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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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Simulating thermal environment, transparent walls, and propellant heating in nuclear light bulb engine
Initial radio frequency gas heating experiments to simulate thermal environment in nuclear light bulb reactor
High intensity RF radiant energy source to simulate thermal environment in nuclear light bulb engine
The planar GaAs Shottky diode frequency multiplier is a critical technology for the local oscillator (LO) for submillimeter- wave heterodyne receivers due to low mass, tenability, long lifetime, and room-temperature operation. The use of a W-band (75-100 GHz) power amplifier followed by a frequency multiplier is the most common for submillimeter-wave sources. Its greatest challenge is to provide enough input power to the LO for instruments onboard future planetary missions. Recently, JPL produced 800 mW at 92.5 GHz by combining four MMICs in parallel in a balanced configuration. As more power at W-band is available to the multipliers, their power-handling capability be comes more important. High operating temperatures can lead to degradation of conversion efficiency or catastrophic failure. The goal of this innovation is to reduce the thermal resistance by attaching diamond film as a heat-spreader on the backside of multipliers to improve their power-handling capability. Polycrystalline diamond is deposited by hot-filament chemical vapor deposition (CVD). This diamond film acts as a heat-spreader to both the existing 250- and 300-GHz triplers, and has a high thermal conductivity (1,000-1,200 W/mK). It is approximately 2.5 times greater than copper (401 W/mK) and 20 times greater than GaAs (46 W/mK). It is an electrical insulator (resistivity approx. equals 10(exp 15) Ohms-cm), and has a low relative dielectric constant of 5.7. Diamond heat-spreaders reduce by at least 200 C at 250 mW of input power, compared to the tripler without diamond, according to thermal simulation. This superior thermal management provides a 100-percent increase in power-handling capability. For example, with this innovation, 40-mW output power has been achieved from a 250-GHz tripler at 350-mW input power, while the previous triplers, without diamond, suffered catastrophic failures. This breakthrough provides a stepping-stone for frequency multipliers-based LO up to 3 THz. The future work for this design is to apply the high output power from both the 250 and 300 GHz to multiple chains in order to generate milliwatts at 2.3 THz. Using the first generation of results for this innovation, 40 mW of output power were produced from a 240-GHz tripler at 350-mW input power, and 27- mW output power was produced from a 300-GHz tripler at 408-mW input power. This is two times higher than the current state-of-the-art output power capability. A finite-element thermal simulation also shows that 30-microns thick diamond dropped the temperature of the anodes by at least 200 C.
Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic- metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.
Four nuclear thermal rocket (NTR) models have been created in the Numerical Propulsion System Simulation (NPSS) framework. The models are divided into two categories. One set is based upon the ZrC-graphite composite fuel element and tie tube-style reactor developed during the Nuclear Engine for Rocket Vehicle Application (NERVA) project in the late 1960s and early 1970s. The other reactor set is based upon a W-UO2 ceramic-metallic (CERMET) fuel element. Within each category, a small and a large thrust engine are modeled. The small engine models utilize RL-10 turbomachinery performance maps and have a thrust of approximately 33.4 kN (7,500 lbf ). The large engine models utilize scaled RL-60 turbomachinery performance maps and have a thrust of approximately 111.2 kN (25,000 lbf ). Power deposition profiles for each reactor were obtained from a detailed Monte Carlo N-Particle (MCNP5) model of the reactor cores. Performance factors such as thermodynamic state points, thrust, specific impulse, reactor power level, and maximum fuel temperature are analyzed for each engine design.
We report over the past decades, several computer codes have been developed for simulation and analysis of thermal-hydraulics and system response in nuclear reactors under operating, abnormal transient, and accident conditions. However, simulation errors and uncertainties still inevitably exist even while these codes have been extensively assessed and used. In this work, a data-driven framework (Optimal Mesh/Model Information System, OMIS) is formulated and demonstrated to estimate simulation error and suggest optimal selection of computational mesh size (i.e., nodalization) and constitutive correlations (e.g., wall functions and turbulence models) for low-fidelity, coarse-mesh thermal-hydraulic simulation, in order to achieve accuracy comparable to that of high-fidelity simulation. Using results from high-fidelity simulations and experimental data with many fast-running low-fidelity simulations, an error database is built and used to train a machine learning model that can determine the relationship between local simulation error and local physical features. This machine learning model is then used to generate insight and help correct low-fidelity simulations for similar physical conditions. The OMIS framework is designed as a modularized six-step procedure and accomplished with state-of-the-art methods and algorithms. A mixed-convection case study was performed to illustrate the entire framework.
Biomass is a source of renewable energy and can undergo thermochemical processes to be converted into gaseous and liquid fuels. Thermal decomposition-induced biomass fouling on the screw feeder is a major challenge in preconversion of pyrolysis, leading to plugging of the feed line. Here, this work intends to gain fundamental understanding of the biomass fouling phenomena across the heat gradient on the screw feeder via a combination of materials characterization and thermal simulation. An actual screw feeder that failed because of biomass fouling was examined to reveal the deposit’s morphology, composition, and mechanical properties. Thermal simulations were developed for both the screw feeder and the wood particles being transported using a combined analytical and numerical approach. The simulated temperature profiles correlated well with the deposit observations, based on which potential mitigations were proposed from both the operation and screw design perspectives and supported by preliminary experimental validation.
A set of tasks have been identified to help demonstrate the viability, performance, and affordability of surface fission systems. Completion of these tasks will move surface fission systems closer to reality by demonstrating affordability and performance potential. Tasks include fabrication and test of a 19-pin section of a Surface Power Unit Demonstrator (SPUD); design, fabrication, and utilization of thermal simulators optimized for surface fission' applications; design, fabrication, and utilization of GPHS module thermal simulators; design, fabrication, and test of a fission surface power system shield; and work related to potential fission surface power fuel/clad systems. Work on the SPUD will feed directly into joint NASA MSFC/NASA GRC fabrication and test of a surface power plant Engineering Development Unit (EDU). The goal of the EDU will be to perform highly realistic thermal, structural, and electrical testing on an integrated fission surface power system. Fission thermal simulator work will help enable high fidelity non-nuclear testing of pumped NaK surface fission power systems. Radioisotope thermal simulator work will help enable design and development of higher power radioisotope systems (power ultimately limited by Pu-238 availability). Shield work is designed to assess the potential of using a water neutron shield on the surface of the moon. Fuels work is geared toward assessing the current potential of using fuels that have already flown in space.
An experimental investigation was made to evaluate two nickel base alloys (Nickel-201 and Inconel-718) in three heat treated conditions. These conditions were: (1) annealed; (2) after thermal exposure simulating a braze cycle; and (3) after a thermal exposure simulating a braze cycle plus one operational lifetime of high temperature service. For the Nickel-201, two different braze cycle temperatures were evaluated. A braze cycle utilizing a lower braze temperature resulted in less grain growth for Nickel-201 than the standard braze cycle used for joining Nickel-201 to Inconel-718. It was determined, however, that Nickel-201, was marginal for temperatures investigated due to large grain growth. After the thermal exposures described above, the mechanical properties of Nickel-201 were degraded, whereas similar exposure on Inconel-718 actually strengthened the material compared with the annealed condition. The investigation included tensile tests at both room temperature and elevated temperatures, stress-rupture tests, and metallographic examination.
The Two-dimensional Implicit Thermal Response and Ablation program, TITAN, was developed and integrated with a Navier-Stokes solver, GIANTS, for multidimensional ablation and shape change simulation of thermal protection systems in hypersonic flow environments. The governing equations in both codes are demoralized using the same finite-volume approximation with a general body-fitted coordinate system. Time-dependent solutions are achieved by an implicit time marching technique using Gauess-Siedel line relaxation with alternating sweeps. As the first part of a code validation study, this paper compares TITAN-GIANTS predictions with thermal response and recession data obtained from arc-jet tests recently conducted in the Interaction Heating Facility (IHF) at NASA Ames Research Center. The test models are graphite sphere-cones. Graphite was selected as a test material to minimize the uncertainties from material properties. Recession and thermal response data were obtained from two separate arc-jet test series. The first series was at a heat flux where graphite ablation is mainly due to sublimation, and the second series was at a relatively low heat flux where recession is the result of diffusion-controlled oxidation. Ablation and thermal response solutions for both sets of conditions, as calculated by TITAN-GIANTS, are presented and discussed in detail. Predicted shape change and temperature histories generally agree well with the data obtained from the arc-jet tests.
Thermal-hydraulic simulations have been performed using computational fluid dynamics (CFD) for the highly-enriched uranium (HEU) design of the IVG.1M reactor at the Institute of Atomic Energy (IAE) at the National Nuclear Center (NNC) in the Republic of Kazakhstan. Steady-state simulations were performed for both types of fuel assembly (FA), i.e. the FA in rows 1 & 2 and the FA in row 3, as well as for single pins in those FA (600 mm and 800 mm pins). Both single pin calculations and bundle sectors have been simulated for the most conservative operating conditions corresponding to the 10 MW output power, which corresponds to a pin unit cell Reynolds number of only about 7500. Simulations were performed using the commercial code STAR-CCM+ for the actual twisted pin geometry as well as a straight-pin approximation. Various Reynolds-Averaged Navier-Stokes (RANS) turbulence models gave different results, and so some validation runs with a higher-fidelity Large Eddy Simulation (LES) code were performed given the lack of experimental data. These singled out the Realizable Two-Layer k-ε as the most accurate turbulence model for estimating surface temperature. Single-pin results for the twisted case, based on the average flow rate per pin and peak pin power, were conservative for peak clad surface temperature compared to the bundle results. Also the straight-pin calculations were conservative as compared to the twisted pin simulations, as expected, but the single-pin straight case was not always conservative with regard to the straight-pin bundle. This was due to the straight-pin temperature distribution being strongly influenced by the pin orientation, particularly near the outer boundary. The straight-pin case also predicted the peak temperature to be in a different location than the twisted-pin case. This is a limitation of the straight-pin approach. The peak temperature pin was in a different location from the peak power pin in every case simulated, and occurred at an inner pin just before the enrichment change. The 600 mm case demonstrated a peak clad surface temperature of 370.4 K, while the 800 mm case had a temperature of 391.6 K. These temperatures are well below the necessary temperatures for boiling to occur at the rated pressure. Fuel temperatures are also well below the melting point. Future bundle work will include simulations of the proposed low-enriched uranium (LEU) design. Two transient scenarios were also investigated for the single-pin geometries. Both were “model” problems that were focused on pure thermal-hydraulic behavior, and as such were simple power changes that did not incorporate neutron kinetics modeling. The first scenario was a high-power, ramp increase, while the second scenario was a low-power, step increase. A cylindrical RELAP model was also constructed to investigate its accuracy as compared to the higher-fidelity CFD. Comparisons between the two codes showed good agreement for peak temperatures in the fuel and at the cladding surface for both cases. In the step transient, temperatures at four axial levels were also computed. These showed greater but reasonable discrepancy, with RELAP outputting higher temperatures. These results provide some evidence that RELAP can be used with confidence in modeling transients for IVG.
Thermal runaway (TR) modeling is one of the primary tools that can be used to overcome challenges associated with lithium-ion battery (LIB) safety. Among all LIB accidents that have occurred over the past decade, Internal Short Circuit (ISC) remains the most common trigger mechanism. Many available models in the literature either use a simplified approach to simulate ISC or completely ignore its contribution. The aim of this study is to understand the nature of the heat released for different types of ISC scenarios, including aluminum-anode, anode–cathode, and copper-cathode ISC. We study ISC behavior using a coupled electrochemical–thermal model with an integrated TR chemical kinetics solver built in the COMSOL Multiphysics framework. The time duration of heat release and the magnitude of the peak ISC current are studied as functions of parameters such as the size of the penetrating filament and the capacity of the cell. The numerical results are used to build an empirical model validated against the published experimental TR propagation data. Our model can be successfully used as a viable low-cost substitute in lower order (lumped) TR simulations to enable TR prevention and mitigation.
Coarse austenite to bainite transformation in low carbon steel under simulated welding thermal cycles was morphologically and crystallographically characterized by means of optical microscope, transmission electron microscope and electron backscattered diffraction technology. The results showed that the main microstructure changes from a mixture of lath martensite and bainitic ferrite to granular bainite with the increase in cooling time. The width of bainitic laths also increases gradually with the cooling time. For a welding thermal cycle with relatively short cooling time (e.g. t{sub 8/5} is 30 s), the main mode of variant grouping at the scale of individual prior austenite grains changes from Bain grouping to close-packed plane grouping with the progress of phase transformation, which results in inhomogeneous distribution of high angle boundaries. As the cooling time is increased, the Bain grouping of variants becomes predominant mode, which enlarges the effective grain size of product phase. - Highlights: • Main microstructure changes and the width of lath structure increases with cooling time. • Variant grouping changes from Bain zone to close-packed plane grouping with the transformation. • The change of variant grouping results in uneven distribution of high angle grain boundary. • Bain grouping is main mode for large heat input, which lowers the density of high angle boundary.
Uranium dioxide (UO 2 ) is the most commonly used fuel in light-water nuclear reactors and thermal conductivity controls the removal of heat produced by fission, thereby governing fuel temperature during normal and accident conditions. The use of fuel performance codes by the industry to predict operational behavior is widespread. A primary source of uncertainty in these codes is thermal conductivity, and optimized fuel utilization may be possible if existing empirical models are replaced with models that incorporate explicit thermal-conductivity-degradation mechanisms during fuel burn up. This approach is able to represent the degradation of thermal conductivity due to each individual defect type, rather than the overall burn-up measure typically used, which is not an accurate representation of the chemical or microstructure state of the fuel that actually governs thermal conductivity and other properties. To generate a mechanistic thermal conductivity model, molecular dynamics (MD) simulations of UO 2 thermal conductivity including representative uranium and oxygen defects and fission products are carried out. These calculations employ a standard Buckingham-type interatomic potential and a potential that combines the many-body embedded-atom-method potential with Morse-Buckingham pair potentials. Potential parameters for UO 2+x and ZrO 2 are developed for the latter potential. Physical insights from the resonant phonon-spin-scattering mechanism due to spins on the magnetic uranium ions are introduced into the treatment of the MD results, with the corresponding relaxation time derived from existing experimental data. High defect scattering is predicted for Xe atoms compared to that of La and Zr ions. Uranium defects reduce the thermal conductivity more than oxygen defects. For each defect and fission product, scattering parameters are derived for application in both a Callaway model and the corresponding high-temperature model typically used in fuel-performance codes. The model is validated by comparison to low-temperature experimental measurements on single-crystal hyperstoichiometric UO 2+x samples and high-temperature literature data. Furthermore, this work will enable more accurate fuel-performance simulations and will extend to new fuel types and operating conditions, all of which improve the fuel economics of nuclear energy and maintain high fuel reliability and safety.
Thermal simulators (highly designed heater elements) developed at the Early Flight Fission Test Facility (EFF-TF) are used to simulate the heat from nuclear fission in a variety of reactor concepts. When inserted into the reactor geometry, the purpose of the thermal simulators is to deliver thermal power to the test article in the same fashion as if nuclear fuel were present. Considerable effort has been expended to mimic heat from fission as closely as possible. To accurately represent the fuel, the simulators should be capable of matching the overall properties of the nuclear fuel rather than simply matching the fuel temperatures. This includes matching thermal stresses in the pin, pin conductivities, total core power, and core power profile (axial and radial). This Technical Memorandum discusses the historical development of the thermal simulators used in nonnuclear testing at the EFF-TF and provides a basis for the development of the current series of thermal simulators. The status of current heater fabrication and testing is assessed, providing data and analyses for both successes and failures experienced in the heater development and testing program.
Accurate thermal performance calculation procedures for opaque spandrel areas in curtain wall and window wall systems are essential for rating systems when comparing spandrel systems. However, there is a lack of consensus in thermal modeling needed for accurately characterizing heat transfer through spandrel assemblies due to the complex arrangement of materials and structural components. Several studies indicate that conventional 2D thermal simulations may overestimate R-values by 30% compared to physical testing and 3D simulations. Detailed simulations and well-curated laboratory test data are necessary to build confidence in simulation models, which will later be used to develop correlations to improve widely used conventional 2D thermal simulations. This study aims to experimentally test heat transfer through various spandrel assemblies to validate 3D simulation models. Also, the challenges of conducting a thorough testing design along with the solutions would be documented. The team developed a design for testing spandrel assemblies, making appropriate modifications to the existing heat, air, and moisture (HAM) chamber to accommodate the testing needs. Two moveable baffles were designed and fabricated to guide airflow direction parallel to the test article surface. The data acquisition capabilities in the chamber were upgraded to add more than two hundred sensors to the climate and indoor side of the chamber. The goal is to provide a quality dataset for validating complex 3D modeling simulations, which will be used to develop improved thermal simulation techniques that more accurately represent the thermal behavior of spandrel assemblies and their integration within the building envelope. This paper will summarize the results for the boundary conditions of the testing and the temperature variation across different locations of the spandrel assemblies.
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