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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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At least 1,279 records · Page 71
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

Systems and methods of generating data marks in data visualizations

An example method of displaying a data visualization includes displaying a plurality of selectable fields and receiving user selections of two different fields from the plurality of selectable fields. The method also includes generating, in accordance with the received user selections, data marks to be displayed in a data visualization, each data mark corresponding to a respective retrieved tuple of data from a multidimensional database, where (i) each data mark has an x-position defined according to data for a first field in the respective tuple and (ii) each data mark has a y-position defined according to data for a second field in the respective tuple. The method also includes displaying the data visualization that includes the generated data marks.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

The principal component analysis method used with polynomial Chaos expansion to propagate uncertainties through critical transport problems

A critical neutron transport problem with random material properties is introduced. The total cross section and the average neutron multiplicity are assumed to be uncertain, characterized by the mean and variance with a log-normal distribution. The average neutron multiplicity and the total cross section are assumed to be uncorrected and the material properties for differing materials are also assumed to be uncorrected. The principal component analysis method is used to decompose the covariance matrix into eigenvalues and eigenvectors and then 'realizations' of the material properties can be computed. A simple Monte Carlo brute force sampling of the decomposed covariance matrix is employed to obtain a benchmark result for each test problem. In order to save computational time and to characterize the moments and probability density function of the multiplication factor the polynomial chaos expansion method is employed along with the stochastic collocation method. A Gauss-Hermite quadrature set is convolved into a multidimensional tensor product quadrature set and is successfully used to compute the polynomial chaos expansion coefficients of the multiplication factor. Finally, for a particular critical fuel pin assembly the appropriate number of random variables and polynomial expansion order are investigated. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Validation of the integration of CFD and SAS4A/SASSYS-1: Analysis of EBR-II shutdown heat removal test 17

Recent analyses have demonstrated the need to model multidimensional phenomena, particularly thermal stratification in outlet plena, during safety analyses of loss-of-flow transients of certain liquid-metal cooled reactor designs. Therefore, Argonne's reactor systems safety code SAS4A/SASSYS-1 is being enhanced by integrating 3D computational fluid dynamics models of the plena. A validation exercise of the new tool is being performed by analyzing the protected loss-of-flow event demonstrated by the EBR-II Shutdown Heat Removal Test 17. In this analysis, the behavior of the coolant in the cold pool is modeled using the CFD code STAR-CCM+, while the remainder of the cooling system and the reactor core are modeled with SAS4A/SASSYS-1. This paper summarizes the code integration strategy and provides the predicted 3D temperature and velocity distributions inside the cold pool during SHRT-17. The results of the coupled analysis should be considered preliminary at this stage, as the exercise pointed to the need to improve the CFD model of the cold pool tank. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

BWR transient analysis using neutronic / thermal hydraulic coupled codes including uncertainty quantification

The KIT is involved in the development and qualification of best estimate methodologies for BWR transient analysis in cooperation with industrial partners. The goal is to establish the most advanced thermal hydraulic system codes coupled with 3D reactor dynamic codes to be able to perform a more realistic evaluation of the BWR behavior under accidental conditions. For this purpose a computational chain based on the lattice code (SCALE6/GenPMAXS), the coupled neutronic/thermal hydraulic code (TRACE/PARCS) as well as a Monte Carlo based uncertainty and sensitivity package (SUSA) has been established and applied to different kind of transients of a Boiling Water Reactor (BWR). This paper will describe the multidimensional models of the plant elaborated for TRACE and PARCS to perform the investigations mentioned before. For the uncertainty quantification of the coupled code TRACE/PARCS and specifically to take into account the influence of the kinetics parameters in such studies, the PARCS code has been extended to facilitate the change of model parameters in such a way that the SUSA package can be used in connection with TRACE/PARCS for the U and S studies. This approach will be presented in detail. The results obtained for a rod drop transient with TRACE/PARCS using the SUSA-methodology showed clearly the importance of some kinetic parameters on the transient progression demonstrating that the coupling of a best-estimate coupled codes with uncertainty and sensitivity tools is very promising and of great importance for the safety assessment of nuclear reactors. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

A mono-dimensional nuclear fuel performance analysis code, PUMA, development from a coupled approach

Multidimensional-multi-physical phenomena in nuclear fuels are treated as a set of mono-dimensional-coupled problems which encompass heat, displacement, fuel constituent redistribution, and fission gas release. Rather than uncoupling these coupled equations as in conventional fuel performance analysis codes, efforts are put into to obtain fully coupled solutions by relying on the recent advances of numerical analysis. Through this approach, a new SFR metal fuel performance analysis code, called PUMA (Performance of Uranium Metal fuel rod Analysis code) is under development. Although coupling between temperature and fuel constituent was made easily, the coupling between the mechanical equilibrium equation and a set of stiff kinetics equations for fission gas release is accomplished by introducing one-level Newton scheme through backward differentiation formula. Displacement equations from 1D finite element formulation of the mechanical equilibrium equation are solved simultaneously with stress equation, creep equation, swelling equation, and FGR equations. Calculations was made successfully such that the swelling and the hydrostatic pressure are interrelated each other. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

A fission matrix based validation protocol for computed power distributions in the advanced test reactor

The Idaho National Laboratory (INL) has been engaged in a significant multi year effort to modernize the computational reactor physics tools and validation procedures used to support operations of the Advanced Test Reactor (ATR) and its companion critical facility (ATRC). Several new protocols for validation of computed neutron flux distributions and spectra as well as for validation of computed fission power distributions, based on new experiments and well-recognized least-squares statistical analysis techniques, have been under development. In the case of power distributions, estimates of the a priori ATR-specific fuel element-to-element fission power correlation and covariance matrices are required for validation analysis. A practical method for generating these matrices using the element-to-element fission matrix is presented, along with a high-order scheme for estimating the underlying fission matrix itself. The proposed methodology is illustrated using the MCNP5 neutron transport code for the required neutronics calculations. The general approach is readily adaptable for implementation using any multidimensional stochastic or deterministic transport code that offers the required level of spatial, angular, and energy resolution in the computed solution for the neutron flux and fission source. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Statistical modeling support for calibration of a multiphysics model of subcooled boiling flows

Nuclear reactor system analyses rely on multiple complex models which describe the physics of reactor neutronics, thermal hydraulics, structural mechanics, coolant physico-chemistry, etc. Such coupled multiphysics models require extensive calibration and validation before they can be used in practical system safety study and/or design/technology optimization. This paper presents an application of statistical modeling and Bayesian inference in calibrating an example multiphysics model of subcooled boiling flows which is widely used in reactor thermal hydraulic analysis. The presence of complex coupling of physics in such a model together with the large number of model inputs, parameters and multidimensional outputs poses significant challenge to the model calibration method. However, the method proposed in this work is shown to be able to overcome these difficulties while allowing data (observation) uncertainty and model inadequacy to be taken into consideration. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

A Multi-Dimensional Benefit Assessment of Automated Mobility Platforms (AMP) for Large Facilities

Automated Mobility Platform (AMP) is a general term for intelligently managed passenger mobility systems that leverage automation and electrification to improve system performance and sustainability through the integration of real-time data streams with advanced decision-making processes tailored for use in large facilities. Using airports as an example, the rapid growth in air travel demand has led to facility expansions and congested terminals, which directly impacts equity (e.g., increased challenges for passengers with reduced mobility [PRMs]) and sustainability - both of which are important metrics often neglected during the engineering design process. Therefore, to evaluate systems and inform critical decisions more effectively, a holistic evaluation framework is proposed that includes equity, sustainability, and infrastructure perspectives - referred to as a multidimensional benefit assessment. More specifically, the following analysis focuses on: (1) mobility, with emphasis on travel time and accessibility within an airport, contrasting existing systems with the proposed AMP system; (2) equity analysis, specifically to the PRM community, but with an eye to the whole of society; (3) energy consumption and greenhouse gas (GHG) emissions associated with intra-airport mobility, comparing and contrasting modes as well as impacts on overall airport operations efficiency; and (4) fundamental changes in building design enabled by AMPs for larger and more flexible, functional, and energy-efficient structures.

ADVANCED PROPULSION SYSTEMS↗
DOE OSTI2023

Use of latent metathesis polymerization systems for photopolymerization-based additive manufacturing

Photolatent olefin metathesis catalysts can be used in combination with metathesis-active monomers or resins for photopolymerization-based additive manufacturing via ring-opening metathesis polymerization (ROMP). A variety of alternative catalysts and photosensitizers are commercially available that allows for tuning of thermomechanical properties, potlifes, activation rates, and irradiation wavelengths. As an example, two optimized catalyst/photosensitizer systems were demonstrated in the rapid stereolithographic printing of complex, multidimensional poly-dicyclopentadiene structures with microscale features under ambient conditions.

Leguizamon, Samuel↗
DOE OSTI2016

An Atomic Mix Closure for Stochastic Media Transport Problems

The accurate modeling of stochastic media in radiation transport calculations remains on ongoing research challenge. The Levermore-Pomraning (LP) closure for the statistical transport equation is simple and inexpensive, but it is often inaccurate. More accurate closures have been proposed, but they can be prohibitively expensive. In this work we present a relatively inexpensive closure that we have observed to be more accurate than LP in almost all cases studied. Obtaining solutions to stochastic transport problems can be quite difficult. Transport calculations on an ensemble of explicit realizations generated by Monte Carlo sampling can be prohibitively expensive. Atomic mix or LP calculations are relatively inexpensive, but they can be inaccurate. The present work illustrates an approach that is less expensive than Monte Carlo sampling yet more accurate than atomic mix or LP approximations. We still need to analyze the reasons why the atomic mix closures derived for thicker problems lead to source iteration instabilities. It is not clear if the closure itself induces an ill-posed problem or if it is merely the iterative process that is problematic. We hope to examine alternative source terms to drive the subsidiary calculations. We also want to extend the work to multigroup and multidimensional problems. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

On the Modeling of Wall Heat Flux Partitioning in Subcooled Flow Boiling

Accurate prediction of void distribution in reactor coolant channels at subcooled boiling conditions is an important factor in the analysis of the operating conditions of both pressurized water reactors (PWRs) and boiling water reactors (BWRs). Its importance arises from purely thermal-hydraulic considerations (thermal limits, CRUD formation, etc.) and the impact of void reactivity feedback on core neutronics. A mechanistic multidimensional model of subcooled boiling was originally developed by Kurul and Podowski. Whereas this model has been implemented in several commercial CFD codes and successfully used in computer simulations of flow and phase distributions in the coolant channels of light water reactor, additional new modeling aspects of forced-convection subcooled boiling heat transfer have been investigated in recent studies. The purpose of this paper is to formulate a modified model of wall heat flux partitioning, which allows for capturing and quantifying the effect of simultaneous evaporation and condensation at the heated wall on the net evaporation rate and wall temperature. The new model has been implemented in the NPHASE-CMFD code, numerically tested and validated against experimental data. A mechanistic wall heat flux partitioning model for subcooled flow boiling has been presented. This model has been combined with extensively tested models of interfacial mass, momentum and energy transfer terms. It turns out that the model predictions of the wall temperature and void fraction are in good agreement with the experimental data of reference. Future work will focus on the continuation of model validation against experimental data, and on extending its scope to include the phenomena governing departure from nucleate boiling. (authors)

42 ENGINEERING↗
DOE OSTI2016

Modeling of Two-Phase Flow around Spacers with Mixing Vanes in Reactor Rod-Bundles

Reactor fuel rod bundles experience complex two-phase flow patterns that are governed by multiple interacting phenomena. Proper understand of the underlying flow hydrodynamics can point to potential hotspots within nuclear reactors. The structural supports, called spacer grids, have mixing vanes fixed to the trailing edges to enhance mixing and swirl within the flow. This is an added phenomenon that makes the flows even more complicated to model. As reactors can experience boiling conditions, it is important to understand how the spacer geometry affects local velocity and phase distributions in these geometries. This work represents a continuation of a study that has been focused on developing, testing, and validation of a consistent multidimensional model of dispersed bubbly flow along PWR coolant channels,. The proposed model has been implemented in the NPHASE-CMFD code the results of computer simulations have been compared to experimental results of Yloenen. One of the objectives of the current study has been to examine the effects of selected recent modeling improvements at RPI in NPHASE-CMFD as a part of the continuing model development and validation efforts. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Studying the hadron structure with PANDA and CLAS using machine learning techniques

The hadron spectroscopy and structure are currently very active fields of research to study the non-perturbative regime of quantum chronodynamics. The first one studies the complex structure of excited hadrons by looking at their decay products, while the latter uses lepton scattering on nucleons. Both methods require reconstruction algorithms with great efficiency and good particle identification and background rejection rates. This work aims to provide these by either improving the existing methods or developing new ones. The first part of this document presents a feasibility study of a predicted hybrid charmonium state for the PANDA experiment. Lattice QCD calculations predict the ground state hybrid charmonium to be a spin exotic with quantum numbers of JP C = 1?+ at a mass of around 4.3 GeV with a width to be around 20 MeV. A machine learning based data analysis scheme is proposed to further improve the signal efficiency and the background reduction, alongside with improvements of the analysis software (PandaRoot), that are vital for this study. These improvements include a reworked clustering algorithm for the electromagnetic calorimeter (EMC) and an optimized monte carlo matching for neutral particles. The second part of this document is about studying the proton structure. A multidimensional study of the structure function ratio Fsin(?)LU /FUU has been performed for K±, based on the measurement of beam-spin asymmetries. It uses the high statistics data recorded with the CLAS12 spectrometer at Jefferson Laboratory. Fsin(?)LU is a twist-3 quantity that provides information about the quark gluon correlations in the proton. This document will present for the first time a simultaneous analysis of two kaon channels over a large kinematic range of z, xB , PT and Q2 with virtualities Q2 ranging from 1 GeV2 up to 8 GeV2 using machine learning techniques for improved particle identification.

Kripko, Aron↗
DOE OSTI2022

Pronghorn Porous Media Model Validation with Pressure Drop Measurements

The verification and validation (V&V) of Pronghorn is imperative to assert its accuracy when predicting the fluid velocity, temperature, and pressure in high temperature gas-cooled reactors. Pronghorn is a coarse-mesh, intermediate-fidelity, and multidimensional thermal-hydraulics (TH) code developed by the Idaho National Laboratory (INL). New pebble bed experiments are used to observe the details of the fluid motion and pressure drop in the porous bed under the reactor normal operation. This paper focuses on the validation of the Pronghorn compressible and incompressible Navier-Stokes equations using the pressure drop measurements performed at the engineering-scale pebble bed facility at the Texas A&M university (TAMU). Various pressure drop correlations and porosity functions are implemented in both Pronghorn and STAR-CCM+ to compare the pressure drop due to the combined viscous and inertial resistances in the porous bed. The correlations accounting for the near-wall effect are also utilized to observe if the pressure drop estimates can be improved. Pronghorn porous media models predict the pressure drop well relative to the STAR-CCM+ simulation results and 1D correlations, and both the finite element method (FEM) and finite volume method (FVM) perform accurately. Pronghorn models are also validated with the experimental measurements given the different Reynolds number ranges and specific aspect ratios. The likelihood of the statistical significance between the pressure drop measurements and specific correlations or simulations is low provided that the overlap of their confidence intervals is more than the half of a single arm. Several validation metrics are reasonable in regard to the similar studies from other literature. The precise average pebble bed porosity estimation has much impact on the pressure drop, and the Foumeny and Montillet (dense packing) models carry out the accurate pressure drop prediction by considering the near-wall effect.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

IMPLEMENTATION OF TURBULENCE MODELING FOR THE COUPLED GRIFFIN-PRONGHORN SIMULATION OF THE MOLTEN SALT FAST REACTOR FOR THE VIRTUAL TEST BED

In support of the Virtual Test Bed (VTB) repository, the Molten Salt Fast Reactor (MSFR) concept is modeled using MOOSE-based tools, notably Griffin and Pronghorn, employing newly implemented initial turbulence models and standard wall functions. The models are adapted to Pronghorn’s spirit of a multidimensional coarse-mesh intermediate fidelity code. Multiphysics simulations of a 2D axisymmetric model of an MSFR are carried out to study the coupling between the different physics.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Advancing Thermal-Hydraulic Modeling Capabilities for Molten Salt Reactors in Pronghorn

This is an M2 report for the NEAMS program. This work focuses on advancing the modeling and simulation capabilities for Molten Salt Reactors (MSRs) within Pronghorn, addressing key thermal-hydraulic challenges that must be addressed for the safe and efficient operation of these advanced reactors. This report summarizes the development and validation of the following new models in Pronghorn: (i) two-equation RANS turbulence models, (ii) near-wall turbulence corrections due to curvature and roughness, and (iii) multidimensional two-phase drift-flux models. For each of these cases, along with the experimental validation, an application study of interest to MSRs is presented. In addition, this report also presents recent improvements in the domain-overlapping coupling between Pronghorn and SAM to address the cases of weakly compressible flow and flow reversal. Finally, an MSR action that greatly simplifies the setup of MSR models using NEAMS tools is described.

42 ENGINEERING↗
DOE OSTI2024

Generating An Advanced Cross-section Library For HTGR Pebble Bed Depletion Calculations Using Reduced-Order Model Generation Techniques

For code development, Advanced Reactor Technologies - Gas Cooled Reactors Program (ART-GCR) rely on a collaboration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program, but the cross sections generation and the methodology definition is part of this program area goals. Based on previous studies in FY23, the size of microscopic cross section libraries increases rapidly with the number of tabulations, requiring significant amount of memory and drastically slowing down the Griffin calculations when evaluating cross sections via the multivariate linear interpolation approach. Rising to these challenges, this work investigates constructing Reduced-order Models (ROMs) for the multi-group microscopic cross sections to accelerate the cross section evaluation in Griffin. A database of multigroup cross sections is first collected considering all possible parameters that a designer could change for optimization. Down-selection of the ROM techniques afterward shows Deep Neural Network (DNN) as the best candidate when jointly consider memory efficiency, predictive accuracy, computational cost, scalability, flexibility and ease of implementation of the algorithms in comparison to the multidimensional interpolation. This work develops a specific interface that enables the cross section predictions using pre-trained DNN models into Griffin leveraging the existing ROM capabilities. DNNs have been trained for all isotopes for use in Griffin. Preliminary Griffin testing shows that DNNs exhibit exceptional predictive accuracy and the use of DNNs provides orders of magnitude improvement in memory efficiency compared to conventional interpolation techniques. With such ROM techniques, it holds great promise to further increase the fidelity of the Pebble Bed Reactor (PBR) simulation by increasing the number of tabulations/state variables during cross section evaluation, while maintaining the computational cost affordable in Griffin.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

WIP: Collaborating to Introduce Second-Life Battery Technology Research in an Informal STEM Learning Environment

This work in progress, innovative practice, paper outlines an interdisciplinary collaborative model utilized by a team of educational researchers, and engineers to design a series of community-centered, place-based learning activities suitable for a middle or high school informal STEM learning environment. Based on the potential applications and benefits of Second-Life Electric Vehicle Batteries, the activities connect cutting-edge STEM research with multidimensional real-world challenges within their communities related to climate change, renewable energy integration and storage, intelligent transportation systems, and optimized energy management, that empowers students to think critically about the role they can play in shaping a more resilient, environmentally sustainable community.

Whiteside, Hope↗