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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 343 records · Page 19
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

ALMA OBSERVATIONS OF THE HH 46/47 MOLECULAR OUTFLOW

The morphology, kinematics, and entrainment mechanism of the HH 46/47 molecular outflow were studied using new ALMA Cycle 0 observations. Results show that the blue and red lobes are strikingly different. We argue that these differences are partly due to contrasting ambient densities that result in different wind components having a distinct effect on the entrained gas in each lobe. A 29 point mosaic, covering the two lobes at an angular resolution of about 3'', detected outflow emission at much higher velocities than previous observations, resulting in significantly higher estimates of the outflow momentum and kinetic energy than previous studies of this source, using the CO(1-0) line. The morphology and the kinematics of the gas in the blue lobe are consistent with models of outflow entrainment by a wide-angle wind, and a simple model describes the observed structures in the position-velocity diagram and the velocity-integrated intensity maps. The red lobe exhibits a more complex structure, and there is evidence that this lobe is entrained by a wide-angle wind and a collimated episodic wind. Three major clumps along the outflow axis show velocity distribution consistent with prompt entrainment by different bow shocks formed by periodic mass ejection episodes which take place every few hundred years. Position-velocity cuts perpendicular to the outflow cavity show gradients where the velocity increases toward the outflow axis, inconsistent with outflow rotation. Additionally, we find evidence for the existence of a small outflow driven by a binary companion.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2026

Impact of Molten Gallium on the Microstructure and Corrosion Behavior of Aluminum and Uranium-Aluminum Alloys for Used Nuclear Fuel Reprocessing

Test reactors around the world utilize highly enriched uranium fuel to achieve high neutron fluxes for materials testing. Once spent, the remaining uranium is a valuable resource for subsequent fuel fabrication. However, some of these test reactor cores consist of curved plate-type fuel elements, fabricated using aluminum alloy 6061 (AA6061) cladding to encapsulate a uranium-aluminum alloy (UAlx) fuel matrix. These assemblies require non-standard reprocessing approaches for uranium recovery, as aluminum dissolves readily in acidic solutions, generating large volumes of waste and complicating downstream chemical separations. In this work, we investigate a novel chemical decladding strategy based on the interaction between AA6061/UAlx and molten gallium (Ga). Ga is known to induce severe degradation of aluminum metal through liquid metal embrittlement (LME), even at relatively low Ga concentrations. By penetrating the aluminum crystal lattice, Ga disrupts grain cohesion and facilitates fracture or dissolution of the aluminum matrix. Thermodynamic analysis of the Al–Ga binary phase diagram suggests that Ga may offer a viable pathway to selectively weaken or dissolve the AA6061 cladding, and potentially the aluminum component of the UAlx fuel matrix within. To this end, parametric experiments were performed at 50 °C and 100 °C across a range of Al–Ga atomic fractions. At lower Al fractions, the AA6061 was completely molten after 2 hours of exposure to the Ga metal. In contrast, samples with higher Al fractions (0.9 Al, 0.1 Ga) contained residual solids after 2 hours, which were characterized by microstructural examination using electron backscatter diffraction (EBSD) and transmission electron microscopy (TEM). These Al-Ga compositions were also evaluated using FactSage thermodynamic modeling to further elucidate the relationship between phase diagram behavior and LME.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Property Measurements of NaCl-UCl 3 and NaCl-KCl-UCl 3 Molten Salts (Rev.1)

Thermochemical and thermophysical property values of several salt compositions of interest are needed by molten salt reactor (MSR) developers to design, license, and operate their reactors. Thermochemical and thermophysical properties being measured at Argonne include thermal transitions, phase behavior, heat capacity, density, volumetric thermal expansion of the liquid phase, thermal diffusivity, thermal conductivity, and viscosity. Several properties of eutectic compositions in the ternary NaCl-KCl-UCl 3 and binary NaCl-UCl 3 systems that may be used by MSR developers as fuel bearing salts are being measured. A 65.8 mol % NaCl–34.2 mol % UCl 3 mixture and a near-eutectic mixture of 50.9 mol % NaCl–24.4 mol % KCl–24.7 mol % UCl 3 were synthesized and the thermochemical properties of the mixtures were measured by using differential scanning calorimetry (DSC). The measured transition temperatures were compared to transition temperatures predicted using two models. A thermodynamic model of the binary NaCl-UCl 3 system was constructed using data in the Molten Salt Thermal Properties Database–Thermochemical Version 2.0 (MSTDB-TC V2.0). A ternary NaClKCl-UCl 3 system model constructed at Argonne and was described in a previous report. These comparisons can be used to validate the models. Thermophysical property values of molten salts are needed to model how salt retains and transfers heat in an MSR system. These property values are essential to the entire MSR design because molten salt is used as both the fuel and the coolant material in a salt fueled reactor. Heat capacity of the synthesized NaCl-UCl 3 and NaCl-KCl-UCl 3 salt mixtures was measured by using DSC and thermal diffusivity was measured by using laser flash analysis (LFA) at temperatures spanning the typical operating range of an MSR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Multiscale modeling of fission gas behavior in U 3 Si 2 under LWR conditions

In this work, we present a model of fission gas behavior in U 3 Si 2 under light water reactor (LWR) conditions for application in engineering fuel performance codes. The model includes components for intra-granular and inter-granular behavior of fission gases. The intra-granular component is based on cluster dynamics and computes the evolution of intra-granular fission gas bubbles and swelling coupled to gas diffusion to grain boundaries. The inter-granular component describes the evolution of grain-boundary fission gas bubbles coupled to fission gas release. Given the lack of experimental data for U 3 Si 2 under LWR conditions, the model is informed with parameters calculated via atomistic simulations. In particular, we derive fission gas diffusivities through density functional theory calculations, and the re-solution rate of fission gas atoms from intra-granular bubbles through binary collision approximation calculations. Further, the developed model is applied to the simulation of an experiment for U 3 Si 2 irradiated under LWR conditions available from the literature. Results point out a credible representation of fission gas swelling and release in U 3 Si 2 . Finally, we perform a sensitivity analysis for the various model parameters. Based on the sensitivity analysis, indications are derived that can help in addressing future research on the characterization of the physical parameters relative to fission gas behavior in U 3 Si 2 . The developed model is intended to provide a suitable infrastructure for the engineering scale calculation of fission gas behavior in U 3 Si 2 that exploits a multiscale approach to fill the experimental data gap and can be progressively improved as new lower-length scale calculations and validation data become available.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2026

Microstructure, Transport, and Mechanics of Compacted Clay Simulated at the 0.1 μm Scale (1400 Smectite Clay Particles) Using a Coarse-Grained Model with Explicit Counterions

Clay-rich geomaterials play a critical role in many subsurface systems. The macroscale properties of these materials (low permeability, high ionic conductivity, high swelling pressure, etc.) are sensitive to molecular-level adsorption and hydration interactions at clay−water interfaces. Efforts to develop multiscale simulation approaches to predict these properties reveal a scale gap between atomistic simulations (typically limited to systems smaller than 10 nm) and continuum-scale models (which use computational grid elements with dimensions ≳ 10 μm). In this study, we present a coarse-grained (CG) framework that partly bridges this gap by simulating compacted smectite clay assemblages with dimensions of 0.1 μm containing 1,400 clay particles across a range of dry densities (1,050 to 1,850 kg·m −3 ) and Na/Ca counterion compositions (Na fraction ranging from 0.2 to 1). The simulated systems, along with their reconstructed binary three-dimensional pore networks, are used to evaluate the microstructure, pore size distribution, tortuosity, ion diffusivity, and swelling pressure of compacted smectite clay. Results show that our approach captures important features of the mesoscale heterogeneity of compacted clays, including tactoid formation, hierarchical porosity, and anisotropic pore networks. Results also reveal how compaction and counterion composition govern emergent behaviors, including dominant pore sizes, directional transport, and electrochemical response. This work highlights the potential of CG simulations to bridge molecular and continuum scales and to advance geotechnical and environmental applications involving clay-rich materials as well as related nanoporous media such as geopolymers and calcium-silicate-hydrate. However, the results also suggest that accurate prediction of certain microstructural and mechanical properties (e.g., swelling pressure) may require even larger-scale systems on the order of 1 μm.

02 PETROLEUM↗
DOE OSTI2025

Diffusion and phase formation in the γ-uranium-technetium system

Phase formation in the U-Tc binary system at 800 °C was investigated using a diffusion couple experiment. Scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS) identified four novel potential intermetallic phases - U 7 Tc 3 , U 13 Tc 12 , U 3 Tc 5 , and UTc 4 . Diffusion coefficients were calculated for the intermetallic phases using the Boltzmann-Matano method and were respectively found to be – 120, 38.2, 15.6, and 1.51 × 10 −18 m 2 /s. Tc also exhibits a solid-solution phase with high penetration into the U with a diffusion coefficient of ∼ 10 −14 m 2 /s. Furthermore, these findings expand the number of known U-Tc phases and provide the first diffusion coefficients for the U-Tc system, and contribute valuable data to the broader field of actinide metallurgy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Phase equilibria of advanced technology uranium silicide-based nuclear fuel

The phases in uranium-silicide binary system were evaluated in regards to their stabilities, phase boundaries, crystal structures, and phase transitions. The results from this study were used in combination with a well assessed literature to optimize the U-Si phase diagram using the CALPHAD method. A thermodynamic database was developed, which could be used to guide nuclear fuel fabrication, could be incorporated into other nuclear fuel thermodynamic databases, or could be used to generate data required by fuel performance codes to model fuel behavior in normal or off-normal reactor operations. The U 3 Si 2 and U 3 Si 5 phases were modeled using the Compound Energy Formalism model with 3 sublattices to account for the variation in composition. The crystal structure used for the USi phase was the tetragonal with an I4/mmm space. Above 450°C, the U 3 Si 5 phase was modeled. The composition of the USi 2 phase was adjusted to USi 1.84 . The calculated invariant reactions and the enthalpy of formation for the stoichiometric phases were in agreement with experimental data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

First-principles study of thermophysical properties of interaction layer products in U-Mo/Al dispersion fuel

Thermophysical properties of interaction layer (IL) formed in U-Mo/Al dispersion fuel are important for the evaluation of its impact on the fuel performance. Binary and ternary phases observed in the IL, however, are not well studied yet. Recently we predicted the thermophysical properties of the binary U-Al compounds in U-Al system using density functional theory (DFT) calculations. In this paper, we investigated the structural, elastic, electronic, vibrational and thermodynamic properties of the ternary U-Mo-Al compounds formed in IL, including (U 0.75 ,Mo 0.25 )Al 3 , U 6 Mo 4 Al 43 and UMo 2 Al 20 , using DFT calculations. The polycrystalline aggregate properties of these compounds were obtained from the predicted single crystal elastic constants. The calculated electronic density of states confirm that all of the compounds exhibit typical metallic behavior with the majority states at the Fermi level dominated by U 5 f electrons. Using quasi-harmonic approximation, we predicted the thermodynamic properties of these compounds by including both electronic and lattice contributions. We also investigated the effect of density on the thermodynamic properties of (U 0.75 ,Mo 0.25 )Al 3 phase. Here, the current results are expected to be helpful to the modeling of the fuel performance of U-Mo/Al dispersion fuel and the thermodynamic modeling of the ternary U-Mo-Al system.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Thermodynamic description of molten salt systems: KCl-LiCl-NaCl and KCl-LiCl-NdCl 3

The mixture of KCl and LiCl has been used as electrolyte in the electrorefining process to recover uranium from used nuclear fuels due to the low melting point. However, lanthanides and sodium in the reactor waste continuously dissolve into it and thus alter its thermodynamic properties. To understand the thermodynamic behavior evolution of the electrolyte with the accumulation of impurities, thermodynamic modeling for KCl-LiCl-NaCl and KCl-LiCl-NdCl 3 and four constituent binary systems in the entire composition space was performed using the CALPHAD (CALculation of PHAse Diagrams) approach. The ionic liquid was described by the two-sublattice model, where neutral species were introduced to consider short-range ordering (SRO) within the melt, whereas the solid solution was modeled based on the Compound Energy Formalism. Literature data on phase equilibria and thermochemical properties were critically evaluated and used during the optimization of thermodynamic parameters for KCl-LiCl-NaCl and KCl-LiCl-NdCl 3 and their subsystems. The calculated phase diagrams and mixing enthalpies are in good agreement with the experimental data. The thermodynamic modeling for the KCl-LiCl-NdCl 3 system was carried out for the first time. To fill the gap in experimental measurement, enthalpy of mixing for the KCl-LiCl-NdCl 3 melt was estimated using the surrounded-ion model. These data then served as critical inputs for thermodynamic optimization. Furthermore, the present study can provide insights into thermodynamic property evolution of the electrolyte and solubility limit of various impurities during the electrorefining process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Status Report on the Development of a Thermodynamic Database for Use with Metallic Fuel Design

The scope of this FY18 milestone involves the development of a high fidelity thermodynamic database for the U-Mo-Ti-Zr system to support the accelerated design of new pseudo binary U-xM metallic fuel, where M is a ternary metallic additive composed of Mo, Ti and Zr and x is the additive amount. Based on the previous assessment of the Mo-Ti-Zr ternary system performed in FY17, a complete set of thermodynamic model parameters describing the U-Mo-Ti-Zr quaternary system over the entire composition and temperature range have been developed in this work using the CALPHAD (CALculation of PHAse Diagrams) approach with inputs from both experimental measurements and density functional theory (DFT) total energy calculations. The two ternary intermetallic compounds, the delta phase in the U-Ti-Zr system and the cubic C15 Laves phase in the Mo-Ti-Zr system, are modeled using physically-based two-sublattice compound energy formalism. The liquid, bcc and hcp phases are described using the random solution model. Using the present database, the melting temperatures and bcc onset temperatures of U-xM metallic fuels are predicted as a function of both additive composition and amount. An optimal metallic additive composition has been identified in this work.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Investigation of Interdiffusion Behavior in the Mo-Zr Binary System via Diffusion Couple Studies

Zirconium has recently garnered attention for use as a diffusion barrier between U–Mo metallic nuclear fuels and Al alloy cladding. In order to gain a fundamental understanding of the diffusional interactions, the interdiffusion behavior in the binary Mo–Zr system was investigated via solid-to-solid diffusion couples annealed in the temperature range of 750 to 1050 degrees C. A combination of scanning electron microscopy, X-ray energy dispersive spectroscopy, and electron probe microanalysis were used to examine the microstructure and concentration profiles across the interdiffusion zone. A large __-Zr (cI2) solid solution layer and a thin (approximately 1–2 um) layer of Mo2Zr (cF24) developed in all couples. Parabolic growth constants and concentration dependent interdiffusion coefficients were calculated for the Mo2Zr and Zr solid solution phases, respectively. The pre-exponential factor and activation energy for growth of the Mo2Zr phase were determined to be approximately 6.5 × 10- 15 m2/s and 90 kJ/mol, respectively. The interdiffusion coefficient in ___-Zr solid solution decreased with an increase in Mo concentration. Both the pre-exponential factors (2 × 10- 8 m2/s at 2 at.% Mo to near 5 × 10- 8 m2/s at 9 at.% Mo) and activation energies (140 kJ/mol at 2 at.% Mo to approximately 155 kJ/mol at 9 at.% Mo) of interdiffusion coefficients were determined to increase with an increase in Mo concentration.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Bio-reduction of ferrihydrite-montmorillonite-organic matter complexes: Effect of montmorillonite and fate of organic matter

Organic matter (OM) is often associated with Fe (hydr)oxides such as ferrihydrite (Fh) in soils and sediments, forming binary Fh-OM complexes. Microbial reduction of Fh results in destabilization of the complexes and mineral/OM transformation. However, little is known about the role of clay minerals in such processes, despite their common co-existence with Fh and OM in natural environments. Here Fh-OM complexes were synthesized in the presence of montmorillonite (SWy-2), forming ternary Fh-(SWy-2)-OM complexes. A metal-reducing bacterium Geobacter sulfurreducens was used to reduce Fh in the complexes under circumneutral pH and anoxic conditions with or without H 2 as extra electron donor. Various spectroscopy and mass spectrometry methods were used to monitor the progress of Fh bio-reduction and mineral/OM transformation. Results showed that G. sulfurreducens utilized mineral-bound OM as electron donor and/or carbon source to couple with Fh reduction. Relative to Fh-OM complex, Addition of SWy-2 to Fh-OM complex enhanced the bio-reduction extent of Fh by increasing the proportion of bioavailable OM that was weakly bound to SWy-2. However, its effect on the bio-reduction rate was variable. SWy-2 initially decreased the rate, because it spatially separated OM (electron donor) from Fh (electron acceptor). During later incubation, SWy-2 increased the reduction rate by sorbing biogenic Fe 2+ that would otherwise passivate the Fh and cell surfaces. Bio-reduction transformed mineral-bound OM to microbial products (e.g. necromass, extracellular polymeric substances), but organic compounds with aromatic structures, carboxyl groups and large molecular weight were more resistant to desorption and oxidation. The persistence of these compounds against bio-reduction induced transformation is likely due to their stronger binding with minerals and/or lower nominal oxidation states of carbon relative to other compounds. Our results provide new insights into the role of clay minerals in regulating biogeochemical cycling of solid-phase Fe and transformation of mineral-associated OM in anoxic soil environments.

58 GEOSCIENCES↗
DOE OSTI2013

Nuclear Metal Fuel Casting Modeling Annual Report 2013

Actinide burning fuels in fast reactor applications have been a traditional approach. Recently, the use of metallic fuels in light-water reactors (LWR) applications has also been suggested. Although there has been some experience with the casting of cylindrical metallic binary or ternary fuels in the past, renewed attention to the casting of metallic fuels arises due to the fact that not only do we need to cast traditional constituents such as Pu, Zr, and U, but also add more difficult actinides such as Am and Np. Casting metallic fuel rods shows great promise, but fuel and transuranic losses must be minimized. Issues critical to the manufacture of transuranic containing fuels include fuel holdup in crucibles and molds, fuel loss into dross from crucible and mold wash coatings, generation of contaminated wastes such as mold pieces, and loss of volatile constituents from the crucible during heating and casting. This final issue potentially constrains the caster to a positive pressure design because Am has a high partial pressure allowing it to volatize from the liquid metal under vacuum. There have been several primary studies investigating the actinide volatilization issue in order to improve the caster design. As part of these efforts, a bench scale casting apparatus has been designed. Fuel manufacturing process modeling and simulation is a cost-effective way to explore some of the above-mentioned issues for the development of new manufacturing processes yielding minimum loss.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Summary and Assessment of Metal Fuel Capabilities in Bison

The work described in this report was performed under funding from the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. This report is issued in satisfaction of the level 2 milestone M2MS-18IN0201012 on Release Bison with Initial Metallic Fuel Performance Capability. Metallic fuel development began in Bison a few years after light water reactor development began (~ 2009). The initial development was funded by the Advanced Fuels Campaign. With recent interest from industry, the Nuclear Regulatory Commission, and the Versatile Test Reactor program, the Nuclear Energy Modeling Simulation Program has increased funding for metallic fuel development and validation in FY18. The material models with the most mature development are binary uranium-zirconium and ternary uranium-plutonium-zirconium alloys for fuel and HT9 for cladding. The first material models were incorporated from open literature sources, which include mechanical and thermal material properties that are functions of temperature, porosity, and zirconium concentration, swelling, fission gas release, zirconium diffusion, creep, and sodium coolant channel boundary conditions. These material models have been utilized to simulate ERB-II fuel pins and the results compared to measurements. The EBR-II experiment measurements mostly consist of cladding strain and zirconium redistribution. As such, comparisons to EBR-II measurements are currently limited to these two figures of merit. Simulations and comparisons to experiment measurements have also been done for transient testing of EBR- II fuel pins in TREAT, which included temperature measurements. Preliminary comparisons between Bison calculations and experiment measurements are favorable. These comparisons have also highlighted the importance of the fuel swelling and fission gas release models on cladding strain. Also, Bison metallic fuel simulations are currently being run in support of ATR experiments and VTR exploratory design calculations. Further development and evaluation of individual material models and fuel system models are planned for FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Analysis and Thermal Property Investigations into Ternary Actinide Chloride Salt Systems Containing UCl3 and PuCl3

While regulators, the scientific community, and MSR developers still lack access to literature data on the thermal properties of clean fuel salts, even less information is available on the properties of fuel salts containing impurities. It is essential to understand, benchmark, and predict crucial data on the changes in thermal properties of fuel salt systems due to impurities arising from moisture, corrosion, and reactor operation (i.e., fission products). This research focuses on two actinide fuel salts (1) to investigate a worst-case scenario buildup of actinide fission product in a NaCl-UCl3 eutectic fuel salt and (2) to investigate NaCl-PuCl3 eutectic salt after 1000 hours of operation in a natural circulation flow loop flow to determine if corrosion or atmospheric (moisture/oxygen) products are present. For the first salt, a conservative assumption or worst-case scenario, for fission product buildup in a fuel salt was investigated by adding PuCl3 to eutectic 67 mol% NaCl – 33 mol% UCl3 salt resulting in a ternary salt having a composition of 61 mol% NaCl – 30 mol% UCl3 – 9mol% PuCl3. Addition of PuCl3 to eutectic NaCl-UCl3 resulted in a ternary salt that had a higher melting temperature than either the NaCl-PuCl3 or NaCl-UCl3 binary eutectic mixture. Addition of PuCl3 also resulted in an increase in density which was expected. The second salt was extracted from a micro loop. The composition of the fuel (primary) salt prior to flow loop operations was determined to be 64 mol% NaCl – 36 mol% PuCl3, however, the post-flow loop salt showed increased levels of MgCl2 and NaCl changing the salt composition to 10 mol% MgCl2 – 63mol% NaCl – 26mol% PuCl3) indicating the primary salt interacted with the rinse salt. Analysis of the post flow loop salt detected low concentrations of Al, Ni, Co, Nb, and Zr, most likely corrosion products from the flow loop material of construction. Contamination of the fuel salt (with the rinse salt NaCl-MgCl2) decreased the density by approximately 10% and reduced the onset of melting temperature by 50 °C, from 451 °C to approximately 400 °C. Results from the fission product simulated salt (61 mol% NaCl – 30 mol% UCl3 – 9mol% PuCl3) and the corrosion product salt (10 mol% MgCl2 – 63mol% NaCl – 26mol% PuCl3) will be included in two separate manuscripts for submission to peer-reviewed journals.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Advanced Simulation Aided Metallic Fuel Design

Nuclear engineering is reaching a domain space in which computational power and methods are starting to surpass the models that are used to simulate nuclear fuel rods. As these models are updated, simulations utilizing different compositions, operating conditions, fabrication techniques, and geometries can provide core designers with an important first step towards fuel qualification. Due to the cost associated with fuel irradiations, it can be expected that time and money spent building better fuel models may provide an attractive return during initial studies of advanced fuels, especially with regards to fuels beyond the extensively studied uranium oxide system. In general, the primary U.S. experience with irradiated fuel rods beyond traditional oxide fuel is with binary (U-Zr) and ternary (U-Pu-Zr) metal fuels. The bulk of the experimental data for metal fuels is obtained from rods irradiated in the EBR-II sodium fast reactor (SFR), thus many of the correlations developed for metal fuels comes from the post-irradiation (PIE) examination of EBR-II rods. One of the unique characteristics that set U-Pu-Zr fuels apart from traditional oxide fuels is the presence of different phases that develop in the fuel during irradiation. Each of these different phases has unique thermal, mechanical, and diffusive behavior. The transitions between these phases are not static as the initially homogenous zirconium diffuses towards the center and edge of the fuel rod, resulting in a low-Zr, high-U ring. Along with the differing material properties that occur in each phase, the increase in uranium corresponds to an increase in localized fission and power production. The direct coupling between the thermal and diffusion solves highlights one of the many examples of tightly coupled phenomenon that are present in nuclear fuel, and motivation for the development of codes that simultaneously solve each piece of physics in a fully-coupled manner.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Effect of Ion Irradiation on Amorphous and Crystalline Ge–Se and Their Application as Phase Change Temperature Sensor

Research on phase change materials is predominantly focused on their application as memory devices or for temperature control which requires low phase change temperature. The Ge–Se binary chalcogenide glass system with its wide glass-forming region is a potential candidate for high-temperature and high-radiation phase change applications. In this study, the concept of employing Ge x Se 100– x glasses to monitor high temperature (450–528 °C) using the phase change effect, is reported. Materials selection, device structure, and performance of prototype sensors are analyzed. In addition, the effect of heavy ion irradiation by Xe ions with energies of 200, 600, and 1000 keV (fluence ≈10 14 cm –2 ) on the Ge x Se 100– x ( x = 30, 33, 40) thin films and phase change devices is studied. The irradiation effect on the amorphous and crystalline structure of the thin films is evaluated by Raman spectroscopy and X-ray diffraction (XRD). Although the changes in the structural units of amorphous films are negligible, in crystalline films orthorhombic-GeSe 2 crystals are found to be most affected by irradiation and a new phase, orthorhombic GeSe is found in the thin films after irradiation. The performance of a sensor with an active film of Ge 40 Se 60 is also shown as an example.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Approaches for Water Removal in Direct-Fired sCO 2 Power Cycles

There is interest in investigation of water removal processes in direct fired sCO 2 flows, as this may potentially lead to greater system efficiency as the removal of this contaminant will result in sCO 2 behaving close to idealized behaviors. Water removal should be split into a two-step process, condensation of the water, followed by separation of the liquid phase water from the sCO 2 . The two main avenues of condensation are manipulation of pressure and temperature for phase change. For this paper, temperature-based phase change is the primary focus through the implementation of heat exchangers. Of the heat exchangers investigated it was found that printed circuit heat exchangers (PCHEs) could be an alternative for this use case, though the specific design of flow channel geometry and flow direction depends on the specific system case and cannot be determined at this point. For water separation there were four processes identified, all of which already assume water is in liquid phase at that point in the system. Of these separation avenues the best candidate is the hydrocyclone as it has a proven history of separating liquid-liquid phase mixtures with small density differences in oilfield use, in addition they have been investigated and modeled specifically for water separation for sCO 2 flows and the footprint is relatively small.

20 FOSSIL-FUELED POWER PLANTS↗