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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 199 records · Page 11

Lower length scale informed improvements to Bison U-Pu-Zr fuel swelling model

Due to renewed interest in metallic fuels from the U-(Pu)-Zr material system, significant improvements have recently been made to BISON's capability to simulate metallic fuel, particularly in the area of swelling. In this report, lower length scale simulations that have been conducted to inform the engineering-scale models in BISON during FY20 are described. For the high-temperature regions of the fuel that are dominated by the $\gamma$ phase of U-(Pu)-Zr, two new models for swelling have implemented in BISON. Phase-field simulations were conducted to calculate parameters for these and other BISON models that control when gaseous swelling ceases and fission gas release begins. For the low-temperature regions of the fuel dominated by the phase of uranium, microstructurally resolved simulations of pore growth and crystal deformation were performed to understand the effect of irradiation-generated crystal shape changes. Conformally meshed microstructures with pores were simulated with a crystal plasticity model and with a burnup eigenstrain model. Plastic deformation will not cause pore size to increase; burnup eigenstrain (irradiation-induced crystal shape changes) will increase the volume of existing porosity and cause it to become anisotropic in shape. Additional model development work for early-stage fission gas cluster/bubble formation in the $\gamma$ phase were performed, by coupling with defect clustering within cluster dynamics framework. Sensitivity analysis were conducted to identify the critical thermo-kinetic parameters needed for rigorous assessment.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Water vapor oxidation of SiC layer in surrogate TRISO fuel particles

Under accidental conditions for high temperature gas-cooled reactors (HTGR), the SiC layer in tri-structural-isotropic (TRISO) fuel particles can be exposed to water vapor. In this study, oxidation behaviors of surrogate TRISO fuel particles were investigated in a He-20 vol% water vapor mixed atmosphere at temperatures up to 1600 °C. The growth of the crystalline oxide passivation layer with cracks and pores followed a parabolic law with time, where the maximum was 2.3 μm at 1600 °C. The oxide layer and the SiC surface under the oxide became flattened with increasing temperature, as a function of the silica viscosity and the diffusion path of water vapor. Volatilization of the oxide layer was analyzed using a mechanistic model that an inert gas in oxidizing atmospheres could influence the magnitude of volatilization. The fracture load and strength of the oxidized and thinned SiC layer were numerically estimated to decrease from 2.27 to 1.69 N and from 317 to 299 MPa, respectively, with the SiC thickness decrease from 35 to 32 μm. This prediction indicates that the oxidized SiC layer should retain fission products. Additionally, the mechanical integrity of each layer in the TRISO fuel particle after oxidation was evaluated. The results in this work provide important data for the safety analysis of accidental scenarios in HTGRs.

36 MATERIALS SCIENCE↗

A quantitative phase-field model for gas bubble evolution in UO2

Due to the large formation energy of vacancies and noble gas atoms in the form of interstitials or substitutional atoms in nuclear fuel (UO2), the thermodynamic equilibrium concentrations of these species are very low in the nuclear fuel matrix even at very high temperatures, which imposes difficulties upon the quantitative study of bubble evolution via the phase-field method. In this study, a quantitative phase-field model is proposed to deal with this problem. The system’s free energy density is derived according to the principles of thermodynamics, with consideration of the elastic effect and with the use of real material parameters from experiments. The model is useful for the study of the kinetics of gas bubble growth with very dilute concentrations of vacancy and gas atoms in the matrix. This model is applied to study single bubble growth and multiple bubble growth under various concentrations of vacancy and gas atoms and at various temperatures. The elastic effect and the effects of the generation rate of vacancies and gas atoms on bubble growth are analyzed.

low defect concentration, nuclear fuel, gas bubble↗

Sensitivity analysis of thermal contact conductance modeling to inform MiniFuel irradiation capsule designs

The MiniFuel irradiation platform has been developed by Oak Ridge National Laboratory as a flexible, high-throughput separate effects testing capability within the High Flux Isotope Reactor (HFIR). Finite element thermal models are relied upon to design MiniFuel experiments to achieve a specific time-averaged irradiation temperature for experimental objectives. A previous study identified that uncertainty in the component heat generation rates and thermal contact conductance (TCC) model are the most significant contributors to predicted fuel temperature variance. To address both sources of uncertainty, this work performs sensitivity analysis on the TCC model to identify high-impact, high-uncertainty parameters that contribute to fuel temperature variance. The TCC model is analyzed in increasing detail, first using a standalone Python code, then again after coupling Python to the BISON fuel performance code. Furthermore, the parameters with the largest contributions to fuel temperature variance which can be reduced through design changes are identified as the initial subcapsule gas pressure, contact pressure between the fuel and dish, and the effective surface roughness of the interface. A set of design recommendations for future capsule designs has been established and applied to reduce the previously quantified average fuel temperature uncertainty ranges of ± 40 °C in the HFIR vertical experiment facilities (VXF) and ± 80 °C in the removable beryllium (RB) reflector to approximately ± 32 °C and ± 53 °C, respectively. This equates to a 21 % and 33 % reduction in the uncertainty range of the average fuel temperature for VXF and RB, respectively.

BISON↗

A MOOSE-Based Model for Fission Product Transport and Source Term Estimation for High-Temperature Gas-Cooled Reactors

Thanks to fuel elements containing tristructural isotropic (TRISO) particles combined with a low core power density and passive feedback mechanisms leading to modest temperature rises in the event of accidental events, high-temperature gas-cooled reactors (HTGRs) offer a high degree of reliability in terms of fission product retention. While the anticipated source term for HTGRs is expected to be very low, it is important to provide a quantitative estimate of radiological releases during nominal and accidental conditions. Here, we propose a computationally efficient mechanistic source term methodology relying on the Multiphysics Object Oriented Simulation Environment (MOOSE) for tracking fission product transport from TRISO particles up to the coolant pressure boundary, as well as modeling the transport and potential deposition of these nuclides inside the reactor coolant loop. The proposed computational scheme is applied to estimate source term inventories for a representative 10-MW(thermal) prismatic high-temperature microreactor and is qualitatively compared against known release fractions. In addition to providing an alternate analysis tool, this MOOSE model can help reactor designers quantify the influence of key design parameters relevant for studies of radiological dose consequences.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Development of a BISON validation case for the TRISO transient irradiations in NSRR using effective heat capacity methods

The current tristructural isotropic (TRISO) fuel assessment and validation database in BISON primarily covers steady- state irradiation and high-temperature furnace testing. Transient assessment cases are potentially needed to support U.S. industry efforts in designing and deploying commercial reactors using TRISO fuels. Historical transient tests in- volving TRISO fuels used highly conservative conditions compared to the typical high-temperature gas-cooled reactor accident scenarios. Despite this, modeling historical transient tests is fundamental for evaluating BISON’s predictive capabilities, adapting material properties for high-temperature and high-particle-power regimes, and developing a sys- tematic validation approach for TRISO transient applications. This work developed a 1D model of transient experiments carried out at the Nuclear Safety Research Reactor using BISON. BISON’s predictions of energy deposition, UO 2 melting onset, and molten volume fractions are compared against experimental measurements. Melting was modeled using an effective specific heat capacity model for UO 2 . We found that BISON’s predictions are in reasonable agreement with experimental data for low-energy-deposition cases, and that BISON overpredicts melting at higher energy depositions. We also discuss the potential causes of discrepancies between the simulated and measured results and propose ways to further develop the model. Although these simulations used conservative conditions compared to those expected for actual TRISO-fueled reactors, they extended the range of conditions reflected in the data in the existing BISON database for TRISO fuels.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Numerical Investigation of Ammonia/n-Heptane Dual-Fuel Spray Flames Using Large Eddy Simulations

Diesel engines are extensively used in heavy-duty transportation, power generation, and marine vehicles due to their superior thermal efficiency and extended high-load operability compared to spark ignition (SI) engines. However, combustion in diesel engines is generally characterized by locally rich fuel–air mixtures and high combustion temperatures, causing significant amounts of soot and NO x emissions from these engines. Utilizing carbon-free alternative fuels and enhancing fuel efficiency represent promising strategies to mitigate greenhouse gas (GHG) and other emissions in the heavy-duty transportation sector. In this context, ammonia (NH 3 ), as a hydrogen carrier, has received significant attention as a viable substitute for hydrocarbon fuels due to its carbon-free composition, relatively high energy density, and well-established infrastructure. Many previous studies have considered combustion and emission characteristics of ammonia-hydrocarbon fuel blends in engines and simplified flames. But, detailed investigations on the effects of ammonia on the performance of hydrocarbon fuels under engine conditions are lacking. In the present study, we perform large eddy simulations (LES) of the ignition and flame processes in a constant-volume combustion reactor, where n-heptane is injected in an ammonia/air ambient mixture in a diesel-like environment. A detailed and validated reaction mechanism containing 302 species and 1981 reactions is employed. The Engine Combustion Network Spray H experimental data is used to validate the spray model under both non-reacting and reacting conditions. Dual-fuel combustion is simulated using the well-stirred reactor (WSR) approach. Results are presented for two spray cases: (1) single fuel (SF) with n-heptane injected into a mixture of air and combustion products and (2) dual-fuel (DF) with the injection of n-heptane in a mixture of air, ammonia, and combustion products. It is observed that the presence of ammonia has a significant effect on the ignition and flame development processes. With ammonia addition, both the first- and second-stage ignition delay times increase, but the effect of ammonia on the second-stage ignition is significantly more prominent. In addition, the ignition kernel size and growth rate decrease noticeably. For SF spray, the main ignition is characterized by multiple ignition kernels near the spray tip, whereas for DF spray, a single relatively small ignition kernel forms and grows slowly in the downstream direction. The flame development and the final quasi-steady flame structure are also modified due to ammonia. Here, the outcome of this research would enable a better understanding of ammonia–diesel dual-fuel spray flame behavior and guide the development of associated engine combustion strategies.

ammonia↗

Integration of Superconducting Magnetic Energy Storage (SMES) Systems Optimized with Second-Generation, High-Temperature Superconducting (2G-HTS) Technology with a Major Fossil-Fueled Asset

Utility companies and fossil fueled generation assets are seeking combinations of “hybrid energy storage” to allow power to be provided in advance of, and to otherwise complement, peaking natural gas-fired generation. American Maglev Technology of Florida Inc. (AMT) learned during the Phase I program based on interactions with NRG Energy (NRG) that energy storage such as superconducting magnetic energy storage (SMES) can qualify as a Black Start unit in most markets, ensuring orderly re-start of grid operations and fossil fueled power plants and serving as an important asset of hybrid energy storage technology. In conjunction with the University of Houston (UH), AMT and NRG are working together to scale up low-cost, high- efficiency, second-generation high-temperature superconducting (2G-HTS) technology for deployment across several markets, with a primary focus on the commercial development of utility grid-scale SMES (g-SMES) systems. Ultimately the program confirmed that the novel g-SMES design can meet the performance and financial requirements of the fossil power plant industry, while exhibiting continuous grid-voltage regulation; cost-effective, peak-hour energy storage with almost infinite life; increased input/output efficiency; and the capability to undergo millions of charging cycles, without degradation, representing a significant improvement over lithium ion and other conventional storage technologies. We believe wide scale deployment of this technology is feasible and practical when compared to the higher life cycle cost and relatively short lifespan of chemical batteries. Based on these findings, the future for deployment of SMES at many fossil-fueled power plants exceeding 100MW capability is bright. Further de-risking g-SMES technology would pave the way for the construction of a subscale prototype. Eventually, scaled-down versions of SMES are expected to be competitive with the multi-billion-dollar lithium-ion market for a wide range of commercial, industrial, and transportation uses. There is long-term potential for commercialization and widespread deployment of this disruptive technology, which would spawn a new worldwide supply chain and create domestic, high-tech manufacturing jobs in the green industry.

20 FOSSIL-FUELED POWER PLANTS↗

OECD-NEA HTTF Benchmark Progress and Updates

General slides discussing High Temperature Test Facility benchmark progress, Prismatic HTGR deployment, modeling and simulation tools for validating systems, and verification and validation issues. Also discuss collaboration and conclusions of RELAP5-3D validation activities based on HTTF with trends in data and international impact to accelerate deployment of prismatic HTGR microreactors by providing an opportunity for designers to assess their codes against experimental data and solutions from other codes.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

A model for droplet heating and evaporation of water-in-oil emulsified fuel

Water-in-oil (W/O) emulsified fuel is a promising alternative fuel by inducing flash boiling of water at high temperature that can improve the atomization of fuel spray. The heating and evaporation process of emulsified fuel droplet is affected by the diffusion and coalescence of small dispersed water droplets in oil during the heating. In this study, a model is developed for the droplet heating and evaporation of W/O emulsified fuel with these key physics considered. The diffusivity of dispersed water droplets in oil is calculated using the Stokes-Einstein equation. The deactivation temperature of surfactant is selected as a criterion for the water coalescence since the water coalescence is due to the deactivation of surfactant. The process of water coalescence is simplified that the dispersed water droplets coalesce into one single water sub-droplet at the center of the oil droplet instantly as the droplet temperature reaches the deactivation temperature of surfactant. The model is validated against experimental data of single droplet under different heating temperatures, surfactant concentrations, and sizes of dispersed water droplets. Based on the proposed model, the effects of fuel properties and heating conditions on the droplet heating and evaporation are analyzed.

Droplet coalescence↗

Computationally Assisted Design of Ion-conducting Polymers for Anion Exchange Membrane Fuel Cells (2nd Year Report) [Slides]

Phosphonated polymers have been proposed as proton conductors for fuel cells; however, the anhydride formation of phosphonic acid functional groups lowers proton conductivity and hinders their use in fuel cell applications. We used calculations to support this hypothesis by correlating calculated Gibbs free energies for the formation of the anhydride with pK values of the corresponding acid. The results showed that stronger acids are less likely to form an anhydride even at elevated temperatures and supported the design of a new polymer and the high-temperature proton exchange membrane fuel cell (HT-PEMFC) that outperformed other state-of-the-art HT-PEMFCs.

30 DIRECT ENERGY CONVERSION↗

Design and Analysis of a 250 MW Plate-fuel Reactor for Nuclear Thermal Propulsion

Nuclear thermal propulsion (NTP) system is a viable option for deep space missions considering its high thrust and lightweight. Promising fuel options being considered for the high operating temperature of NTP include the graphite composite fuel and the tungsten cermet fuel. As there remains uncertainties in the fabrication and performance of these two fuel elements, alternative designs using fuel plates and more standard fuels that are based on proven and tested nuclear technologies are being considered. This paper provides a summary of thermal hydraulic and neutronic analyses of the straight fuel plate concept used for a 250 MW NTP core. The summary of technical findings from a series of scoping studies might be useful for the future NTP engine designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

High‐Temperature Mechanical Behavior of Single‐Crystal FeCrAl Alloy Under In Situ Micropillar Compression

FeCrAl cladding is one of the candidate materials for the near‐term accident‐tolerant fuel technologies under development. Research on high‐temperature mechanical behaviors of single‐crystal FeCrAl alloy is rather limited. Previous studies have reported the mechanical property of low‐index orientation in single‐crystal FeCrAl alloy at room temperature. However, the critical resolved shear stress to activate slip systems can be orientation and temperature dependent. Here, single‐crystal grains in a coarse‐grained FeCrAl alloy with different crystallographic orientations are selected to preferentially activate {110}<111> slip systems or {112}<111> slip systems. Micropillars are fabricated in the selected single‐crystal grains and tested at elevated temperatures in situ in a scanning electron microscope. The critical resolved shear stresses of {110}<111> slip systems and {112}<111> slip systems are determined at various temperatures. The critical resolved shear stress shows a temperature dependence and orientation independence. This study provides important insight for understanding the deformation mechanisms of FeCrAl alloys at elevated temperatures.

Sun, Tianyi↗

RELAP5-3D Modeling of the OECD-NEA HTTF Benchmark

HTTF at Oregon State University (OSU) Reference: General Atomics’ modular high-temperature gas-cooled reactor Helium cooled, electrically heated (2.2 MW) Prismatic graphite blocks in the core and reflectors Alumina ceramic blocks are used to simulate the core and top and bottom reflectors One-fourth scale in length and diameter Most of the coolant channels in the core are full scale Lower pressure compared to the prototype reactor (0.7 MPa) Over 500 instruments Designed primarily to investigate depressurized (DCC) and pressurized (PCC) conduction cooldown transients

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

High-temperature compressive creep tests of U 3 Si 2 with spark plasma sintering: Experiments and finite element modeling

This paper reports a systematic high-temperature creep behavior of dense U 3 Si 2 pellets using a spark plasma sintering (SPS) apparatus at elevated temperatures and pressures under vacuum conditions. The stress exponent was subsequently derived to be 3.21 at 1173 K and 2.17 at 1223 K, respectively, indicating a grain boundary sliding creep mechanism. The creep activation energy was determined to be 203.6 ± 19.0 kJ/mol, which agrees well with the literature. Finite element modeling was performed using the creep parameters fitted from the strain-time plot. Additionally, the results suggest an excellent match with the experimental data, confirming the validity of the experiments. Microstructure characterizations indicate that the main phase of the specimens after creep tests remains to be U 3 Si 2 , with a 4 μm thick layer of nano-sized particles induced from the diffusion between U 3 Si 2 and alumina disc used to avoid electric current passing through the sample. The successful conduct of creep experiments demonstrates the great potential of SPS to perform high-temperature mechanical testing of nuclear fuels under vacuum conditions. The subsequent finite element modeling exhibits excellent capabilities for accurately predicting material performance in the creep tests and provides a practical tool in evaluating nuclear fuels’ performance for a much-extended time scale.

36 MATERIALS SCIENCE↗

Assembly of MiniFuel Targets for Irradiation of TRISO Fuel Compacts in the High Flux Isotope Reactor

To support the development of Kairos Power’s fluoride-salt-cooled high-temperature reactor, irradiation testing of tristructural isotropic (TRISO) fuel compacts was performed at the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) to collect experimental data on TRISO fuel during high particle-power operation and validate fuel performance models. Fuel compacts containing enriched uranium oxycarbide (UCO), natural UCO (NUCO), or uranium dioxide (UO 2 ) TRISO particles were fabricated at ORNL and inserted into MiniFuel targets for HFIR irradiation. Five MiniFuel targets were successfully assembled, welded, tested, and delivered to HFIR, along with their quality assurance documentation. The targets were inserted into HFIR’s inner vertical experiment facility within the permanent beryllium reflector. Each target contains six fuel compacts and will be irradiated in HFIR for four cycles, with target temperatures of 500, 700, and 900°C. This report summarizes the experiment design, test matrix, and fabrication. This work was performed under the Nuclear Science User Facility program.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Rheology of a Wet Waste Feedstock

The current study provides rheological characterization of a representative HTL feedstock, Wet Waste Feedstock # 22 (WW22), as a function of HTL-prototypic temperature and shear rates. Feed WW22 is derived from a regionally representative blend of food, waste water sludges, and FOG (fats, oils, and greases). Rheological characterization of WW22 as a function of temperature provides, in part, key information for development and economic assessment of HTL wet waste slurry handling and heating operations, which may facilitate improved design rigor and optimization of HTL unit operations.

09 BIOMASS FUELS↗

NDMAS Portal Updates

Overview of current ART-GCR Data, including fuel fabrication, irradiation monitoring, fuel and graphite, post-irradiation examination, graphite characterization, high temperature metals mechanical tests, methods validation data (including: HTTR, NSTF, and HTTF), addition of AGR fuel data, other GCR and related data, updated high temperature metals, ongoing and upcoming work, users guide, and the link to those interested in access to NDMAS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗