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At least 73 records · Page 4

Nuclear fusion powered Titan aircraft

This paper discusses a system for Titan exploration enabled by nuclear fusion power. Titan is one of the most interesting locations in the solar system with a thick atmosphere, surface oceans, under-ice oceans and complex terrain. This paper provides a conceptual design of a fusion-powered system to explore many parts of Titan and enable the use of high-power instruments. The design includes a fusion-powered orbital transfer vehicle and an electric Titan science aircraft. A Direct Fusion Drive (DFD) propulsive engine could bring a sizable spacecraft to Titan orbit in less than two years. A second fusion reactor, configured as a closed-loop power generator, would be used for an electric Titan science aircraft. Both reactors are based on the Princeton Field-Reversed Configuration (PFRC) concept which combines an FRC with a magnetic mirror. PFRC uses a novel radio-frequency plasma heating system and deuterium-helium-3 fuel. A lower temperature plasma flows around the closed-field FRC region removing the fusion products. In the DFD propulsive configuration, this secondary flow permits direct and variable thrust and exhaust velocity. The science aircraft would do a powered entry to Titan and then have the capability to fly anywhere on the moon at subsonic speeds. The DFD-powered transfer vehicle would allow the in-orbit transfer stage to change inclination as needed to cover different areas of the surface.

33 ADVANCED PROPULSION SYSTEMS↗

Hot Hydrogen Exposure of U x Zr 1-x C y Nuclear Fuel: The Effect of Additional Carbon

Nuclear thermal propulsion (NTP) using hydrogen as propellant in a solid-state fission reactor to reach temperatures of up to 3000K can achieve significantly higher specific impulse than chemical propulsion. A promising fuel that is again considered for NTP is U x Zr 1-x C y . To reduce the overall mass loss as well as the uranium fuel loss, a range of experiments using additional carbon in hot hydrogen (2600 K, 6 SLPM flow rate) were performed for up to 5-6 h. Both a CH 4 addition (0.1 and 0.2 vol%) in the hydrogen stream and the use of sacrificial graphite pieces upstream of the samples were tested, with sample compositions from 5 at% UC to 30 at% UC. The results showed a strong reduction of mass loss, up to a factor of 3, as well as a reduction of surface uranium losses in the presence of additional carbon in the atmosphere. Even with additional carbon in the atmosphere, material with 30 at% UC was not stable, but material with 20 at% UC was stable for up to 6h. The sample surfaces showed texturing after processing, possibly by grain growth on the surface. This led to a separation into grains with high uranium content close to the (100)-orientation, and with low uranium content in grains close to the (111)-orientation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hot Hydrogen Testing of W-Coated UN Kernels in a Mo30W Matrix

Ceramic uranium mononitride (UN) is being considered as a reactor fuel for nuclear thermal propulsion. To avoid or reduce the dissociation of UN at the high temperatures needed, embedding it in a metallic matrix (cermet) has been proposed. To assess the viability of this concept, hot hydrogen testing of tungsten-coated UN kernels embedded in a Mo-30 wt% W (Mo30W) alloy matrix has been performed at temperatures from 1800°C to 2300°C. Both the isolated kernels and kernels consolidated by spark plasma sintering in the Mo30W matrix were tested. In addition to direct observations and mass loss measurements, the samples were analyzed by X-ray diffraction (XRD) and scanning electron microscopy (SEM)/energy dispersive X-ray spectroscopy (EDS) after each run. The decomposition of UN started at 1800°C despite the coating and matrix, and increased at 2000°C. Uranium seeped through the tungsten grain boundaries of the coating at all temperatures. The consolidated sample expanded irregularly at 2000°C through the formation of voids, and SEM/EDS analysis showed uranium-containing veins in the matrix consisting of U 2 Mo according to the XRD data. The observed pore generation at 2000°C was explained by the formation of water vapor from residual oxides and diffused hydrogen. At 2200°C and above, both the kernels and the consolidated samples melted through the formation of uranium or low–melting point uranium-molybdenum alloys.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Density functional theory calculations of the mixing enthalpy of ternary uranium carbide compounds

The high melting point of uranium-zirconium carbides (U,Zr)C makes them an ideal fuel for nuclear thermal propulsion (NTP) reactors. Gaps remain in the current understanding of the U-Zr-C system due to the difficulty of conducting thermodynamic experiments at NTP operation conditions. Density functional theory calculations using the Hubbard U model (DFT+U) were performed using orbital matrix occupation (OMC) to obtain the mixing enthalpy for UC and ZrC for (U,Zr)C ternary compounds. Similarly, DFT+U calculations were also carried out for the (U,Nb)C and (U,Ta)C systems. In conclusion, the DFT results are envisioned to be used in thermodynamic assessments of the uranium carbide systems based on the CALPHAD approach to supplement the lack of experimental data for the mixing thermodynamics.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Rare Earth Carbide (Nd-C and Ce-C) Synthesis and Characterization to Inform Phase Equilibrium in Advanced Nuclear Fuels

As advances are being made regarding the performance of nuclear fuels, uranium carbides, and composites, such as (U,Zr)C and UO2 + UCx, have recently gained significant interest for deployment in nuclear space propulsion and high temperature gas-cooled reactors, respectively. However, the phase equilibria of several fission products in carbide systems remain unknown and may impact the overall fuel performance, specifically for particle nuclear fuels that are designed for commercial nuclear energy. Furthermore, comprehensive thermodynamic data on Rare Earth (RE) carbides, such as the Nd-C and Ce-C binary systems, remain limited. Presented in this study are the synthesis methods and characterizations of several Nd-C and Ce-C compositions. The findings from this research provide insights on the stability of RE-C binaries that form in irradiated nuclear fuels and address a critical knowledge gap in the current state of thermodynamics for two key RE-C systems.

Cavazos, Steven J. (ORCID:0009000130329363)↗

Tabulated Fluid Properties Research Report

The Multiphysics Object-Oriented Simulation Environment (MOOSE) enables a wide range of advanced nuclear reactor simulations.[6] Under the guidance of MOOSE’s Thermal Hydraulics Team,I worked to expand the capabilities of Tabulated Fluid Properties (TFP) in the fluid properties module. The fluid properties module allows the user to determine a variety of fluid properties by interpolating points between tabulated data. I implemented the ability to use bilinear interpolation instead of bicubic interpolation for interpolating tabulated data. I also changed the method of variable set inversions to use a 2-dimensional Newton’s Method utility that I created. Variable set inversions are often done from (v,e) to (p,T), where v is specific volume, e is specific internal energy, p is pressure and T is temperature. New routines have also been added into TFP such that it can be used with more applications, such as the Navier Stokes and Thermal Hydraulics modules in MOOSE for Pronghorn[5] and RELAP-7[1] respectively. This work was spurred by interest from NASA in testing a Nuclear Thermal Propulsion (NTP) engine system. NTP engines have drastically different fluid properties throughout the engine and Tabulated Fluid Properties provides the flexibility needed to properly simulate and test these engines.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material, the interior surface of which is lined in higher temperature regions with an insulation layer of porous refractory ceramic material. Continuous insulation layer extends the length of the fuel assembly or separate insulation layer sections have a thickness increasing step-wise along the length of the fuel assembly from upper (inlet) section towards bottom (outlet) section. A fuel element positioned inward of the insulation layer and between support meshes has a fuel composition including HALEU and the form of a plurality of individual elongated fuel bodies or one or more fuel monolith bodies containing coolant flow channels. Fuel assemblies are distributively arranged in a moderator block, with upper end of the outer structural member attached to an inlet for propellant and lower end of the outer structural member operatively interfaced with a nozzle forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material (e.g., SiC—SiC composite), insulation layer of porous refractory ceramic material (e.g., zirconium carbide with open-cell foam structure or fibrous zirconium carbide), and interior structural member of refractory ceramic-graphite composite material (e.g., zirconium carbide-graphite or niobium carbide-graphite). Spacer structures between various layers provide a defined and controlled spacing relationship. A fuel element bundle positioned between support meshes includes a plurality of distributively arranged fuel elements or a solid, unitary fuel element with coolant channels, each having a fuel composition including high assay, low enriched uranium (HALEU). Fuel assemblies are distributively arranged in a moderator block and the upper end of the outer structural member is attached to a metallic inlet tube for hydrogen propellant and the lower end of the outer structural member is interfaced with a support plate, forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗

Status Report on Design of In-situ Thermomechanical Testing at LANSCE

Nuclear fuel encounters severe thermomechanical environments in which its mechanical response is determined by its microstructure, temperature and stress level histories. Simulating the response of such microstructures is crucial for predicting both performance and transient fuel mechanical responses and experimental verification of such predictions is therefore of great interest. While most of the deformation in a nuclear fuel rod occurs in the cladding, deformation of the fuel itself is still of interest with deformation mechanisms at operating temperature and above including creep, swelling, cracking as well as pellet-clad interaction. Characterization of these properties and understanding of the underlying deformation phenomena at operating or excursion temperatures is therefore of great importance for development and ultimately licensing of improved and novel nuclear fuel forms. Diffraction techniques offer unique insight on the atomistic (e.g. crystal structure) and microstructure (e.g. phase transformations, texture, defects) length scales and have a long history of providing unique data to inform relevant deformation models that enable the required predictive capabilities. For example, dislocations lead to diffraction peak broadening that can be characterized to estimate the dislocation density and study the role of dislocations on the deformation while measuring lattice strains allows to studie load sharing in two phase materials. In this report the requirements for a sample environment for high temperature deformation of nuclear fuels are defined. The HIPPO neutron time-of-flight diffractometer at LANSCE will host this sample environment and is also described. This instrument covers diffraction angles from 140° to 40° and is also equipped with an event-mode neutron imaging detector system, enabling energy-resolved neutron imaging in parallel with the diffraction that could measure sample temperature from Doppler broadening of neutron absorption resonances or measure pore densities from changes in the attenuation. Designs of devices to characterize thermomechanical properties of nuclear fuel without diffraction are also considered to guide the design. While this report is focused on applications for nuclear fuels, the device can also characterize cladding, moderator or structural materials and therefore contribute to other fields of research and development for advanced reactors. The temperatures planned to be reached are above 2000℃, thus enabling characterization of LWR reactor fuels under accident scenarios but also reaching temperatures of fuels developed for nuclear thermal propulsion and providing opportunities to characterize those. In conjunction with the energy-resolved neutron imaging detector, this setup would allow to measure neutron cross-sections at high temperatures, filling a gap towards development of reactors operating at high temperatures.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Editorial: Benchmark experiments, development and needs in support of advanced reactor design

Advanced nuclear reactor designs will for the most part be a departure from low enrichment light water reactor (LWR) designs currently operated around the world. Such advanced designs include but are not limited to new TRISO-fueled high temperature gas reactors, heat-pipe cooled micro-reactors, fluoride salt cooled high-temperature reactors, molten salt reactors, lead cooled fast reactors, nuclear thermal propulsion concepts, and include LWR designs with advanced fuel and clad types. Modeling and simulation methods for advanced reactors is necessary for regulators to approve license requests. However, regulators also require that modeling approaches be validated against experimental measurements. Hence, there is a crucial need for data for advanced reactor systems that will support validation of analysis methods. To this end, this Research Topic includes eleven papers organized into topical seven categories relevant for advanced reactor design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Mobility assessment of the BCC and carbide phases in the C-Nb, C-U and Nb-U systems

Uranium carbides with refractory metal additions are considered for Gen IV nuclear reactors and nuclear thermal propulsion as fuels for their high-temperature and corrosion resistant properties. Understanding kinetic effects that dictate microstructural evolution during fabrication and operating conditions is essential to advance technological development of these fuels. This work presents the development of an atomic mobility database for C-Nb-U systems based off available experimental data supported with ab-initio methods. The mobility assessments and uncertainty quantification (using Markov chain Monte Carlo) were conducted in the Kawin software. Carbon diffusion is considered dominant, as metal diffusion is much slower, with niobium diffusion being even slower and rate limiting than uranium metal. We provide a comprehensive and self-consistent thermo-kinetic database that is validated by diffusion couple simulations through Kawin. In conclusion, this enables prediction of microstructural and phase evolution critical for the development and lifetime assessment of next generation nuclear fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

TREAT Slide Decks for Alabama A&M

Slides to introduce TREAT and two senior design projects to a professor and senior engineering students at Alabama A&M. Slide Deck 1 is an overview of the TREAT Reactor and its mission. Use for Space Nuclear Thermal Propulsion Testing is mentioned, but no details of space reactor fuel are included. Review for unlimited release is requested. Slide deck 2 presents technical background for neutron radiography, with contrast to x-radiography, and then presents the application of TREAT for Flash Radiography. Slide Deck 3 describes the background behind the second senior design project, which is to design powered actuators for TREAT core clamps. Slide Deck 3 includes photos of different sides of the TREAT reactor and sketches of internal components associated with the core clamping system. These were presented via Teams on October 5, but we want to provide the slidedeck files to the university for student reference.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Uncertainty Quantification and Sensitivity Analysis of Non-Nuclear Advanced Controls Testbed Reactor Mockup

The research presented in this report describes our progress in applying stochastic methods and uncertainty quantification, parametric study, and variance-based sensitivity analysis (also known as Sobol sensitivity analysis) to a full-core model of a nuclear thermal propulsion (NTP) system simulated with Griffin, with the goal of developing a reduced order (surrogate) model which can be rapidly sampled while perturbing multiple input parameters. In this NTP system, reactivity and power feedback affect the rotation of control drums, which are controlled by a hybrid proportional, integral and derivative (PID) controller, actuated by the power demand and reactivity feedback from the numerical model. This model uses reactor kinetic feedback (mean generation time and $\beta$ from a transient Griffin simulation executed with the improved quasi-static method to provide the kinetic parameters) as inputs to functions which control the CD rotation angle. Using a number of stochastic method approaches, we developed a dual purpose training-surrogate model of the NTP system using polynomial regression. The trained model can be rapidly sampled while simultaneously perturbing various input parameters of the model, such as coefficients on the PID control, or temperature (directly affect the neutron cross section). The surrogate model delivers accurate results orders-of-magnitude faster (minutes, not days) than the base model. Once the base model has been trained, distributions of the uncertain parameters can be changed at will to investigate the effects of perturbing multiple inputs and their effect on the output. For example, coefficients used in the PID control system may vary due to some physical interference, or there may be uncertainty in the temperature of the neutron cross sections in various regions of the reactor. A distribution can be placed on these parameters and operational boundaries can be determined. The goal of this work is to support development of an advanced control system to operate CDs in a functioning NTP system.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Nuclear Thermal Rocket Emulator for a Hardware-in-the-Loop Test Bed

To support NASA’s mission to use nuclear thermal rockets for future Mars missions, an instrumentation and control test bed has been built at Oak Ridge National Laboratory. The system is designed as a hardware-in-the-loop test bed for testing control elements and autonomous control algorithms for nuclear thermal propulsion rockets. The mock reactor system consists of a modular and scalable framework, using inexpensive components and open-source software. The hardware system consists of a two-phase flow loop and a mock reactor with six control drums. A single-board computer (NVIDIA Jetson) handles reactor core emulation and hosts a message queuing telemetry transport broker that allows user-deployed control algorithms to interact with the system hardware. The reactor emulator receives sensor data from the hardware and provides the simulated performance of the reactor under steady-state, transient, and fault conditions. The emulator uses a reactivity lookup table and the point kinetics equations to solve for the reactor dynamics in real time. Emulated reactor dynamics and sensor input inform the autonomous control algorithm’s decision-making in a closed-loop manner. The current system is capable of operating at 10 Hz, but faster cycle rates are an area of ongoing research. This test bed will enable NASA and other space vendors to rigorously test their autonomous control systems for NTP rockets under transient (reactor startup and shutdown), steady-state, and fault conditions to reduce development time and risk for autonomous control systems in future missions.

autonomous control↗

Post-irradiation examination of commercial tantalum alloys following neutron irradiation

Post irradiation examination of two commercial tantalum alloys (based on Ta-8 %W) revealed significant irradiation hardening following neutron irradiation in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Tensile samples of commercial Ta alloys T-111 and ASTAR-811C were irradiated to a total neutron fluence (E > 0.1 MeV) in the range (2–4)x10 21 n/cm 2 (0.39–0.75 dpa) at an irradiation temperature near 900 °C. Tensile testing conducted at room temperature and at 800 °C revealed significant irradiation hardening under all test conditions. Samples that were tensile tested at room temperature exhibited brittle failure, whereas ductility was maintained for elevated temperature testing. The irradiation hardening and embrittlement noted for both alloys was attributed to irradiation-induced dislocation loop formation as confirmed through scanning transmission electron microscopy analysis. Corresponding energy dispersive x-ray spectroscopy revealed radiation-induced segregation of Hf in these loop populations. This Hf enrichment is shown to result in eventual co-precipitation of elongated (Hf,O)-rich precipitates along the dislocation loops. Further, these results show that irradiation hardening is enhanced via irradiation-enhanced precipitation at elevated irradiation temperatures. Consequently, the minimum recommended operating temperature window for Ta alloys should be increased to account for observed hardening/embrittlement at irradiation temperatures as high as 900 °C (1,173 K).

36 MATERIALS SCIENCE↗

Effects of a refractory metal protective barrier on uranium carbonitride ceramic–metallic composites exposed to a high temperature hydrogen environment

Uranium nitride has been under consideration at NASA and the Department of Energy for nuclear thermal propulsion applications intended for deep space exploration vehicles. It is critical to the success of the technology development that the behavior of the fuel system is properly characterized in operational environments. This body of work fabricated ceramic metal composites consisting of uranium ceramic fuel in a molybdenum and molybdenum–tungsten metal matrix, both with and without an exterior molybdenum barrier. Specimens were introduced to a high temperature hydrogen environment and characterized through x-ray diffraction, scanning electron microscopy, and energy dispersive spectroscopy to identify characteristic changes. A portion of the specimens was fabricated with a 1.27 mm thick molybdenum barrier in order to isolate the material from the hydrogen environment. Instability of the uranium nitride occurred as low as 2073 K, leading to aggressive instability approaching 2658 K. Carbon reactions with the metal matrix were observed as well as alloying between the uranium and metal matrix constituents. The liquid uranium resulting from the instability of the uranium nitride was observed to attack the grain boundaries of the refractory metal, resulting in the formation of a liquid uranium molybdenum eutectic melt. Specimens encased in a barrier experienced similar results. The authors conclude this is a significant challenge to applications of this cermet at these temperatures. Alternate configurations with resilient hermetic seals or lower temperature applications such as power reactors may have more success.

36 MATERIALS SCIENCE↗

Thermocouple and RTD Drift Due to Neutron Irradiation and Temperature for Space Nuclear Applications: A Review

Temperature measurement is a key factor in the safe and reliable operation of nuclear reactors. Thermocouples and resistance temperature detectors (RTDs) are among the most common and reliable tools deployed for temperature measurement, and some are capable of operating in extreme environments such as very high temperature and neutron radiation fields anticipated in nuclear thermal propulsion (NTP) or fission surface power. These devices are known to exhibit drifting behavior in their temperature readings throughout their operational lifetimes. Here, to help select appropriate instrumentation, it is important to identify the factors that influence drift as well as typical behavior in these environments. This work is focused on identifying contributing factors in the performance of thermocouples and RTDs. An extensive literature review is included that discusses several experimental results in different environments. Although individual instrument performance may vary with construction methods, materials selection, and vendors, this review aims to assist in the selection process and identify typical expected behaviors of several devices.

Chesser, Ryan J. [Oak Ridge National Laboratory (O↗

Plenum plate standoff spools for separating a pair of plenum plates in a nuclear reactor

A nuclear reactor has first and second plenum plates disposed in a pressure vessel. Both plenum plates have a plurality of apertures. The second plenum plate is parallel to the first plenum plate. A fuel element includes a fuel element coolant flow tube which extends through aligned apertures of the parallel plenum plates. A fuel element standoff spool is disposed about a portion of the fuel element coolant flow tube which is located between the plenum plates. The nuclear reactor is also usable in nuclear thermal propulsion.

Inman, James B.↗