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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.

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All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Leveraging Intrinsic Principal Directions for Multifidelity Uncertainty Quantification

In this work we propose an approach for accelerating Uncertainty Quantification (UQ) analysis in the context of Multifidelity applications. In the presence of complex multiphysics applications, which often require a prohibitive computational cost for each evaluation, multifidelity UQ techniques try to accelerate the convergence of statistics by leveraging the in- formation collected from a larger number of a lower fidelity model realizations. However, at the-state-of-the-art, the performance of virtually all the multifidelity UQ techniques is related to the correlation between the high and low-fidelity models. In this work we proposed to design a multifidelity UQ framework based on the identification of independent important directions for each model. The main idea is that if the responses of each model can be represented in a common space, this latter can be shared to enhance the correlation when the samples are drawn with respect to it instead of the original variables. There are also two main additional advantages that follow from this approach. First, the models might be correlated even if their original parametrizations are chosen independently. Second, if the shared space between models has a lower dimensionality than the original spaces, the UQ analysis might benefit from a dimension reduction standpoint. In this work we designed this general framework and we also tested it on several test problems ranging from analytical functions for verification purpose, up to more challenging application problems as an aero-thermo-structural analysis and a scramjet flow analysis.

42 ENGINEERING↗
DOE OSTI2018

Report on Aging and Irradiation Theoretical Model Development for Alloy 690 in Grizzly

This report describes the current efforts to implement a model to predict the engineering scale behavior of irradiated and aged Ni 2 Cr model alloy and Alloy 690 into the Grizzly software. Grizzly is a multiphysics simulation code under development at Idaho National Laboratory to simulate aging mechanisms and their effects on the integrity of critical LWR components. Engineering scale behavior of materials are strongly impacted by the underlying microstructure: the distribution and interaction of dislocations and defects within the individual grains dictates how the material will respond to loading. The evolution of these dislocations and defects determines how the engineering scale material will perform under different loading and environmental conditions. Mesoscale models can function as a bridge to connect microstructure evolution models to engineering models of the ductile-brittle transition temperature (DBTT) curve and fracture. The toughness of a material, which governs the engineering scale fracture behavior, is a function of both elastic deformation and plastic deformation. While the yield stress measurements capture the elastic contribution to a material’s toughness, information about the dislocation movement is required to understand the plastic component of toughness. Crystal plasticity models predict dislocation density evolution and thereby connect microstructure evolution to engineering scale behavior predictions. These models calculate the plastic strain within a metal as a function of dislocations and other crystal defects, including precipitates and interstitial loops, among others, by tracking dislocation movement and the interaction of these dislocations with the crystal defects. Within NiCr alloys, long-range ordered (LRO) precipitates form under aging conditions. These precipitates act as barriers to dislocation motion and are either sheared by dislocations or pin dislocations. Additional defects, such as stacking fault tetrahedra (SFTs) from irradiation damage, also impede dislocation motion in these alloys. A crystal plasticity mode for Ni 2 Cr model alloy and Alloy 690 is being currently implemented in the Grizzly code at INL. This report includes a description of the theoretical models incorporated into the Grizzly crystal plasticity code. These theoretical models include mechanisms for dislocation glade and twinning as well as dislocation-defect interaction relationship for both LRO precipitates and SFTs. We have selected these dislocation models based on the deformation mechanisms observed in experimental micropillar compression tests conducted on the Ni 2 Cr model alloy. The models are then extended to Alloy 690 with ongoing parameter fitting.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Advances in Geologic Disposal Safety Assessment and an Unsaturated Alluvium Reference Case

The Spent Fuel and Waste Science and Technology (SFWST) Campaign of the U.S. Depat ment of Energy (DOE) Office of Nuclear Energy (NE), Office of Fuel Cycle Technology (OFCT) is conducting research and development (R&D) on geologic disposal of spent nuclear fuel (SNF) and high level nuclear waste (HLW). Two high priorities for SFWST disposal R&D are design concept development and disposal system modeling (DOE 2011, Table 6). These priorities are directly addressed in the SFWST Geologic Disposal Safety Assessment (GDSA) work package, which is charged with developing a disposal system modeling and analysis capability for evaluating disposal system performance for nuclear waste in geologic media. This report describes specific GDSA activities in fiscal year 2018 (FY 2018) toward the development of GDSA Framework, an enhanced disposal system modeling and analysis capability for geologic disposal of nuclear waste. GDSA Framework employs the PFLOTRAN thermal-hydrologic-chemical multiphysics code (Hammond et al. 2011a; Lichtner and Hammond 2012) and the Dakota uncertainty sampling and propagation code (Adams et al. 2012; Adams et al. 2013). Each code is designed for massivelyparallel processing in a high-performance computing (HPC) environment. Multi-physics representations in PFLOTRAN are used to simulate various coupled processes including heat flow, fluid flow, waste dissolution, radionuclide release, radionuclide decay and ingrowth, precipitation and dissolution of secondary phases, and radionuclide transport through engineered barriers and natural geologic barriers to the biosphere. Dakota is used to generate sets of representative realizations and to analyze parameter sensitivity.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Dual-Purpose Canister Filling Demonstration Project Progress Report, Rev. 1

This report discusses the initial progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs)using filler materials to demonstrate that the probability of criticality in DPCs during disposal to be below the probability for inclusion in a repository performance assessment. In the initial phase of a multi-phase effort that will result in a full-scale demonstration, a computational fluid dynamics (CFD) model was developed to gauge the filling process and to uncover any unforeseen issues. The initial filling simulations of the lower region (mouse holes) of a prototypic DPC show successful removal of the inner space voids and smooth,even progression of the liquid level. In the initial phase, flow through a pipe that is similar to the drain pipe in a DPC will be investigated separately to gain valuable insight of flow regime inside a pipe. The initial experimental setups for validating the computational filling model have been designed,and the various assembly parts are being procured. The experience gained from the initial experiments will be applied to the next steps toward a full-scale demonstration and to the validation of multiphysics filling simulation models.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

Release 1.0 of MARMOT: A Mesoscale Fuel Performance Code

MARMOT is the mesoscale fuel performance code under development as part of the US DOE Nuclear Energy Advanced Modeling and Simulation Program. With this report, we release a new version of MARMOT, summarize its implemented capabilities, and document the ongoing development of new capabilities planned for the next release. The purpose of MARMOT is to predict the coevolution of microstructure and material properties of nuclear fuel and cladding due to stress, temperature, and irradiation damage. It accomplishes this using the phase field method coupled to solid mechanics and heat conduction. MARMOT is based on the Multiphysics Object-Oriented Simulation Environment (MOOSE), and much of its basic capability in the areas of the phase field method, mechanics, and heat conduction come directly from MOOSE modules.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Summary of BISON Development and Validation Activities - NEAMS FY15 Report

The objective of the NEAMS ToolKit is to develop a “pellet-to-plant” simulation capability useful for predicting performance and safety for a broad range of nuclear reactor power systems. The NEAMS ToolKit has been organized into a Fuels Product Line (FPL) and a Reactor Product Line (RPL) and is modular in design. Within the FPL, a multiscale approach has been adopted in which simulations of fuel performance at the engineering scale are informed by material property and irradiation performance models developed from mesoscale simulations of microstructural evolution. The focus in this report is on development and validation of the engineering-scale fuel performance analysis tool within the FPL, which is BISON. This summary report contains an overview of work performed under the work package enti- tled “FY2015 NEAMS INL-Engineering Scale Fuel Performance & Interface with RPL Tools.” A first chapter identifies the specific FY-15 milestones, providing a basic description of the asso- ciated work and references to related detailed documentation. Where applicable, a representative technical result is provided. A second chapter summarizes major additional accomplishments including 1) a BISON validation paper, 2) a multiphysics coupling paper, 3) application of BI- SON to Reactivity Insertion Accident analysis and 4) use of BISON for uncertainity analysis in fuel modeling. A final chapter outlines FY-16 future work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

High Brightness Superconducting RF Gun Cavity

The Phase I goals were to numerically optimize a TM020/TM010 SRF gun with a demountable back wall, using SLAC’s state-of-the-art massive parallel ACE3P suite of 3D Finite Element Analysis codes available at the National Energy Research Scientific Computing Center (NERSC) , which includes: a) TM020 cell shape, including impact of RF focusing on beam dynamics; b) optimum number of TM010-cells to achieve low emittances; c) multipacting simulations back wall opening; d) dark current investigations; e) studies on LOM and HOM damping with waveguides/port on TM020-cells using wakefield solver can be conducted as well but are not required stringently; f) conceptual design of a power input coupler; g) multiphysics simulations to optimize the gun thermal management and cooling requirements for a superconducting joint; h) multivariate beam dynamical optimization in conjunction with the gun optimization; and i) finalize an engineering design for vertical high power tests at JLab.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

HTTR 3 D Cross Section Generation with Serpent and MAMMOTH

The High Temperature Test Reactor (HTTR) is a graphite moderated thermal reactor with a nominal power output of 30 MWth. The purpose of this work is to prepare a set of neutron cross sections for use in the transient analyses of three Loss of Forced Cooling (LOFC) events planned for this reactor. The cross sections are generated with the Serpent 2 3 D Monte Carlo code to include the axial heterogeneity and the strong axial coupling of this core. The approach yields good agreement of results when compared to previously used 2 D lattice models, which require a fine spatial discretization and, surprisingly, comes at a higher computational cost. In addition, the temperature distribution must be considered to obtain reasonable cross sections and the axial leakage component dominates the spectral effects in the core. Consequently, a temperature model as a function of the fuel and moderator temperatures is adopted, and the core is depleted to 390 effective full-power days (EFPDs). A full tabulation of cross sections is prepared at this burnup point for a variety of fuel and moderator temperatures. Next, the multiphysics reactor application MAMMOTH is used to evaluate the quality of these cross sections using a diffusion solver. As expected, the homogenization error in the cross sections is significant and the Super Homogenization (SPH) correction from MAMMOTH is necessary to preserve key reaction rates. The SPH corrected MAMMOTH results are in excellent agreement with the Serpent results and reproduce both the reference power profile and temperature coefficients at each tabulation point. This fact confirms the accuracy of the SPH corrected cross sections and demonstrates the maturity of using Serpent 2 and MAMMOTH for 3 D cross section generation, even for reactors as complex as the HTTR. Finally, the authors recommend that the cross sections should be parametrized with the local burnup and that a full core burnup calculation with coupled thermal fluids would provide a better estimate of the initial condition for any subsequent transient analysis. Unfortunately, this was not within the scope of this work, and direct analysis with these tabulations should be considered as a first order approximation, at best.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Integrated FHR technology development: Tritium management, materials testing, salt chemistry control, thermal hydraulics and neutronics, associated benchmarking and commercial basis

This report summarizes the results of a 3-year Integrated Research Project (IRP) sponsored by the U.S. Department of Energy to address major technical challenges in the development of the Fluoride-salt-cooled High-temperature Reactor (FHR). The IRP universities included the Massachusetts Institute of Technology (MIT), the University of California at Berkeley (UCB), the University of Wisconsin (UW), and the University of New Mexico (UNM). The FHR is a new reactor concept less than 20 years old that combines (1) a clean fluoride salt coolant, (2) the graphite-matrix coated-particle fuel originally developed for High- Temperature Gas-Cooled Reactors and (3) passive decay heat removal systems from sodium fast reactors. The base-line design is a pebble-bed FHR that delivers heat between 600 to 700°C with a 7 Li 2 BeF 4 (flibe) salt coolant to the power cycle or industrial users. The Executive Summary describes what was accomplished and summarizes major conclusions. The main report contains stand-alone sections in each major area of work. Highlights are summarized below. Tritium. The coolant salt produces tritium necessitating understanding and control of tritium. Experimental work included measuring hydrogen uptake on carbon, tritium uptake during irradiation of graphite in 700°C salt and the permeability of tritium through metals with and without coatings. An FHR systems model was developed to predict tritium behavior and enable design of tritium control systems. The leading candidate for tritium removal from the flibe salt is carbon bed outside the reactor core. Thermal hydraulics and Neutronics. A 10-meter high FHR integral effects test facility, using Dowtherm A®, has been built and used to simulate the integral response of a representative FHR to improve understanding of FHR thermal hydraulics and validate thermal hydraulics codes used for the FHR. Neutronic analysis included code-to-code benchmarking, multiphysics modeling, uncertainty analysis for nuclear data uncertainties and shielding analysis. Twisted-tube heat exchangers were examined that may provide significant cost reductions because of their improved heat-transfer performance with liquid salts. The impacts on safety and design caused by uncertainties in physical properties of the salt were assessed with recommendations on where added physical property measurements are needed. Work began on understanding radiative heat transfer that becomes important in salt systems at higher temperatures. Materials. Static corrosion tests were conducted on graphite and materials in 700°C salt -- both in the UW laboratory and in the MIT reactor. In addition to static corrosion testing, a circulating salt corrosion loop was built and is operating. Methods to measure salt properties including redox were developed. The presence of carbon increases corrosion. With good redox control, the evidence indicates stainless steel can be used as a material of construction. Benchmarking. An additional series of workshops where held to enable benchmarking and comparison of results between different investigators. Commercialization. The FHR delivers higher average-temperature (~650°C) heat to the power cycle or industry than other reactors. This enables power cycles with lower-cost heat storage for variable electricity to the grid while the reactor operates at base-load -- providing a major competitive advantage by maximizing revenue from electricity sales. In October, 2016, a new startup company named Kairos Power was formed, and opened its first office in Jack London Square, Oakland, California, in January 2017. Kairos Power exists largely due to the work that was accomplished by this IRP and the previous DOE-funded FHR IRP that performed research on FHR technology from 2012 to 2015. In September, 2018, as this IRP is concluding, Kairos Power was staffed at 60 full time employees, working to develop technology derived from this IRP. Many of these employees worked with the IRP before joining Kairos Power 1 (https://kairospower.com). Other students and researchers from this IRP have joined other startup companies, including the Terrapower Molten Chloride Fast Reactor and the Terrestrial Energy development efforts.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Full-System Linearization for Floating Offshore Wind Turbines in OpenFAST: Preprint

The wind engineering community relies on multiphysics engineering software to run nonlinear time-domain simulations, e.g. for design-standards-based loads analysis. Although most physics involved in wind energy are nonlinear, linearization of the underlying nonlinear system equations is often advantageous to understand the system properties and exploit well-established methods and tools for analyzing linear systems. This paper presents the development of the new linearization functionality of the open-source engineering tool OpenFAST for floating offshore wind turbines (FOWTs), as well as the concepts and mathematical background needed to understand and apply it correctly.

17 WIND ENERGY↗
DOE OSTI2018

Evaluation of statistical variation of microstructural properties and temperature effects on creep fracture of Grade 91

This report describes three improvements made to a physically-based model for creep and creep rupture in Grade 91 steel developed as part of the Advanced Reactor Technologies program: 1. Progress on transitioning the model framework to the MOOSE finite element package to improve parallel scalability, increase the physical size of the simulations, and incorporate multiphysics effects into the model. 2. The development of a physically-based method of scaling the model parameters to accurately capture creep in Grade 91 for temperatures in the range of 450° to 500° C. 3. Extending the model to account for microstructural statistical variations, specifically capturing the effect of grain boundary energy on the physical parameters underling the model for grain boundary void nucleation and cavitation. Put together, these improvements result in an accurate, predictive model for the physical response of Grade 91 steel over the expected use temperature range for the material in future liquid metal cooled fast reactors. The model can be used to more accurately predict engineering properties of the material for very long service lives, which could lead to safer, more economical future advanced reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Dual-Purpose Canister Filling Demonstration Project Progress Report

This report discusses the initial progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs) using filler materials to demonstrate that the probability of criticality in DPCs during disposal to be below the probability for inclusion in a repository performance assessment. In the initial phase of a multi-phase effort that will result in a full-scale demonstration, a computational fluid dynamics (CFD) model was developed to gauge the filling process and to uncover any unforeseen issues. The initial filling simulations of the lower region (mouse holes) of a prototypic DPC show successful removal of the inner space voids and smooth, even progression of the liquid level. In the initial phase, flow through a pipe that is similar to the drain pipe in a DPC will be investigated separately to gain valuable insight of flow regime inside a pipe. The initial experimental setups for validating the computational filling model have been designed, and the various assembly parts are being procured. The experience gained from the initial experiments will be applied to the next steps toward a full-scale demonstration and to the validation of multiphysics filling simulation models.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Initial Assessment for a Generic Horizontal Fissile Solution Vessel with an Electron Beam-based Neutron Generator

The LANL NA-23 program office has provided an extensive design study for a medical isotope (i.e. Molybdenum-99) production facility based on a fissile solution vessel with various accelerator technologies. These concept designs and associated R&D have provided affordable pathways for system optimization for industry counter partner such as SHINE and Global Medial Isotope Supply (GMIS). A horizontal fissile solution vessel design with an electron beam (E-beam) accelerator is proposed and studied with modeling capabilities developed for previous concepts (i.e. a generic vertical fissile solution vessel with a deuterium-tritium (DT) accelerator). An initial system assessment is performed by estimating theoretical Molybdenum-99 production yield with MCNP6.2. A baseline system configuration is constructed based on early design discussions between LANL and GMIS. Since only an initial design concept is available at this stage, all associated system design parameters can be adjusted to fit the best system performance characteristics. In this work, unlike previous vertical system studies, the cooling configuration of the horizontal fissile solution vessel is not considered, and only fission reactions are used to quantify the theoretical product yield. Five parametric design studies are proposed, preliminary results are briefly reported, and extended design studies beyond what is proposed here are also discussed. It is concluded that a comprehensive system analysis via multiphysics coupled calculation developed by LANL should be further conducted for the final design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

SIERRA Code Coupling Module: Arpeggio User Manual (Version 4.46)

The SNL Sierra Mechanics code suite is designed to enable simulation of complex multiphysics scenarios. The code suite is composed of several specialized applications which can operate either in standalone mode or coupled with each other. Arpeggio is a supported utility that enables loose coupling of the various Sierra Mechanics applications by providing access to Framework services that facilitate the coupling. More importantly Arpeggio orchestrates the execution of applications that participate in the coupling. This document describes the various components of Arpeggio and their operability. The intent of the document is to provide a fast path for analysts interested in coupled applications via simple examples of its usage.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Bison Robustness and Performance Improvements

In last year’s Bison robustness report, the authors demonstrated that improving the Jacobian matrix can drastically increase solution convergence rates. The development of an automatic differentiation (AD) capability in the MOOSE framework has created the possibility of having exact Jacobian matrices for all of our material models. A large focus in this year’s report is the application of AD to MOOSE physics modules, such as the tensor mechanics module, and the application to Bison fuel and cladding material models. A second main focus in this year’s report is the investigation of a set of algorithmic solver changes to improve the robustness and performance of Bison LWR simulations. Beyond providing increased solver robustness, the ability to solve strongly coupled multiphysics problems, and solving large deformation problems with ease, these automatically generated perfect Jacobians also drastically reduce development time for new models. In models converted to AD, Jacobian computation made up the majority of the code base. With this code removed, both readability and maintainability of the MOOSE/Bison code base are improved. The work documented here significantly improved the robustness of the Bison LWR assessment case suite. Through algorithmic changes in solver infrastructure, such as the application of predictors and improvements in timestepping, tangible improvements in performance were observed. These improvements are documented using a new standard set of metrics that are collected in all assessment cases, and a suite of statistical analysis tools developed for this report.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

SIERRA Code Coupling Module: Arpeggio User Manual (V. 4.40)

The SNL Sierra Mechanics code suite is designed to enable simulation of complex multiphysics scenarios. The code suite is composed of several specialized applications which can operate either in standalone mode or coupled with each other. Arpeggio is a supported utility that enables loose coupling of the various Sierra Mechanics applications by providing access to Framework services that facilitate the coupling. More importantly Arpeggio orchestrates the execution of applications that participate in the coupling. This document describes the various components of Arpeggio and their operability. The intent of the document is to provide a fast path for analysts interested in coupled applications via simple examples of its usage.

42 ENGINEERING↗
DOE OSTI2019

Modeling Continuous-Flow Reactor Improvements over Batch Reactions for Enzyme Catalyzed Microsphere Surface Reactions

Our research objective was to study enzymatic reactions occurring on microsphere surfaces. Specifically, our work involved enzymatic polymerization reactions (chain lengthening) taking place both in batch and in a continuous flow reactor. My research involved creating simulations using COMSOL Multiphysics 5.4a to further study variable interactions within the continuous flow reactor. Our goal was to use suitable software to create a simulation platform suite that could be used for advanced research studies of continuous flow enzymatic surfaces reactions. In addition, we studied various characteristics of our present enzymatic polymerization reactor assembly, modeled inlet species flow (Species M) and estimated outlet species production (Species P). Our simulations spanned three levels of detail: Single-Bead, Packed-Bead Cell and Full Reactor Column. Furthermore, the Packed-Bead Cells were studied using three possible regular arrangement of solid spheres: Face-Centered Cubic (FCC), Hexagonal Close Packed (HCP) and Cubic Close Packed (CCP).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗