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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 1,207 records · Page 67
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

NRC Multiphysics Analysis Capability Deployment (FY2021 - Part 2)

This report details the progress and activities of Idaho National Laboratory (INL) on the Nuclear Regulatory Commission (NRC) project “Development and Modeling Support for Advanced Non-Light Water Reactors.” Task 4b was completed for this report. INL developed a sample problem showing how to use Serpent 2 to calculate macroscopic cross sections for use in Griffin for a typical sodium fast reactor Unprotected Loss of Flow (ULOF) transient. The complicating factor is the relatively large axial gradient of the coolant (and hence fuel) temperature and its evolution during the transient. A 3D Griffin model of the Advanced Burner Test Reactor (ABTR) is coupled to the System Analysis Module (SAM) to perform the ULOF transient. The model includes various explicit feedback mechanisms for fast reactors, including Doppler, radial expansion from the displacement of the support plate, and axial expansion from the displacement of the fuel pins. The results for the ULOF transient are consistent with published values. We observed small effects in the transient results that arise from the superhomogenization equivalence correction of uniform and nonuniform temperature data sets, but they diminish as more dominant reactivity mechanisms are added to the model. Potential improvements to the neutronics model include: adding the differential control rod and sodium feedback mechanisms, implementing cross section re-homogenization based on the partial volume of the various materials, and enhancing the kinetics parameters. Furthermore, the fluids model only includes four subchannels to represent the entire active core due to performance limitations in SAM when attempting to model one subchannel per assembly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Effect of double heterogeneity treatment on neutronics modeling of HTGR unit cell - 18524

The High Temperature Gas-cooled Reactor (HTGR) utilizes unique coated fuel particle design, which is embedded within a graphite matrix and shaped into spherical or hexagonal fuel element. The feature of the inherent double heterogeneity in the HTGR fuel element therefore raises difficulties in modeling and requires special simulation strategies to address its impact on the self-shielding calculation. In this study, neutronics calculations with various levels of heterogeneity processing were performed for both spherical and hexagonal HTGR fuel elements. The results indicate that ignoring either level of double heterogeneity can lead to significant underestimation of self-shielding effect and consequently yield as much as 8% deviation in the prediction of the multiplication factor. Fission and capture rates were compared for the spherical HTGR fuel model and over 30% deviation was found in the resonance energy range if homogenized model used. Further investigation on the neutron spectrum demonstrated similar tendency for both types of fuel elements when applying the simplified self-shielding models, that is, {approx}20% underestimation of neutron flux in 6.5-6.75 eV energy interval was observed due to the incomplete evaluation of the cross section resonance self-shielding effect. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transient thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2019

Numerical Modeling of Supersonic Combustion in a Non- Premixed Rotating Detonation Engine

A computational fluid dynamics (CFD) model was developed to simulate a representative non-premixed RDE combustor. Unsteady Reynolds-Averaged Navier-Stokes (RANS) simulations were performed for the full combustor geometry (including the separate fuel and air injection ports), with hydrogen as the fuel and air as the oxidizer. Adaptive mesh refinement (AMR) was incorporated to achieve a good trade-off between model accuracy and computational expense. A finite-rate chemistry model was employed along with a 10-species detailed kinetic mechanism to describe the H2-Air combustion chemistry. Two different operating conditions were simulated, corresponding to the same global equivalence ratio of unity but different fuel and air mass flow rates. Numerical results were compared against available experimental data on detonation wave frequency/height, reactant fill height, oblique shock angle and axial pressure distribution in the channel. The CFD model predicted the sensitivity of these wave characteristics to operating conditions with good accuracy, both qualitatively and quantitatively. The present CFD model offers a potential capability to perform rapid design space exploration and/or performance optimization studies for realistic full-scale RDE configurations.

Pal, Pinaki↗
NASA NTRSDate not supplied

Validation of Transient Spacecraft Refueling Model with Gateway Breadboard Test Data

The ability to refuel spacecraft on-orbit is a key enabling technology for deep space human exploration. On-orbit refueling requires priming a pressurized liquid propellant from a source tank into fluid transfer lines at vacuum conditions, which could result in excessive surge pressures beyond the system’s qualified operating limits and potentially damage hardware. Therefore, it is imperative to employ numerical models during the development of flight hardware to parametrically characterize system performance over a wide range of operating conditions. Since priming transients are difficult to model accurately, it is important to validate those models with test data. As part of the Gateway bipropellant refueling system development, a simplified fluidic breadboard system was created using water as a simulant for fuel to support early program risk reduction of refueling operations. The objectives of breadboard testing were to gather performance data to characterize and demonstrate critical refueling operations and to validate numerical models that can be extended to predicting flight system performance. A description of the breadboard test system and the results of numerical model validation is presented. The model is shown to have very good agreement with test data.

Brian S Lusby↗
DOE OSTI2019

Resolution Limits of Near IR and IR Thermography Applied to Imaging Thermal Heterogeneities in Nuclear Fuel Surrogate Materials

ABSTRACT The main goal of the project is: To develop an IR thermographic method to monitor fuel pellet cracking inside the hostile environment of the nuclear core. This method will use a high temperature and radiation resistant fiber optic bundle to monitor the fuel pellet inside the core and transmit the information outside to an IR camera. This method will give a visual cue (using IR video) as to when the crack started to form and how it progresses. To achieve this goal, we identified the following as major milestones of the project: 1. Perform 1D modeling of temperature and blackbody radiation intensity across a crack. 2. Image a crack on a fuel cell surrogate using an IR camera with free-space optics and high power laser setup. 3. Simulate cracking in the surrogate sample through inductive heating method, and demonstrate imaging crack formation using the free-space optics setup. 4. Determine parameters for a suitable fiber optic bundle compatible with the IR camera based on the results from the free-space optics experiments. 5. Demonstrate imaging using fiber bundle and integrate viewing system with nuclear core at INL. This report focuses on Goals 1-3. We have successfully achieved Goals 1-3 and are making steady progress towards achieving Goals 4-9. Goals 2 and 4 are very similar and require 2 basic parts: the imaging system and the method to crack the fuel surrogate to replicate the fuel pellet cracking in the nuclear reactor core. The following sections describe the 1D modeling of temperature and blackbody radiation intensity across a crack (Goal 1), procedure to create crack on fuel surrogates and results (Goals 2 and 3), and the IR imaging system and results (Goal 2).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Project Startup: Evaluating the Performance of Frito Lay's Electric Delivery Trucks (Fact Sheet)

The Fleet Test and Evaluation Team at the National Renewable Energy Laboratory (NREL) is evaluating the in-service performance of 10 medium-duty Smith Newton electric vehicles (EVs) and 10 comparable conventional diesel vehicles operated by Frito Lay North America in the Seattle, Washington, area. Launched in late 2013, the on-road portion of this 12-month evaluation focuses on collecting and analyzing vehicle performance data, such as fuel economy and maintenance costs, to better understand how to optimize the use of such vehicles in a large-scale commercial operation. In addition to the on-road portion of this evaluation, NREL is analyzing charging data to support total cost of ownership estimations and investigations into smart charging opportunities. NREL is also performing a battery life degradation analysis to quantify battery pack health, track battery performance over time, and determine how various drive cycles and battery charging protocols impact battery life.

25 ENERGY STORAGE↗
DOE OSTI2023

Numerical investigation and experimental validation of water condensation in the gas diffusion layer with different properties

Liquid water in porous media hinders the transport of reactant to catalyst layer, where the electrochemical reactions occur, which in turn, reduces performance and lifetime of a proton exchange membrane fuel cell (PEMFC). In this study, two types of gas diffusion layer (GDL) materials are used to study the two-phase saturation distribution in a PEMFC. The correlations between the effective and anisotropic transport properties and porosity of the two GDLs are solved by the pore scale model. In addition, the anisotropic liquid water permeability and the relationship between saturation pressure and capillary pressure is determined by using the Lattice-Boltzmann method. Fuel cell performance as well as liquid water distribution in the GDL using neutron radiography are obtained for validation. Under wet conditions, the cell with Freudenberg GDL performs better than that using Toray GDL, especially at high current densities. The results of the two-phase model simulation show that the peak water saturation for Toray GDL occurs in the catalyst layer and can reach 40%, while the peak water saturation for Freudenberg GDL occurs inside the GDL directly under the ribs and only reaches 25% saturation. The combined results provide key insights to enable high power density operation of a PEMFC through GDL material optimization.

42 ENGINEERING↗
DOE OSTI2019

Status Report on Development of MARCH Module Designs

The Transient Reactor Test (TREAT) facility was constructed in the late 1950’s, provided thousands of transient irradiations before being placed in standby in 1994, and was restarted in 2017 in order to resume its crucial role in nuclear heated safety research. Advances in modern computational capabilities and a resurgence of interest in novel reactor technology have created an opportunity for emphasizing modern science-based and separate effects test capabilities at TREAT. An innovative approach to this type of testing leverages the relatively low radioisotope accumulation during brief TREAT irradiations by arranging small fresh fuel specimens in low activation hardware “modules” so that they can be easily extracted and shipped for examination within weeks. The concept was termed the Minimal Activation Retrievable Capsule Holder (MARCH) irradiation vehicle system, and its maturation began under a Lab Directed Research and Development (LDRD) project starting in fiscal year 2017. Within the first year of this project, enough design and predictive modeling had been performed to prove the principle and ascertain support from major direct-funded DOE fuel development programs including the Accident Tolerant Fuels (ATF) program that sponsored the first modern fueled test to be performed in TREAT using the MARCH system. Since the design was modular in nature, it was straightforward for other sources to adopt and mature viable test modules while the LDRD project continued concurrently for two more years in conceptualizing and demonstrating the viability of other innovative modules. Many of these modules were adopted for continued development and irradiation under the Advanced Fuels Campaign and other direct sources. This report summarizes the MARCH modules that were conceptualized, designed, and deployed to date with the primary aim of providing reference information to future experiment designers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

BISON Capabilities for LWR Fuel Behavior Analysis During Accident and High-burnup Conditions

The U.S. Department of Energy (DOE)'s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program aims to develop predictive capabilities using computational methods for the analysis and design of advanced reactor and fuel cycle systems. This program has been supporting the development of BISON which is a high-fidelity and high-resolution fuel performance tool at the engineering scale. Increasing recent interest in applications at extended burnups motivated this study to incorporate more physically based models in BISON. This document details integration of newly implemented modeling capabilities into BISON, which includes (1) new thermal conductivity models that are valid up to 100 GWd/t, (2) models for the formation of the high-burnup structure (HBS), (3) two porosity correction methods beingapplied on the thermal conductivity due to the conducting pores during the HBS formation. BISON's results are verified and validated to test the new modeling capabilities

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Refinements of a treat model for transient analysis

In order to prepare for the TREAT reactor being taken off standby, the Gateway for Advanced Modeling and Simulation Lab has used a set of KENO models to simulate certain TREAT temperature- limited transients using the Transient Reactor excursion simulator (T-ReX) code, a hybrid deterministic-stochastic code. These transients are from the M8CAL experiment series, a set of experiments done to calibrate the reactor for the planned M8 series of fuel tests. Recently, the model has been updated in media definitions and geometry. Models were examined and calculations were performed to low uncertainty values to find the k{sub eff} and reactivity insertions. Alterations to modeled values of shutdown and initial transient rod insertion depth were made to create models which had experimental values of reactivity insertion, pre-transient k{sub eff}, or both. It was found that two of the altered media inputs, fuel and Zircaloy-3 cladding, had significant effect on the k{sub eff}. Surprisingly, the Zircaloy-3 composition had a larger effect on k{sub eff}. The reactivity insertions in this model's 2855, 2856 and 2857 temperature-limited transients were found to be above experimental values by a factor of 5.5% to 7%, and that pre-transient k{sub eff} was too high, at 1.002. It was found that increasing shutdown rod insertion by 2 to 2.5 cm decreased pre-transient k{sub eff} to 1.0, and that increasing shutdown rod insertion by 3 to 5 cm and decreasing initial transient rod insertion by 1 to 2 cm gave perfect pre-transient k{sub eff} and total reactivity insertion. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Transient uncertainty quantification and Global Sensitivity Analysis of the open-source Molten Chloride Reactor Experiment (MCRE) using GP-PCA surrogate models

Uncertainties in the thermophysical properties of molten salts impact both the steady-state and transient behavior of Molten Salt Reactors (MSRs). In this work, we aim to quantify the influence of such uncertainties on the transient operation of the Molten Chloride Reactor Experiment (MCRE), utilizing the open-source specifications provided for this reactor. Seven representative transient scenarios are considered. For each scenario, we evaluate the impact of thermophysical property uncertainties on four key multiphysics model output variables of interest (VoIs): maximum power density, maximum fuel temperature, maximum reflector temperature, and average fuel velocity magnitude. In addition, we perform a Global Sensitivity Analysis (GSA) by computing Sobol’ indices for the uncertain input parameters to determine their contribution to the variability of each VoI. Conducting GSA is computationally intensive due to the large number of required evaluations of the high-fidelity multiphysics model. To mitigate this cost, we develop a surrogate modeling framework that combines Gaussian Process (GP) regression with Principal Component Analysis (PCA), enabling efficient sample generation for the GSA. Our results show that for energy-related VoIs, thermal conductivity is the dominant contributor to uncertainty. In contrast, for flow-related VoIs, density and dynamic viscosity are the primary sources of uncertainty. The specific heat of the fuel salt was found to play a secondary role in the transient analyses.

42 - ENGINEERING↗
DOE OSTI2019

MODELING SHOCK AND VIBRATION ON USED NUCLEAR FUEL DURING NORMAL CONDITIONS OF TRANSPORTATION

In 2017, the United States Department of Energy (DOE) collaborated with Spanish and Korean organizations to perform a multimodal transportation test to measure shock and vibration loads imparted to used nuclear fuel (UNF) assemblies. This test used real fuel assembly components containing surrogate fuel mass to approximate the response characteristics of real, irradiated used nuclear fuel. Pacific Northwest National Laboratory was part of the test team and used the data collected during this test to validate numerical models needed to predict the response of real used nuclear fuel in other transportation configurations. This paper summarizes the modeling work and identifies lessons learned related to the modeling and analysis methodology. The modeling includes railcar dynamics using the NUCARS software code and explicit dynamic finite element modeling of used nuclear fuel cladding in LS-DYNA. The NUCARS models were validated against railcar dynamics data collected during captive track testing at the Federal Railroad Administration's Transportation Technology Center in Pueblo, CO. The LS-DYNA models of the fuel cladding were validated against strain gage data collected throughout the test campaign. One of the key results of this work was an assessment of fuel cladding fatigue, and the methods used to calculate fatigue are detailed in this paper. The validated models and analysis methodologies described in this paper will be applied to evaluate future UNF transportation systems.

Klymyshyn, Nicholas A.↗
DOE OSTI2016

Identification of fuel cycle simulator functionalities for analysis of transition to a new fuel cycle

Dynamic fuel cycle simulation tools are intended to model holistic transient nuclear fuel cycle scenarios. As with all simulation tools, fuel cycle simulators require verification through unit tests, benchmark cases, and integral tests. Model validation is a vital aspect as well. Although compara-tive studies have been performed, there is no comprehensive unit test and benchmark library for fuel cycle simulator tools. The objective of this paper is to identify the must test functionalities of a fuel cycle simulator tool within the context of specific problems of interest to the Fuel Cycle Options Campaign within the U.S. Department of Energy s Office of Nuclear Energy. The approach in this paper identifies the features needed to cover the range of promising fuel cycle options identified in the DOE-NE Fuel Cycle Evaluation and Screening (E&S) and categorizes these features to facilitate prioritization. Features were categorized as essential functions, integrating features, and exemplary capabilities. One objective of this paper is to propose a library of unit tests applicable to each of the essential functions. Another underlying motivation for this paper is to encourage an international dialog on the functionalities and standard test methods for fuel cycle simulator tools.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Influence of Silane Coupling Agents on Filler Network Structure and Stress-Induced Particle Rearrangement in Elastomer Nanocomposites

Filled rubber materials are key in many technologies having broad impact on the economy and sustainability, the most obvious being tire technology. Adding filler dramatically improves the strength of rubber by reinforcement and tailoring the type of filler and the chemistry of the interface between filler and rubber matrix is important for optimizing performance metrics such as fuel efficiency. In a highly loaded, silica-filled, crosslinked model rubber closely mimicking commercial materials, both the filler network structure and the dynamics of the silica filler particles change when the silica surface is modified with silane coupling agents. Reduction in size scales characteristic of the structure are quantified using ultra-small-angle X-ray scattering (USAXS) measurements and the particle dynamics probed with X-ray photon correlation spectroscopy (XPCS). While the structure averaged over the scattering volume changes little with aging after step strain, the dynamics slow appreciably in a manner that varies with the treatment of the silica filler. The evolution of filler particle dynamics depends on the chemical functionality at the silica surface and observing these differences suggests a way of thinking about the origins of hysteresis in nanoparticle-reinforced rubbers. These microscopic filler dynamics are correlated with the macroscopic stress relaxation of the filled materials. The combination of static and dynamic X-ray scattering techniques with rheological measurements is a powerful approach for elucidating the microscopic mechanisms of rubber reinforcement.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Experimentally Known Properties of U-10Zr Alloys: A Critical Review

More than 14000 U-10Zr fuel rods (alloys of U with 10 wt% Zr) Zr were been irradiated in the Experimental Breeder Reactor II and Fast Flux Test Facility, with some irradiated to burnups of almost 20 at%. However, very little information about properties of un-irradiated U-10Zr alloys is available. This data is needed to help with interpretation of data from irradiated materials, develop and validate phase diagrams and models of U-Zr and more complex systems, develop and validate fuel-performance codes, design fuels for future fast reactors, and design future experimental investigations to fill in crucial gaps in knowledge. This paper provides a summary and critical review of the available experimental data on phases and phase diagrams, electrical properties, thermal expansion, thermal conductivity, heat capacity, mechanical properties, vapor pressures, and thermodynamic properties of un-irradiated U-10Zr alloys. As a result, much of this data was published before ~1970 and is available only in obscure reports.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Tansmutation Research program

Six years of research was conducted for the United States Department of Energy, Office of Nuclear Energy between the years of 2006 through 2011 at the University of Nevada, Las Vegas (UNLV). The results of this research are detailed in the narratives for tasks 1-45. The work performed spanned the range of experimental and modeling efforts. Radiochemistry (separations, waste separation, nuclear fuel, remote sensing, and waste forms) , material fabrication, material characterization, corrosion studies, nuclear criticality, sensors, and modeling comprise the major topics of study during these six years.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗