Heat Transfer Challenges for MVDC Power Cables Used in Wide Body All Electric Aircraft Under Low Pressures
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42 ENGINEERING↗
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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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Conjugate Heat Transfer Analysis of the Supercritical CO2 Based Counter Flow Compact 3D Heat Exchangers
Compact heat exchangers using supercritical fluids such as CO2 are preferred due to their high heat transfer capacity and smaller footprint. Three-dimensional conjugate forced convection heat transfer analysis was performed on several shell-and-tube counter-flow microchannel heat exchangers. Numerical simulations were conducted to test effect of change in mass flow rate, hydraulic diameter and various cross sections on the heat transfer. Increasing mass flow rate improved heat transfer up to a maximum value and then decreased downstream with increasing turbulence. Maximum heat transfer was obtained for the micro channel with the smallest hydraulic diameter. Amongst the cross sections analyzed (circular, square, circular with radial ribs, and square with radial ribs), the most uniform distribution of temperature and maximum heat transfer were obtained for circular cross section with radial ribs. An optimally efficient operation of such a heat exchanger can be attained by considering these factors during multi-objective constrained optimization of geometric parameters and requirements for additive manufacturing of such compact heat exchangers.
Chitale, Janhavi↗
Recent Progress in Developing Nb 3 Sn Conductors With Increased Specific Heat
A possible approach to reducing Nb 3 Sn magnet training is to increase the energy margin of Nb 3 Sn conductors by enhancing their specific heat (C p ). For this study, we have been developing Nb 3 Sn conductors with increased C p by incorporating substances with high C p at 2-10 K based on a conductor design that is compatible with standard Nb 3 Sn strand production. In the past couple of years our efforts have been mainly focused on improving strand design (e.g., position of high- C p filaments, thickness of the Cu tube for the high-C p filaments, ratio of the Cu powder to the high-C p substance, filament spacing, etc.) in order to obtain good strand drawability and to reduce degradation after rolling, which is needed for production of Rutherford cables. We also tried a new high-C p substance, Gd 2 O 2 S, and verified that it has much higher C p over the whole magnetic field range than the Gd 2 O 3 we used before. With that development work we can now produce high-C p strands with good drawability and low levels of degradation after rolling. This paper reports our findings and the current status of the development of high-C p Nb 3 Sn conductors.
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
Effect of material properties on batch‐to‐glass conversion kinetics
Abstract A recently developed model of the cold cap—the reacting glass batch (melter feeds) floating on molten glass in an electric glass melter—couples heat transfer with the feed‐to‐glass conversion kinetics. The model allows for determining the distributions of temperature and various properties within the cold cap. In the present study, this model is applied to four melter feeds designed for high‐level and low‐activity nuclear wastes. Profiles of temperature, conversion degree, cold cap porosity and density, condensed matter velocity, and heating rate were determined using the material properties of the cold cap. Effects of vigorous foaming at the cold cap bottom were considered. Density, thermal conductivity, and glass production rate strongly affect the cold cap thickness and the fraction of undissolved silica entering the melt under the cold cap. The heating rate profile in the cold cap is highly nonlinear, with high heating rates observed in the foam layer.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Construction of a Code Verification Matrix for Heat Conduction With Finite Element Code Applications
When establishing the pedigree of a simulation tool, code verification is used to ensure that the implemented numerical algorithm is a faithful representation of its underlying mathematical model. During this process, numerical results on various meshes are systematically compared to a reference analytic solution. The selection of analytic solutions can be a laborious process, as it is difficult to establish adequate code confidence without performing redundant work. In this report we address this issue by applying a physics-based process that establishes a set of reference problems. In this process, code simulation options are categorized and systematically tested, which ensures that gaps in testing are easily identified and addressed. The resulting problems are primarily intended for code verification analysis but may also be useful for comparison to other simulation codes, troubleshooting activities, or training exercises. The process is used to select fifteen code verification problems relevant for the one-dimensional steady-state heat conduction equation. These problems are applicable to a wide variety of simulation tools, but, in this work, a demonstration is performed using the finite element-based nuclear fuel performance code BISON. Convergence to the analytic solution at the theoretical rate is quantified for a selection of the problems, which establishes a baseline pedigree for the code. Not only can this standard set of conduction solutions be used for verification of other codes, but also the physics-based process for selecting problems can be utilized to quantify and expand testing for any simulation tool.
42 ENGINEERING↗
Transformational challenge reactor design characteristics
The Transformational Challenge Reactor (TCR) program was conceived with the goal to reduce costs and time frames associated with advanced reactor deployment by leveraging developments in advanced manufacturing, advanced materials, data science, and rapid prototyping and testing. The final deliverable of the TCR program was to be an operational test of a novel reactor design. The TCR core design incorporates a dense tri-structural-isotropic/SiC fuel form and volumetrically efficient yttrium hydride moderator, both of which were manufactured and characterized under the TCR program. The TCR is a 3 MW{sub th} He-cooled experimental nuclear reactor designed to reach a total integrated burnup of less than 24 effective full-power hours to keep the radioactive source term to a very low level. TCR design process revealed a positive moderator coefficient; however, the negative doppler coefficients for the fuel and thermal expansion of fuel, moderator, and core support plate yield an overall negative reactivity coefficient. Calculated fuel element temperatures and stresses are well within safety margins. The maximum hypothetical accident (i.e., de-pressurized loss of forced cooling) yields only a modest increase in reactor temperatures that are all within safety margins. This paper summarizes the high-level TCR design characteristics, which were derived from neutronics, thermohydraulics, thermomechanics, and safety analyses.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Brady's Geothermal Field DAS and DTS Surface and Borehole Array Metadata
This metadata submission includes the coordinates of the DAS and DTS surface and borehole arrays, the list of file names, and the list of recorded files during testing at the PoroTomo Natural Laboratory at Brady Hot Spring in Nevada. Testing was completed during March 2016.
15 GEOTHERMAL ENERGY↗
Brady's Geothermal Field - Metadata for DTS and DAS Surveys
Metadata for DTS and DAS datasets for both borehole 56-1 and trenched cables.
15 GEOTHERMAL ENERGY↗
Utah FORGE: High-Resolution DAS Microseismic Data from Well 78-32
This regards a high-resolution DAS microseismic dataset produced by Silixa from Utah FORGE Phase 2C seismic monitoring well 78-32 during stimulation testing of well 58-32. It is a very large dataset and as such it is currently not directly available on GDR. However, it is available for download from the Center for High Performance Computing (CHPC) at the University of Utah using the shell script below. Additional survey information and tips for running the script are available in the attached downloadable Word Document file.
15 GEOTHERMAL ENERGY↗
Temperature Profile of the Solution Vessel of an Accelerator-Driven Subcritical Fissile Solution System
Dynamic System Simulation (DSS) models of fissile solution systems have been developed and verified against a variety of historical configurations. DSS techniques have been applied specifically to subcritical accelerator-driven systems using fissile solution fuels of uranium. Initial DSS models were developed in DESIRE, a specialized simulation scripting language. In order to tailor the DSS models to specifically meet needs of system designers they were converted to a Visual Studio implementation, and one of these subsequently to National Instrument’s LabVIEW for human factors engineering and operator training. Specific operational characteristics of subcritical accelerator-driven systems have been examined using a DSS model tailored to this particular class using fissile fuel.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
Argonne Bubble Experiment Thermal Model Development
This report will describe the Computational Fluid Dynamics (CFD) model that was developed to calculate the temperatures and gas volume fractions in the solution vessel during the irradiation. It is based on the model used to calculate temperatures and volume fractions in an annular vessel containing an aqueous solution of uranium . The experiment was repeated at several electron beam power levels, but the CFD analysis was performed only for the 12 kW irradiation, because this experiment came the closest to reaching a steady-state condition. The aim of the study is to compare results of the calculation with experimental measurements to determine the validity of the CFD model.
38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
Conjugate Heat Transfer and Thermal Mechanical Analysis for the Fast Spectrum Neutron Source for Materials Irradiation
A conjugate heat transfer and thermal structural analysis was completed, with the objective of determining the following: Lead bismuth eutectic (LBE) peak temperature, free convective velocity patterns in the LBE, peak beam window temperature, and thermal stress/deformation in the window.
43 PARTICLE ACCELERATORS↗
Conjugate Heat Transfer and Thermal Mechanical Analysis for Liquid Metal Targets for High Power Electron Beams.
A conjugate heat transfer and thermal structural analysis was completed, with the objective of determining the following: Lead bismuth eutectic (LBE) peak temperature, free convective velocity patterns in the LBE, peak beam window temperature, and thermal stress/deformation in the window.
43 PARTICLE ACCELERATORS↗
Uncertainty Quantification of Calculated Temperatures for the AGR 3/4 Experiment
A series of Advanced Gas Reactor (AGR) irradiation experiments are being conducted within the Advanced Reactor Technology (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaign are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and accident conditions, and support development and validation of fuel performance and fission product transport models and codes (PLN 3636, “Technical Program Plan for INL Advanced Reactor Technologies Technology Development Office/Advanced Gas Reactor Fuel Development and Qualification Program”). The AGR 3/4 test was inserted in the Northeast Flux Trap position in the Advanced Test Reactor (ATR) core at Idaho National Laboratory (INL) in December 2011 and successfully completed irradiation in mid-April 2014, resulting in irradiation of the tristructural isotropic (TRISO) fuel for 369.1 effective full-power days (EFPDs) during approximately 2.4 calendar years. The AGR 3/4 data, including the irradiation data and calculated results, were qualified and stored in the Nuclear Data Management and Analysis System (NDMAS). To support the U.S. TRISO fuel performance assessment and to provide data for validation of fuel performance and fission product transport models and codes, the daily as run thermal analysis has been performed separately on each of twelve AGR 3/4 capsules for the entire irradiation as discussed in ECAR-2807, “AGR 3/4 Daily As Run Thermal Analyses”. The ABAQUS code’s finite element-based thermal model predicts the daily average volume average (VA) fuel temperature (FT), peak FT, and graphite matrix, sleeve, and sink temperature in each capsule. The JMOCUP simulation codes were also created to perform depletion calculations for the AGR 3/4 experiment (ECAR-2753, “JMOCUP As-Run Daily Physics Depletion Calculation for the AGR 3/4 TRISO Particle Experiment in ATR Northeast Flux Trap”). This depletion analysis provides fast fluence and fission heat rate data for all components (fuel compacts, graphite rings, stainless steel retainer, etc.), which are used as inputs for the thermal analysis codes. The graphite temperatures from thermocouples (TCs) in each capsule were used to calibrate these thermal analysis codes. However, given the high rate of TC failure under the harsh irradiation and thermal conditions in the AGR capsules, the thermal analysis results are very useful in aiding TC data qualification, increasing the confidence in delineating failures of the measuring instruments (TCs) from physical mechanisms that may have shifted the system thermal response.
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Year End Progress Report on Rattlesnake Improvements
Rattlesnake is a MOOSE-based radiation transport application developed at INL to support modern multi-physics simulations. At the beginning of the last year, Rattlesnake was able to perform steady-state, transient and eigenvalue calculations for the multigroup radiation transport equations. Various discretization schemes, including continuous finite element method (FEM) with discrete ordinates method (SN) and spherical harmonics expansion method (PN) for the self-adjoint angular flux (SAAF) formulation, continuous FEM (CFEM) with SN for the least square (LS) formulation, diffusion approximation with CFEM and discontinuous FEM (DFEM), have been implemented. A separate toolkit, YAKXS, for multigroup cross section management was developed to support Rattlesnake calculations with feedback both from changes in the field variables, such as fuel temperature, coolant density, and etc., and in isotope inventory. The framework for doing nonlinear diffusion acceleration (NDA) within Rattlesnake has been set up, and both NDA calculations with SAAF-SN-CFEM scheme and Monte Carlo with OpenMC have been performed. It was also used for coupling BISON and RELAP-7 for the full-core multiphysics simulations. Within the last fiscal year, significant improvements have been made in Rattlesnake. Rattlesnake development was migrated into our internal GITLAB development environment at the end of year 2014. Since then total 369 merge requests has been accepted into Rattlesnake. It is noted that the MOOSE framework that Rattlesnake is based on is under continuous developments. Improvements made in MOOSE can improve the Rattlesnake. It is acknowledged that MOOSE developers spent efforts on patching Rattlesnake for the improvements made on the framework side. This report will not cover the code restructuring for better readability and modularity and documentation improvements, which we have spent tremendous effort on. It only details some of improvements in the following sections.
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
COMSOL Multiphysics Model for HLW Canister Filling
The U.S. Department of Energy (DOE) is building a Tank Waste Treatment and Immobilization Plant (WTP) at the Hanford Site in Washington to remediate 55 million gallons of radioactive waste that is being temporarily stored in 177 underground tanks. Efforts are being made to increase the loading of Hanford tank wastes in glass while meeting melter lifetime expectancies and process, regulatory, and product quality requirements. Wastes containing high concentrations of Al 2 O 3 and Na 2 O can contribute to nepheline (generally NaAlSiO 4 ) crystallization, which can sharply reduce the chemical durability of high level waste (HLW) glass. Nepheline crystallization can occur during slow cooling of the glass within the stainless steel canister. The purpose of this work was to develop a model that can be used to predict temperatures of the glass in a WTP HLW canister during filling and cooling. The intent of the model is to support scoping work in the laboratory. It is not intended to provide precise predictions of temperature profiles, but rather to provide a simplified representation of glass cooling profiles within a full scale, WTP HLW canister under various glass pouring rates. These data will be used to support laboratory studies for an improved understanding of the mechanisms of nepheline crystallization. The model was created using COMSOL Multiphysics, a commercially available software. The model results were compared to available experimental data, TRR-PLT-080, and were found to yield sufficient results for the scoping nature of the study. The simulated temperatures were within 60 ºC for the centerline, 0.0762m (3 inch) from centerline, and 0.2286m (9 inch) from centerline thermocouples once the thermocouples were covered with glass. The temperature difference between the experimental and simulated values reduced to 40 ºC, 4 hours after the thermocouple was covered, and down to 20 ºC, 6 hours after the thermocouple was covered. This level of precision is considered acceptable for the scoping nature of the model and the subsequent laboratory glass studies Using the model, two additional glass pouring cycles were conducted. Representative thermocouple data were plotted to show the variations between the two cycles. This provides preliminary data that will be used in laboratory experiments to determine the potential for controlling nepheline crystallization in glass by varying the glass pouring conditions.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Brine Transport Experiments in Granular Salt
To gain confidence in the predictive capability of numerical models, experimental validation must be performed to ensure that parameters and processes are correctly simulated. The laboratory investigations presented herein aim to address knowledge gaps for heat-generating nuclear waste (HGNW) disposal in bedded salt that remain after examination of prior field and laboratory test data. Primarily, we are interested in better constraining the thermal, hydrological, and physicochemical behavior of brine, water vapor, and salt when moist salt is heated. The target of this work is to use run-of-mine (RoM) salt; however during FY2015 progress was made using high-purity, granular sodium chloride.
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
Study of the internal structure, instabilities, and magnetic fields in the dense Z-pinch
Z-pinches are sources of hot dense plasma which generates powerful x-ray bursts and can been applied to various areas of high-energy-density physics (HEDP). The 26-MA Z machine is at the forefront of many of these applications, but important aspects of HEDP have been studied on generators at the 1 MA current level. Recent development of laser diagnostics and upgrade of the Leopard laser at Nevada Terawatt Facility (NTF) give new opportunities for the dense Z-pinch study. The goal of this project is the investigation of the internal structure of the stagnated Z pinch including sub-mm and micron-scale instabilities, plasma dynamics, magnetic fields, and hot spots formation and initiation. New plasma diagnostics will be developed for this project. A 3D structure and instabilities of the pinch will be compared with 3D MHD and spectroscopic modeling and theoretical analysis. The structure and dynamics of stagnated Z pinches has been studied with x-ray self-radiation diagnostics which derive a temperature map of the pinch with a spatial resolution of 70-150 µm. The regular laser diagnostics at 532 nm does not penetrate in the dense pinch due to strong absorption and refraction in trailing plasma. Recent experiments at NTF showed that shadowgraphy at the UV wavelength of 266 nm unfolds a fine structure of the stagnated Z-pinch with unprecedented detail. We propose to develop laser UV diagnostics for Z pinches with a spatial resolution <5 μm to study the small-scale plasma structures, implement two-frame shadowgraphy/interferometry, and develop methods for investigation of strong magnetic fields. New diagnostics will help to understand better basic physical processes in Z pinches. A 3D internal structure of the pinch and characteristic instabilities will be studied in wire arrays with different configurations and compared with 3D MHD simulations and analytical models. Mechanisms of “enhanced heating” of Z-pinch plasma will be studied. Fast dynamics of stagnated plasma will be studied to estimate its contribution to the Doppler broadening of x-ray lines. Development of “necks” and “hot spots” will be studied with high-resolution UV diagnostics, an x-ray streak camera, and x-ray spectroscopy. Laser initiation of hot spots in Z pinches will be tested. A Faraday rotation diagnostic at 266 nm will be applied to 1-10 MG magnetic fields. For magnetic fields B>20 MG, suggested in micropinches, Cotton-Mouton and cutoff diagnostics will be applied. A picosecond optical Kerr shutter will be tested to increase a sensitivity of UV methods for application at multi-MA Z pinches. The proposal is based on the experimental capability of NTF. The Zebra generator produces 1-1.7 MA Z-pinches with electron plasma density of 10 20 -10 21 cm -3 , electron temperature of 0.5-1 keV, and magnetic fields >10 MG. The Leopard laser was upgraded to energy of 90-J at 0.8 ns. This regime will be used for laser initiation of hot spots. A further upgrade to energy of 250-J is suggested for laser-Z-pinch interaction. A picosecond regime will be used for optical gating. A 10-TW Tomcat laser at NTF is available for the high energy UV laser probing of the Z-pinch. Two graduate students will develop new optical and x-ray diagnostics, carry out experiments, and process experimental data. Other students will be involved in the design and fabrication of loads, supporting regular optical and x-ray diagnostics, and data processing. The new plasma diagnostics may be applied to HEDP experiments at NTF and other multi-MA generators. The feasibility of the research plan is based on the experience of the scientific team in Z-pinch plasma physics, laser physics, development of new plasma diagnostics, and the experimental capability of NTF. The experimental group of Dr. V. V. Ivanov (UNR) collaborates with a group for Z pinch MHD modeling of Dr. J. P. Chittenden (Imperial College, London), and theoretical group of Dr. D. D. Ryutov (LLNL). The suggested research ideas are supported by preliminary experiments.
70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗