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At least 721 records · Page 40

Assessment of Heavy-Duty Fueling Methods and Components

Chevron, NLR, ANL, and NextEnergy partnered in the development of a comprehensive assessment of heavy-duty (HD) fuel cell electric vehicle fueling protocols. The project leveraged and built upon existing international heavy-duty (HD) fueling protocols and fueling component development activities to deliver component performance assessments, modeling tools and methods evaluations, techno-economic assessments of industry-selected protocol structures and experimental validations of the strategies performed at NLR's HD hydrogen fueling station.

08 HYDROGEN↗

Simulations of Evaporating Multicomponent Fuel Drops

A paper presents additional information on the subject matter of Model of Mixing Layer With Multicomponent Evaporating Drops (NPO-30505), NASA Tech Briefs, Vol. 28, No. 3 (March 2004), page 55. To recapitulate: A mathematical model of a three-dimensional mixing layer laden with evaporating fuel drops composed of many chemical species has been derived. The model is used to perform direct numerical simulations in continuing studies directed toward understanding the behaviors of sprays of liquid petroleum fuels in furnaces, industrial combustors, and engines. The model includes governing equations formulated in an Eulerian and a Lagrangian reference frame for the gas and drops, respectively, and incorporates a concept of continuous thermodynamics, according to which the chemical composition of a fuel is described by use of a distribution function. In this investigation, the distribution function depends solely on the species molar weight. The present paper reiterates the description of the model and discusses further in-depth analysis of the previous results as well as results of additional numerical simulations assessing the effect of the mass loading. The paper reiterates the conclusions reported in the cited previous article, and states some new conclusions. Some new conclusions are: 1. The slower evaporation and the evaporation/ condensation process for multicomponent-fuel drops resulted in a reduced drop-size polydispersity compared to their single-component counterpart. 2. The inhomogeneity in the spatial distribution of the species in the layer increases with the initial mass loading. 3. As evaporation becomes faster, the assumed invariant form of the molecular- weight distribution during evaporation becomes inaccurate.

Bellan, Josette↗

Fuel Cell Model Validation with Power Module Test Data

The AES Modular Power Systems (AMPS) Fuel Cell team at the NASA Glenn Research Center (GRC) has created a Microsoft Excel model of a regenerative fuel cell (RFC) system for potential lunar and Martian applications. Due to some limitations of this model, including the ability to predict transient effects in the thermal and fluidic performance of the system, a system-level transient thermal/fluids model was desired. GT-SUITE is a transient multi-physics simulation tool that is popular in the automotive industry and has a built-in template for proton exchange membrane (PEM) fuel cells. However, terrestrial fuel cells used in the automotive industry vary significantly from aerospace fuel cells. A validation effort was undertaken to compare GT SUITE model results to actual data collected during the 2015 AMPS (Advanced Exploration Systems (AES) Modular Power Systems) Power Module Demonstration in which the AMPS team used a PEM fuel cell system to power a 478 kilogram rover assembly at the Dunes test site located at GRC. The purpose of the model was to evaluate the efficacy of using GT SUITE to model a non-flow through PEM fuel cell system. The validation effort examined electrochemical performance including fuel cell stack voltage and power supplied, the thermal performance of the stack namely stack temperature and heat rejection rates, and the fluidic performance of the system including pressures, reactant consumption, and water production rates.

Model Validation↗

Energy Economics of Farm Biogas in Cold Climates

Anaerobic digestion of farm and dairy waste has been shown to be capital intensive. One way to improve digester economics is to co-digest high-energy substrates together with the dairy manure. Cheese whey for example represents a high-energy substrate that is generated during cheese manufacture. There are currently no quantitative tools available that predict performance of co-digestion farm systems. The goal of this project was to develop a mathematical tool that would (1) predict the impact of co-digestion and (2) determine the best use of the generated biogas for a cheese manufacturing plant. Two models were developed that separately could be used to meet both goals of the project. Given current pricing structures of the most economical use of the generated biogas at the cheese manufacturing plant was as a replacement of fuel oil to generate heat. The developed digester model accurately predicted the performance of 26 farm digesters operating in the North Eastern U.S.

09 BIOMASS FUELS↗

Particle Image Velocimetry Measurements in a Wire-Wrapped 61-Pin Hexagonal Fuel Bundle

Liquid metal fast reactor fuel assemblies using sodium as a coolant typically utilize a tightly packed triangular lattice of fuel pins enclosed in a hexagonal duct. The fuel pins are typically wrapped with a helical wire spacer. The primary function of the wire spacers is to maintain a gap between neighboring fuel pins. They also mitigate vortex-induced vibration and enhance sub-channel mixing to increase convective heat transfer. Many experiments have been performed to study thermal-hydraulic characteristics, primarily pressure drops, of the wire-wrapped fuel bundles. Novendstern developed a semi-empirical model to predict pressure losses in a hexagonal array of wire-wrapped pins in the turbulent flow regime. Rehme proposed pressure drop correlations for a wide range of geometrical parameters, such as pitch-to-diameter ratio P/D from 1.125 to 1.417, lead-to-wire diameter ratio H / dm from 6 to 45, and number of pins n from 7 to 61. Cheng and Todreas introduced their hydrodynamic models for subchannel friction factors and mixing parameters in a 37-pin fuel bundle, which covered the laminar, transition and turbulent flow regimes. Computational fluid dynamics (CFD) offers an opportunity to validate the above correlations and investigate the flow characteristics in the subchannels. Several numerical studies of the wire-wrapped fuel bundles have been performed and most of them are Reynolds-Averaged Navier-Stokes (RANS) based simulations with various turbulence models. Ahmad and Kim performed RANS simulations of 7- and 19-pin bundles using the k-omega SST turbulent model, while Gajapathy et al. performed RANS simulations on 7-pin bundles using the k-epsilons model. Recently, Fischer et al. carried out large-eddy simulation (LES) calculations on wire-wrapped fuel bundles to obtain higher-resolution numerical solutions of subchannel behavior. The geometrical complexity of the hexagonal wire-wrapped fuel bundle has limited experimental activities to pressure drop measurements. A database of pressure drops and flow-field measurements in a 61-pin wire-wrapped hexagonal fuel bundle is not available to benchmark the existing correlations and validate the CFD calculations. Texas A and M University is conducting isothermal flow experiments in a wire-wrapped 61-pin hexagonal fuel bundle to support the research on advanced nuclear fuel development sponsored by the Department of Energy (DOE). The purpose of these tests is to perform high spatial and temporal resolution measurements of the flow and pressure measurements at different locations in the fuel bundle. The experimental facility employs matched-index-of-refraction (MIR) techniques and laser diagnostic velocity measurement techniques such as particle image velocimetry (PIV) and laser Doppler velocimetry (LDV). The result of this experimental activity will be an experimental database of pressure and flow-field measurements that will be suitable for validating CFD computer codes. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Developmental Objectives for Advanced Reactor Fuels

For any technology, the pace of development is determined in part by the time and expense required to explore alternatives to the current state of the art. For example, the capabilities of software are growing exponentially in large part because the costs of experimentation are so low (e.g., a programmer can often try out a new idea in a matter of minutes). For nuclear fuels and materials, the time and cost of iterating on alternatives is extremely high, which in turn helps to explain why nuclear reactor development can be excruciatingly slow. The pace of development, however, can be accelerated in two distinct ways: (1) improvements in the productivity of our irradiation testing and post-irradiation examinations infrastructure; (2) the use of a variety of alternatives to radiological work, including physically-based models, extensive literature reviews, increased reliance on out-of-reactor testing, and evaluation of the impact of fuel system developments on reactor plant performance. Viewed in this context, a fuel development program effort consists not only of test pin fabrication, irradiation testing, and post-irradiation examinations, but also on physically-based models, out-of-reactor testing, benchmarking of models against historic fuel performance data, and literature reviews. Of course, elements of these efforts are already in place, but increased emphasis on these activities are expected to speed the fuel development effort. Quite simply, the approach calls for obtaining as much knowledge and understanding as possible using whatever means necessary while minimizing the amount of radiological work that is required. In this manner, once an irradiation test program is designed and initiated, there is not only an improved likelihood that the fuel system will behave as expected but there is also a better chance, if the testing is successful, that the fuel design will result in substantial improvements to reactor plant economics and/or safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Flow visualization and computational studies of a reverse flow circular combustor

The development of a reverse-flow circular combustor is described in terms of the visualization techniques that are employed to optimize the flow and mixing of fuel and air. The Fluent computer program is used in conjunction with flow visualization studies to investigate the performance of the combustor. The combustor has multiple mini-combustion zones (MCZs) and fuel injection nozzles along its circular geometry, and the fuel injection design is modeled to determine temperature distribution, including the performance at the combustor exit. The yellow dye smoke used in the visualization demonstrated desirable flow patterns with rapid circulation patterns within the MCZs and extensive interaction between the channel flow and the MCZs. The results are similar to those of CFD models, and the main conclusions are that fuel atomization can be further optimized and that the fuel injector nozzle can control swirls, fuel spray angle, and jet penetration.

Ghorashi, B.↗

FEASIBILITY ASSESSMENT FOR DEVELOPING AN INTEGRAL LOCA TESTING CAPABILITY AT THE TRANSIENT RESEARCH TEST (TREAT) REACTOR

The study of Light Water Reactor (LWR) fuel performance in Loss of Coolant Accidents (LOCAs) has been a significant and evolving field since the beginning of the commercialization of LWR technology. The following paper presents the results of feasibility assessment to develop an in-pile, integral LOCA testing capability using the recently restarted TREAT reactor located at the Idaho National Laboratory (INL). Critical performance objectives for the testing platform are derived from a review of known LWR fuel degradation phenomena in LOCA scenarios. Advanced modeling and simulation tools were then used to gain additional insight into both the system performance and fuel performance characteristics of LOCA testing at TREAT. Results show that TREAT is well suited to conduct in-pile integral LOCA testing.

42 ENGINEERING↗

A thermal conductivity model for U-­Si compounds

U 3 Si 2 is a candidate for accident tolerant nuclear fuel being developed as an alternative to UO 2 in commercial light water reactors (LWRs). One of its main benefits compared to UO 2 is higher thermal conductivity that increases with temperature. This increase is contrary to UO 2 , for which the thermal conductivity decreases with temperature. The reason for the difference is the electronic origin of thermal conductivity in U 3 Si 2 , as compared to the phonon mechanism responsible for thermal transport in UO 2 . The phonon thermal conductivity in UO 2 is unusually low for a fluorite oxide due to the strong interaction with the spins in the paramagnetic phase. The thermal conductivity of U 3 Si 2 as well as other U-­Si compounds has been measured experimentally [1-­4]. However, for fuel performance simulations it is also critical to model the degradation of the thermal conductivity due to damage and microstructure evolution caused by the reactor environment (irradiation and high temperature). For UO 2 this reduction is substantial and it has been the topic of extensive NEAMS research resulting in several publications [5, 6]. There are no data or models for the evolution of the U 3 Si 2 thermal conductivity under irradiation. We know that the intrinsic thermal conductivities of UO 2 (semi-conductor) and U 3 Si 2 (metal) are very different, and we do not necessarily expect the dependence on damage to be the same either, which could present another advantage for the silicide fuel. In this report we summarize the first step in developing a model for the thermal conductivity of U-­Si compounds with the goal of capturing the effect of damage in U 3 Si 2 . Next year, we will focus on lattice damage. We will also attempt to assess the impact of fission gas bubbles.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Assembly Level Modeling and Transportation Damage Prediction of Used Nuclear Fuel (UNF) Cladding

Under U.S. Nuclear Regulatory Commission (NRC) regulations, UNF must maintain its integrity in such a way that it can withstand the physical forces of handling and transportation associated with restaging the fuel and moving it to a different location such as an interim storage site, a geologic repository, or a treatment/recycling facility. Hence, understanding mechanical performance under cumulative loading stemming from normal conditions of storage (NCS), transfer from storage container to transport container if needed, and normal conditions of transport (NCT) is necessary as it establishes part of the safety basis via maintaining the fuel confining boundary (geometry), maintaining criticality safety, and is one of the critical components to the preservation of retrievability. An initial demonstration of modeling and simulation capabilities was performed as part of a collaborative effort between Pacific Northwest National Laboratory (PNNL), Sandia National Laboratory (SNL), Idaho National Laboratory (INL), Oak Ridge National Laboratory (ORNL), and the Transportation Technology Center Inc. (TTCI). This demonstration completed preliminary deterministic evaluations of moderate-to-high burnup UNF mechanical performance under NCS and NCT conditions. This paper focuses on work performed by PNNL on the assembly level modeling of a UNF assembly and fuel cladding damage predictions under NCT loads. The Finite Element (FE) model created for a 17 x 17 Pressurized Water Reactor (PWR) optimized fuel assembly (OFA) is discussed along with the communications to and from the fuel rod-level and cask-level models. Additionally, the methodology for determining cumulative damage is discussed including the 3000 mile damage prediction from a 10 second simulation. The damage results are shown for five shock and vibration broadened loading histories. It is found that predicted cladding strains were not large enough to cause structural failure, but cyclic strains roughly projected for the entire route were significant in some cases. The total damage from summation of the worst shock and vibration cases evaluated is approximately 18% of the expected fatigue limit. Therefore it is predicted that the fuel rods do not fail during NCT given the assumptions used throughout this initial demonstration.

used nuclear fuel, UNF, trasportation, damage, fue↗

Physics-Based Model to Represent Membrane-Electrode Assemblies of Solid-Oxide Fuel Cells Based on Gadolinium-Doped Ceria

This paper reports a physics-based model that predicts membrane-electrode assembly (MEA) performance of solid-oxide fuel cells (SOFCs) with Ce 0.9 Gd 0.1 O 2− δ (GDC10) electrolyte membranes. The paper derives self-consistent thermodynamic and transport properties for GDC1o mobile charged defects (oxide vacancies and reduced-ceria small polarons) by fitting published measurements of oxygen non-stoichiometry and conductivity over ranges of temperature and O 2 partial pressures. The button-cell model is applied to evaluate how mixed ionic-electronic conductivity influences the performance of an SOFC MEA with a GDC10 electrolyte sandwiched between a porous, composite Ni-GDC10 anode and a porous, composite cathode of Sm 0.5 Sr 0.5 CoO 3− δ (i.e., SSC) and GDC10. SSC properties are also derived by fitting published conductivity and oxygen non-stoichiometry measurements. Mixed conductivity of GDC10 and competing charge transfer reactions at both electrodes reduce open circuit voltages due to leakage current and buildup of defect concentrations at electrode-electrolyte interfaces. To fit polarization data, the button-cell model includes heterogeneous reaction rates for defect incorporation on the GDC10 surface along with Butler–Volmer expressions derived for competing charge transfer reaction rates from rigorous analyses assuming rate-limiting, elementary charge transfer reactions for each electrode. The calibrated MEA model can support rigorous SOFC modeling with GDC10 electrolytes over the range of conditions within a fully operating cell.

Electrochemistry↗

Triso Analysis Tool For Predictive Source Terms

Source term modeling for TRi-structural ISOtropic (TRISO) fuel has been performed for previous reactor designs, but few are available in the open literature. Thus, there is a need to develop a simple, versatile, and mechanistic model of fission product release and transport in gas reactor cores that can be applied to a variety of reactors through user inputs and reactor-specific radionuclide inventories. To meet this need, the TRISO Analysis Tool for Predictive Source terms (TRISO-ATOPS) was developed. This model calculates the release of the key safety-important fission products by diffusion through the kernel, silicon carbide and graphite based on fuel and graphite temperatures in the reactor under normal operation. These releases from the fuel enter the coolant where they can plate-out on cooler surfaces. A clean-up model is included for designs with a coolant purification system for removing fission gases. This initial distribution of fission products in the reactor serves as an initial condition for potential releases under postulated accident conditions. From this initial condition, the model calculates the fission product release for any transient temperature profile, and the fission product releases can then be used to assess radiological dose to the workers and the public using conventional radiological dose tools. Data on the diffusion of fission products is based on historic German TRISO experiments and the more current Department of Energy Advanced Gas Reactor TRISO fuel development program. The example cases in this work demonstrate the flexibility of the model

Stoyer, Benjamin [Idaho National Laboratory (INL),↗

Thermal-Hydraulic Results for the Aboveground Configuration of a Dry Cask Simulator

The thermal performance of a commercial nuclear spent fuel dry storage cask is evaluated through detailed analytical modeling. These modeling efforts are performed by the vendor to demonstrate the performance as well as regulatory compliance and are independently verified by the Nuclear Regulatory Commission (NRC). The majority of commercial dry storage casks in use today are aboveground. Cooling of the assemblies located inside the sealed canister is achieved by naturally-convective air flowing in the bottom of the cask and exiting out the top. Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating design and performance models. Numerous studies have been previously conducted. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and belowground configurations. While several testing programs have been previously conducted, these earlier validation efforts did not capture the effects of elevated helium pressures or accurately portray the external convection of aboveground and belowground canistered dry cask systems. Thus the enhanced performance of modern dry storage casks cannot be fully validated using previous studies.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Sensitivity analysis of VERA-CS and FRAPCON coupling in a multiphysics environment

A demonstration and description of the LOCA Toolkit for US light water reactors (LOTUS) is presented. Through LOTUS, the core simulator VERA-CS developed by CASL is coupled with the fuel performance code FRAPCON. The coupling is performed with consistent uncertainty propagation with all model inconsistencies being well-documented. Monte Carlo sampling is performed on a single 17 × 17 fuel assembly with a three cycle depletion case. Both uncertainty quantification (UQ) and sensitivity analysis (SA) are used at multiple states within the simulation to elucidate the behavior of minimum departure from nucleate boiling ratio (MDNBR), maximum fuel centerline temperature (MFCT), and gap conductance at peak power (GCPP). The SA metrics used are the Pearson correlation coefficient, Sobol sensitivity indices, and the density-based, delta moment independent measures. Results for MDNBR show consistency among all SA measures, as well for all states throughout the fuel lifecycle. MFCT results contain consistent rankings between SA measures, but show differences throughout the lifecycle. GCPP exhibits predominantly linear relations at low and high burnup, but highly nonlinear relations at intermediate burnup due to abrupt shifts between models. As a result, such behavior is largely undetectable to traditional regression or variance-based methods and demonstrates the utility of density-based methods.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Test Plan for the Boiling Water Reactor Dry Cask Simulator

The thermal performance of commercial nuclear spent fuel dry storage casks are evaluated through detailed numerical analysis . These modeling efforts are completed by the vendor to demonstrate performance and regulatory compliance. The calculations are then independently verified by the Nuclear Regulatory Commission (NRC). Carefully measured data sets generated from testing of full sized casks or smaller cask analogs are widely recognized as vital for validating these models. Recent advances in dry storage cask designs have significantly increased the maximum thermal load allowed in a cask in part by increasing the efficiency of internal conduction pathways and by increasing the internal convection through greater canister helium pressure. These same vertical, canistered cask systems rely on ventilation between the canister and the overpack to convect heat away from the canister to the environment for both above and below-ground configurations. While several testing programs have been previously conducted, these earlier validation attempts did not capture the effects of elevated helium pressures or accurately portray the external convection of above-ground and below-ground canistered dry cask systems. The purpose of the investigation described in this report is to produce a data set that can be used to test the validity of the assumptions associated with the calculations presently used to determine steady-state cladding temperatures in modern vertical, canistered dry cask systems. The BWR cask simulator (BCS) has been designed in detail for both the above-ground and below-ground venting configurations. The pressure vessel representing the canister has been designed, fabricated, and pressure tested for a maximum allowable pressure (MAWP) rating of 24 bar at 400 deg C. An existing electrically heated but otherwise prototypic BWR Incoloy-clad test assembly is being deployed inside of a representative storage basket and cylindrical pressure vessel that represents the canister. The symmetric single assembly geometry with well-controlled boundary conditions simplifies interpretation of results. Various configurations of outer concentric ducting will be used to mimic conditions for above and below-ground storage configurations of vertical, dry cask systems with canisters. Radial and axial temperature profiles will be measured for a wide range of decay power and helium cask pressures. Of particular interest is the evaluation of the effect of increased helium pressure on allowable heat load and the effect of simulated wind on a simplified below ground vent configuration. While incorporating the best available information, this test plan is subject to changes due to improved understanding from modeling or from as-built deviations to designs. As-built conditions and actual procedures will be documented in the final test report.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

Understanding Model Inadequacy in TRISO Nuclear Fuel Fission Products Release Models: Empirical and Mechanistic Approaches

The increasing use of tristructural isotropic (TRISO) particle fuel in both advanced and existing reactors necessitates a thorough evaluation of uncertainties and shortcomings in TRISO fission product release models. These inadequacies arise from the simplifications made in computational models compared to experimental data. Utilizing the BISON fuel performance code and experimental data from the Advanced Gas Reactor (AGR) program provides a unique chance to rigorously assess these inadequacies within a Bayesian uncertainty quantification (UQ) framework. This study contrasts the standard Bayesian framework with the Kennedy-O'Hagan (KOH) framework, which explicitly accounts for modeling inadequacies, in the context of UQ for TRISO silver release models. It examines both the traditional Arrhenius equation and a more advanced lower-length-scale (LLS)-informed model that incorporates microstructure information. The inverse UQ process applied to AGR-2 and AGR-3/4 datasets identified modeling inadequacy as the primary source of uncertainty, with experimental noise also being significant, while model parameter uncertainty was minimal. Both the Arrhenius and LLS-informed models showed similar levels of modeling inadequacy. For forward predictive UQ using the AGR-1 dataset, the KOH framework enhanced the accuracy and quality of quantified uncertainties by approximately 30% and 40%, respectively, compared to the standard Bayesian framework. This improvement was observed for both the Arrhenius and LLS-informed models. At the engineering scale, both models performed similarly, but the LLS-informed model outperformed the Arrhenius equation at the mesoscale. These findings underscore the importance of explicitly considering modeling inadequacy in the UQ process and highlight the need for ongoing refinement of physics-based models to address these shortcomings.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Understanding Water Transport in Polymer Electrolyte Fuel Cells Using Coupled Continuum and Pore-Network Models

Water management remains a critical issue for polymer electrolyte fuel cell performance and durability, especially at lower temperatures and with ultrathin electrodes. To understand and explain experimental observations better, water transport in gas diffusion layers (GDLs) with macroscopically heterogeneous morphologies was simulated using a novel coupling of continuum and pore-network models. X-ray computed tomography was used to extract GDL material parameters for use in the pore-network model. The simulations were conducted to explain experimental observations associated with stacking of anode GDLs, where stacking of the anode GDLs increased the limiting current density. Through imaging, it is shown that the stacked anode GDL exhibited an interfacial region of high porosity. The coupled model shows that this morphology allowed more efficient water movement through the anode and higher temperatures at the cathode compared to the single GDL case. As a result, the cathode exhibited less flooding and hence better low temperature performance with the stacked anode GDL.

30 DIRECT ENERGY CONVERSION↗

Coupled Bison/Thermochimica Simulation of Hydride Formation in Zircalloy Cladding

The report describes implementation of the thermodynamics models for simulation of zirconium hydride formation in LWR fuel cladding. The current models for ZrH x formation are based on empirical rate models that must calibrated to materials and regimes of interest. The thermodynamic approach enables hydride formation calculations based on fundamental thermodynamics principles and material composition. Thermodynamic formulations and models were implemented in the Idaho National Laboratory fuel performance code Bison using Material, Kernel and AuxKernel methods. The results for the Material and Kernel implementation are shown in this report. The Material method implementation naturally fits the concept of modeling evolution of material state in the material models. Given the relatively high cost of thermodynamic computation, the advantage of the AuxKernel compared to Material implementation is the ability to control the execution and computational cost. The implemented formulations are illustrated on example problems which fit the experimental data and satisfy the computational requirements of conservation equations.

36 MATERIALS SCIENCE↗