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Cobalt-base alloy which combines high temperature strength and magnetic properties has a composition in weight percent of 7-1/2 tungsten, 2-1/2 iron, 1 titanium, 1/2 zirconium, 1/2 carbon, and the balance cobalt. It may be used as construction material for electric motors and generators operating at high temperatures.
The Advanced Fuels Campaign Fission Accelerated Steady-state Test (FAST) at Idaho National Laboratory (INL) completed its first irradiation cycle within the Advanced Test Reactor (ATR). The test focused on the irradiation of alloy fuel forms for use in sodium fast reactors. Tests included the use of fuel additives (Sn, Sb, and Pd) for reducing the gettering of lanthanide fission products, zirconium liners to mitigate fuel-cladding chemical interaction, and annular fuel geometry for sodium-free fuel designs. The first cycle of FAST testing was completed and a series of rodlets were removed for the initial post irradiation examination (PIE) of FAST rodlets. The rodlet irradiation conditions was evaluated using Monte Carlo N-Particle (MCNP) for as-run power history and COMSOL for temperature analysis. These rodlets include a set of low burnups (~2.5 % heavy metal [%FIMA]) control rodlets and a helium bonded annular rodlet (4.7 %FIMA). Initial non-destructive PIE has been completed and includes, visual inspection, neutron radiography and gamma scanning of the FAST rodlets. Radiography confirmed the integrity of the experiments as well as showing the complete filling of the annulus in the annular fuel at a modest burnup (4.7 %FIMA) and potential slumping of the cooler rodlets at lower burnup. Precision gamma scanning indicated mostly usual fission product behavior with the exception of the ceasium in the He-bonded annular fuel. Future destructive PIE will be necessary in order to fully understand the effects of accelerated irradiation on U-Zr metallic fuel behavior.
The U.S. Nuclear Regulatory Commission (NRC) is anticipating licensing applications and commercial use of new fuel types in advanced nuclear power reactors that would be designed and built in the United States. Pacific Northwest National Laboratory (PNNL) is providing technical assistance to the NRC related to the newly proposed nuclear fuel and cladding designs that would be deployed in these reactors. This report focuses specifically on the metallic fuel that is being considered for sodium-cooled fast reactors and specifically on mechanisms that would cause damage or failure to the fuel under reactor operating conditions and design basis accident conditions. There is historic experience with both metallic and oxide fuels in experimental sodium-cooled fast reactors, but this report will focus solely on metallic fuel. Currently two U.S.-based reactor designers are engaged with the NRC in the application or pre-application review and considering a sodium-cooled fast reactor. TerraPower is engaged with NRC in pre-application review of its Natrium reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel clad in HT9 stainless steel. ARC Clean Technology is engaged with NRC in pre-application review of its ARC-100 reactor. The current design for this fuel uses a sodium-bonded uranium-10wt% zirconium (U10Zr) fuel with steel cladding. This report will focus on this fuel system specifically, with broader information given regarding other metallic fuel systems with other stainless steel alloy claddings. To support the NRC’s readiness efforts, this report will identify and discuss degradation and failure modes of metallic fuel concepts for sodium-cooled fast reactors, including fuel performance characteristics that may not be addressed within existing regulatory documents.
Tensile stress effect on critical current-carrying capacity of niobium-zirconium wire and niobium-tin ribbon superconductors at liquid helium temperature
Emittance and thermal cycling tests of stainless steel tubes coated with iron and calcium titanate, and niobium-zirconium tubes coated with iron titanate
Advance high density monolithic alloy HALEU fuels have been scanned with a silicon drift detector scanning system to measure low-energy characteristic X-rays emitted by uranium daughters, as well as zirconium present as a foil diffusion barrier. Four peaks were measured across the surfaces of three plates, and are consistent with other thorium and protactinium peaks present in depleted uranium samples that are induced by the decay of 235 U and 238 U. Of the four peaks detected, three are dominated by the decay of 235 U, and one peak at 15.70 keV covers two characteristic X-rays from the zirconium diffusion barrier (Kα 1 ) and from a 231 Th 15.62 keV X-ray induced by 235 U α-decay. The low energies of these lines suggest that they are mostly emitted from the fuel meat surface, and the poor counting statistics caused by strong self-attenuation make these data a poor candidate for 235 U/ 238 U ratio analysis. We suggest, then, that these X-rays could be used for passive measurement of variations in the cladding thickness of these fuels to complement manual thickness measurement methods.
This research effort is aimed at providing further insight into this multi-variable dependent phenomena by looking at the effects of gravity on the ignition and combustion behavior of metals. Since spacecraft are subjected to higher-than-1g gravity loads during launch and reentry and to zero-gravity environments while in orbit, the study of ignition and combustion of bulk metals at different gravitational potentials is of great practical concern. From the scientific standpoint, studies conducted under microgravity conditions provide simplified boundary conditions since buoyancy is removed, and make possible the identification of fundamental ignition mechanisms. The effect of microgravity on the combustion of bulk metals has been investigated by Steinberg, et al. on a drop tower simulator. However, no detailed quantitative work has been done on ignition phenomena of bulk metals at lower or higher-than-normal gravitational fields or on the combustion characteristics of metals at elevated gravity. The primary objective of this investigation is the development of an experimental system capable of providing fundamental physical and chemical information on the ignition of bulk metals under different gravity levels. The metals used in the study, iron (Fe), titanium (Ti), zirconium (Zr), magnesium (Mg), zinc (Zn), and copper (Cu) were selected because of their importance as elements of structural metals and their simple chemical composition (pure metals instead of multi-component alloys to avoid complication in morphology and spectroscopic studies). These samples were also chosen to study the two different combustion modes experienced by metals: heterogeneous or surface oxidation, and homogeneous or gas-phase reaction. The experimental approach provides surface temperature profiles, spectroscopic measurements, surface morphology, x-ray spectrometry of metals specimens and their combustion products, and high-speed cinematography of the heating, ignition and combustion stages of the metal specimen. This paper summarizes the results obtained to date from experiments conducted under normal and high-gravity conditions.
Uranium–molybdenum (U–Mo) alloys in monolithic fuel foil are the primary candidate for the conversion of high-performance research reactors in the USA. Monolithic fuel is utilized in a plate-type design with a zirconium diffusion barrier and aluminum cladding. These fuel types are unique in that they contain no plenum for the release of fission gases, which, in conjunction with the aluminum cladding, can lead to large stress states within the fuel. The nature of how fundamental processes of radiation damage, including the evolution of point defects, under such stresses occur is unknown. In this work, we present molecular dynamics simulations of the formation energy of point defects under applied stress. We report this work will allow for the implementation of stress-dependent microstructural evolution models of nuclear fuels, including those for both fission gas bubble growth and creep, which are critical to ensure the stable and predictable behavior of research reactor fuels.
We report alpha-zirconium (α-Zr) is a hydrogen atom host material used in nuclear energy and semiconducting applications due to Zr's excellent properties such as corrosion resistance, low thermal neutrons cross-section, and thermodynamically favorable interstitial sequestering of H atoms. Herein we use density functional theory simulations to systematically analyze the initial stages of H insertion in the Zr matrix, and determine optimal hydrogen localization, distribution, and clustering tendencies from a thermodynamics and kinetics perspective. Our results suggest that H occupation of tetrahedral sites is strongly favored, consistent with previous experimental and theoretical studies. Further, a cooperative effect in H localization is shown where H atoms favor clustering. The H pair is found to be the most stable by occupying tetrahedral sites with pairs at second nearest-neighbor positions (TTn2). Among the four H atom clusters, the most favorable configuration has two TTn2 pairs forming "box " or "zig-zag" shaped-cluster with the pairs separated by 3.03 or 3.25 Å, respectively. Nudged elastic band simulations suggest that the formation of TTn2 is facile and is limited by the small diffusion barrier of the monomer of less than 0.1 eV. Further, the formation routes for the 4H atom clusters are also presented. The current study can serve as a foundation to systematically model H atom clustering in metals and alloys.
Here, this work demonstrates how first-principles statistical mechanics approaches within a Bayesian framework can quantify and propagate uncertainties to downstream thermodynamic calculations. To address the issue of Bayesian prior selection, knowledge of 0 K ground states in the material system of interest is incorporated into the prior. The effectiveness of this framework is shown by creating a phase diagram for the fcc zirconium nitride system, including confidence intervals on order-disorder transition temperatures.
Zirconium carbide (ZrC) is a candidate material for use in advanced high temperature reactors, including space nuclear thermal propulsion applications. Thermal neutron scattering laws (TSLs) are generated for carbon bound in ZrC, C(ZrC), and zirconium bound in ZrC, Zr(ZrC), using ab initio lattice dynamics methods. These evaluations are to be submitted for inclusion in ENDF/B-VIII.1 and use the incoherent approximation for inelastic scattering as well as the new mixed elastic scattering treatment. The application of disordered alloy theory is introduced to appropriately capture the isotopic composition of Zr and C in the elastic scattering cross section. Localized higher energy vibrations in the C(ZrC) phonon density of states that are separated from lower energy modes result in quantized oscillations in the inelastic contributions to the TSL with a significant likelihood of large energy down-scattering and up-scattering interactions, where the latter increases in probability with temperature. The quanta of energy transfer during neutron thermalization is substantially greater than classically expected within the thermal neutron energy range. MC21 critical mass calculations of ZrC mixtures with high-enriched uranium demonstrate an impact of the TSLs when compared to free-gas treatment for 235 U concentrations less than 0.2 g/cm 3 . Additional MC21 critical mass calculations with homogenous mixtures of ZrC and reactor-grade graphite also demonstrate sensitivity to the ZrC TSL for thermal spectrum driven fission systems.
A high temperature oxidation resistant, thermal barrier coating system is disclosed for a nickel cobalt, or iron base alloy substrate. An inner metal bond coating contacts the substrate, and a thermal barrier coating covers the bond coating. NiCrAlR, FeCrAlR, and CoCrAlR alloys are satisfactory as bond coating compositions where R=Y or Yb. These alloys contain, by weight, 24.9-36.7% chromium, 5.4-18.5% aluminum, and 0.05 to 1.55% yttrium or 0.05 to 0.53% ytterbium. The coatings containing ytterbium are preferred over those containing yttrium. An outer thermal barrier coating of partial stabilized zirconium oxide (zirconia) which is between 6% and 8%, by weight, of yttrium oxide (yttria) covers the bond coating. Partial stabilization provides a material with superior durability. Partially stabilized zirconia consists of mixtures of cubic, tetragonal, and monoclinic phases.
High temperature research in vacuum - magnesium oxide and zirconium dioxide interaction with refractory metal carbides, nitrates, and borides
Improved x-ray source for x-ray photoelectron spectroscopy (XPS) contains both zirconium anode with beryllium window and magnesium anode with aluminum window. Previously unresolvable peaks of electron-energy spectrum become resolvable. Developed specifically for use in analyzing distributions of chemical constituents in surface layers of specimens of 2219 aluminum alloy and in determining the depths of surface oxide layers and relative proportions of aluminum and oxide in layers. Also used to study chemical constituents of surface layers in other material systems - for example, thin oxide films on silicon-based semiconductor devices, oxide films on alloys, and surface layers affecting adhesion of paints or bonding materials.
Microstructural evolution was studied in samples of wrought and vacuum plasma sprayed (VPS) NARloy-Z exposed to temperatures up to 970 C (1,780 F) for up to 60 h. Samples were heated in a vacuum furnace, followed by rapid quenching in helium (He) gas at a cooling rate of 166 C (300 F) per second. Microstructural analyses were conducted using optical microscopy, scanning electron microscopy (SEM), and electron probe microanalysis (EPMA). In both the wrought and VPS conditions, precipitates rich in silver (Ag) and zirconium (Zr) were present in the matrix and at the grain boundaries even after long exposure to elevated temperatures. Islands rich in oxygen (O2) and Zr were also observed, as well as incipient melting at the grain boundary triple points. Results indicated that the alloy cannot be homogenized by heat treatment at elevated temperatures.
To increase the power cycle efficiency and lower the levelized cost of electricity (LCOE) of concentrated solar power (CSP) plants, printed circuit heat exchangers (PCHEs) capable of operating above 700 °C with molten chloride salt and a sCO 2 -based fluid are needed. In this paper, the design of a high-pressure, high-temperature, 2 MW PCHE comprised of a thermomechanically-robust, zirconium carbide/tungsten (ZrC/W) composite is conducted for CSP plants. The ZrC/W composite is a material with high thermal conductivity, high stiffness, and high failure strength at high temperatures, along with excellent resistance to thermal cycling and thermal shock. In this work, a thermomechanical design analysis was conducted to select appropriate material thicknesses of the ZrC/W plates, and to determine the geometrical dimensions and the thermal performance of the PCHE. The influences of the plate number and heat exchanger length on power density and pressure drop have also been systematically investigated. Economic analyses were conducted to compare the cost of ZrC/W-based PCHEs to those comprised of IN740H (a state-of-the-art, nickel-based superalloy) and 316 stainless steel (316SS). Finally, at a sufficiently high plate production rate, the manufacturing cost of ZrC/W-based PCHEs can be significantly lower, while achieving a much higher power density, compared with state-of-the-art, nickel alloy-based and stainless steel-based PCHEs.
The NASA Marshall Space Flight Center (MSFC) Electrostatic Levitator (ESL) Laboratory is a unique facility for investigators studying high-temperature materials. The laboratory boasts two levitators in which samples can be levitated, heated, melted, undercooled, and resolidified, all without the interference of a container or data-gathering instrument. The ESL main chamber has been upgraded with the addition of a rapid quench system. This system allows samples to be dropped into a quench vessel that can be filled with a low melting point material, such as a gallium or indium alloy. Thereby allowing rapid quenching of undercooled liquid metals. Up to 8 quench vessels can be loaded into the quench wheel, which is indexed with LabVIEW control software. This allows up to 8 samples to be rapidly quenched before having to open the chamber. The system has been tested successfully on several zirconium samples. Future work will be done with other materials using different quench mediums. Microstructural analysis will also be done on successfully quench samples.
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