactors (SMR), and the Next Generation Nuclear Power Plant (NGNP) programs. These programs all involve computational modeling and simulation (M&S) of nuclear reactor systems, components and processes, and it is envisioned that NE-KAMS will help to coordinate and facilitate collaboration and sharing of resources and expertise for V&V and UQ across these programs. In addition, from the outset, NE-KAMS will support the use of computational M&S in the nuclear industry by developing guidelines and recommended practices aimed at quantifying the uncertainty and assessing the applicability of existing analysis models and methods. The NE-KAMS effort will initially focus on supporting the use of computational fluid dynamics (CFD) and thermal hydraulics (T/H) analysis for M&S of nuclear reactor systems, components and processes, and will later expand to include materials, fuel system performance and other areas of M&S as time and funding allow.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2011

Fundamental Drop Dynamics and Mass Transfer Experiments to Support Solvent Extraction Modeling Efforts

In support of the Nuclear Energy Advanced Modeling Simulation Safeguards and Separations (NEAMS SafeSep) program, the Idaho National Laboratory (INL) worked in collaboration with Los Alamos National Laboratory (LANL) to further a modeling effort designed to predict mass transfer behavior for selected metal species between individual dispersed drops and a continuous phase in a two phase liquid-liquid extraction (LLE) system. The purpose of the model is to understand the fundamental processes of mass transfer that occur at the drop interface. This fundamental understanding can be extended to support modeling of larger LLE equipment such as mixer settlers, pulse columns, and centrifugal contactors. The work performed at the INL involved gathering the necessary experimental data to support the modeling effort. A custom experimental apparatus was designed and built for performing drop contact experiments to measure mass transfer coefficients as a function of contact time. A high speed digital camera was used in conjunction with the apparatus to measure size, shape, and velocity of the drops. In addition to drop data, the physical properties of the experimental fluids were measured to be used as input data for the model. Physical properties measurements included density, viscosity, surface tension and interfacial tension. Additionally, self diffusion coefficients for the selected metal species in each experimental solution were measured, and the distribution coefficient for the metal partitioning between phases was determined. At the completion of this work, the INL has determined the mass transfer coefficient and a velocity profile for drops rising by buoyancy through a continuous medium under a specific set of experimental conditions. Additionally, a complete set of experimentally determined fluid properties has been obtained. All data will be provided to LANL to support the modeling effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2012

Multispecies Diffusion Capability For The AMP Nuclear Fuel Performance Code (LANL Milestone M31MS060301 Final Report)

This work addresses only diffusion. The contact solver in AMP was not sufficiently developed this year to attempt treatment of species contact. A cylindrical tensor diffusion coefficient model was added to the AMP code, with the KHHS model [1] implemented into the AMP material library as a specific example. A cylindrical tensor diffusion operator manufactured solution verification example was coded. Before meeting the full text of the milestone task, it remains to: (1) code and run a cylindrical tensor diffusion solver manufactured solution (2) code and run the validation example of [1] (3) document results. These are dependent on developing new capabilities for the AMP code requiring close collaboration with the AMP team at ORNL. The model implemented provides a good intermediate first step toward a general multi-species solver. The multi-species capability of the AMP nuclear fuel code [2] is intended to allow the modeling of radiation-driven redistribution of various elements through solid metal nuclear reactor fuels. The initial model AMP provides for U-Pu-Zr fuels is based on the analysis of the Integral Fast Reactor (IFR) fuel development program experiment X419 post-irradiation data described in [1], referred to here as the KHHS model. This model may be specific to that experiment, but it was thought to provide a good start for the AMP code, because it (1) is formulated at the engineering scale, (2) decouples the species from each other, (3) predetermines the phase boundaries so that reference to a phase diagram is not needed, and (4) one of the authors (Hayes) was the NEAMS Fuels IPSC manager for FY11. The KHHS model is formulated for radial fluxes as little axial redistribution is seen experimentally. As U-Pu-Zr fuel is irradiated, the constituents migrate to form three annular regions. The center region is Zr-enriched and U-depleted, the middle region is Zr-depleted and U-enriched, and the outer region is Zr-enriched and U-depleted. The Pu concentration stays roughly constant throughout with slight enrichment in the center and depletion near the surface. Pu acts as a solvent for the mixture. The experiment was only run to 1.9% burnup, so the model is not at this time applicable to the high-burnup scenarios that the AMP code is intended to eventually model.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Uranium vacancy mobility at the sigma Σ5 symmetric tilt grain boundary in UO 2

An important consequence of the fissioning process occurring during burnup is the formation of fission products. These fission products alter the thermo-mechanical properties of the fuel. They also lead to macroscopic changes in the fuel structure, including the formation of bubbles that are connected to swelling of the fuel. Subsequent release of fission gases increase the pressure in the plenum and can cause changes in the properties of the fuel pin itself. It is thus imperative to understand how fission products, and fission gases in particular, behave within the fuel in order to predict the performance of the fuel under operating conditions. Fission gas redistribution within the fuel is governed by mass transport and the presence of sinks such as impurities, dislocations, and grain boundaries. Thus, to understand how the distribution of fission gases evolves in the fuel, we must understand the underlying transport mechanisms, tied to the concentrations and mobilities of defects within the material, and how these gases interact with microstructural features that might act as sinks. Both of these issues have been addressed in previous work under NEAMS. However, once a fission product has reached a sink, such as a grain boundary, its mobility may be different there than in the grain interior and predicting how, for example, bubbles nucleate within grain boundaries necessitates an understanding of how fission gases diffuse within boundaries. That is the goal of the present work. In this report, we describe atomic level simulations of uranium vacancy diffusion in the pressence of a Σ5 symmetric tilt boundary in urania (UO 2 ). This boundary was chosen as it is the simplest of the boundaries we considered in previous work on segregation and serves as a starting point for understanding defect mobility at boundaries. We use a combination of molecular statics calculations and kinetic Monte Carlo (kMC) to determine how the mobility of uranium vacancies is altered at this particular grain boundary. Given that the diffusion of fission gases such as Xe are tied to the mobility of uranium vacancies, these results given insight into how fission gas mobility differs at grain boundaries compared to bulk urania.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Adding a MOAB Geometry Interface to SHARP Structural Mechanics

The authors briefly summarize the development of, and test experience with, an initial data interface between the structural mechanics code Diablo and the SHARP reactor simulation system data hub MOAB. That interface has been exercised both to write MOAB databases from Diablo, and then also to use such a database to read in part of a simulation definition for a subsequent Diablo execution. All enhancements are integrated into the central Diablo source repository. The SHARP software system for advanced simulation of nuclear reactors and power plant systems is sponsored by DOE's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. SHARP has been architected as a federation of single-physics simulation tools to permit flexibility in programming languages and leveraging of past and on-going investments. Solution of multi-physics problems will be coordinated by, and data passed through, a central 'hub'. SHARP's hub implementation is utilizing MOAB: a Mesh-Oriented datABase. This same data hub approach is also intended to enable multi-resolution simulations, e.g, lower-dimension plant-scale simulations can be informed by high-fidelity 3D models of particular critical components.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2012

FY2012 summary of tasks completed on PROTEUS-thermal work

PROTEUS is a suite of the neutronics codes, both old and new, that can be used within the SHARP codes being developed under the NEAMS program. Discussion here is focused on updates and verification and validation activities of the SHARP neutronics code, DeCART, for application to thermal reactor analysis. As part of the development of SHARP tools, the different versions of the DeCART code created for PWR, BWR, and VHTR analysis were integrated. Verification and validation tests for the integrated version were started, and the generation of cross section libraries based on the subgroup method was revisited for the targeted reactor types.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Scoping analysis of the Advanced Test Reactor using SN2ND

A detailed set of calculations was carried out for the Advanced Test Reactor (ATR) using the SN2ND solver of the UNIC code which is part of the SHARP multi-physics code being developed under the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in DOE-NE. The primary motivation of this work is to assess whether high fidelity deterministic transport codes can tackle coupled dynamics simulations of the ATR.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

A VU Governance Process Applied to a Bison Fuel Rod Validation Exercise

A VU governance process, known as Discovery, Accumulation, and Assessment, was applied to a NEAMS Bison Fuel Rod Validation Exercise. The main outcomes were a VU plan for the Bison Validation effort and an improved DAA process. This document provides a sketch of the DAA process and its implementation in the Synopsis software. The application to Bison showed a tendency to skip important parts of the governance process. This points out the need for mechanisms such as Synopsis to improve VU governance. An appendix shows the data and information collected during the process, as it pertains to the Bison Validation effort.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Large Scale DD Simulation Results for Crystal Plasticity Parameters in Fe-Cr And Fe-Ni Systems

The development of viable nuclear energy source depends on ensuring structural materials integrity. Structural materials in nuclear reactors will operate in harsh radiation conditions coupled with high level hydrogen and helium production, as well as formation of high density of point defects and defect clusters, and thus will experience severe degradation of mechanical properties. Therefore, the main objective of this work is to develop a capability that predicts aging behavior and in-service lifetime of nuclear reactor components and, thus provide an instrumental tool for tailoring materials design and development for application in future nuclear reactor technologies. Towards this end goal, the long term effort is to develop a physically based multiscale modeling hierarchy, validated and verified, to address outstanding questions regarding the effects of irradiation on materials microstructure and mechanical properties during extended service in the fission and fusion environments. The focus of the current investigation is on modern steels for use in nuclear reactors including high strength ferritic-martensitic steels (Fe-Cr-Ni alloys). The effort is to develop a predicative capability for the influence of irradiation on mechanical behavior. Irradiation hardening is related to structural information crossing different length scales, such as composition, dislocation, and crystal orientation distribution. To predict effective hardening, the influence factors along different length scales should be considered. Therefore, a hierarchical upscaling methodology is implemented in this work in which relevant information is passed between models at three scales, namely, from molecular dynamics to dislocation dynamics to dislocation-based crystal plasticity. The molecular dynamics (MD) was used to predict the dislocation mobility in body centered cubic (bcc) Fe and its Ni and Cr alloys. The results are then passed on to dislocation dynamics to predict the critical resolved shear stress (CRSS) from the evolution of local dislocation and defects. In this report the focus is on the results obtained from large scale dislocation dynamics simulations. The effect of defect density, materials structure was investigated, and evolution laws are obtained. These results will form the bases for the development of evolution and hardening laws for a dislocation-based crystal plasticity framework. The hierarchical upscaling method being developed in this project can provide a guidance tool to evaluate performance of structural materials for next-generation nuclear reactors. Combined with other tools developed in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, the models developed will have more impact in improving the reliability of current reactors and affordability of new reactors.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Verification test suite for systems analysis tools

This report documents strategies and test problems to support Verification and Validation (V&V) of the systems analysis tools currently being developed under the Reactor Product Line (RPL) of DOE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. A suite of verification test cases has been developed. This test suite supports development of new codes as well as maintenance of existing codes. For advanced tools that are under development like RELAP-7 for light water reactor analysis [1] and the MOOSE based component models for SFR (M-SFR) analyses [2], verification of simulation results must be performed throughout the entire development phase. This test suite is also useful for established codes as part of automated testing to ensure that general code maintenance does not change existing results.

Sumner, T.↗
DOE OSTI2013

Sensitivity Analysis of the Fission Gas Behavior Model in BISON

This report summarizes the result of a NEAMS project focused on sensitivity analysis of a new model for the fission gas behavior (release and swelling) in the BISON fuel performance code of Idaho National Laboratory. Using the new model in BISON, the sensitivity of the calculated fission gas release and swelling to the involved parameters and the associated uncertainties is investigated. The study results in a quantitative assessment of the role of intrinsic uncertainties in the analysis of fission gas behavior in nuclear fuel.

Swiler, Laura Painton↗