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Viscosity and normal density measurements at lambda point intervals for pure helium-4 and helium-3 mixtures with helium-4
We report results of object kinetic Monte Carlo (OKMC) simulations aimed at understanding the effect of helium flux on the near-surface helium accumulation in plasma-facing tungsten, which is initially defect-free and has a W(100) surface orientation. These OKMC simulations are performed at 933 K for fluxes ranging from 10 22 to 4 × 10 25 He/m 2 s, with 100 eV helium atoms impinging on a W(100) surface up to a maximum fluence of 4×10 19 He/m 2 . In the near-surface region, helium clusters interact elastically with the free surface. The interaction is attractive and results in the drift of mobile helium clusters towards the surface as well as increased trap mutation rates. The associated kinetics and energetics of the above-mentioned processes obtained from molecular dynamics simulations are also considered. The OKMC simulations indicate that as the flux decreases, the retention of implanted helium decreases, and its depth distribution shifts to deeper below the surface in initially pristine tungsten. Furthermore, the fraction of retained helium diffusing into the bulk increases as well, so much so that for 10 22 He/m 2 s, almost all of the retained helium diffused into the bulk with minimal/negligible near-surface helium accumulation. At a given flux, with increasing fluence, the fraction of retained helium initially decreases and then starts to increase after reaching a minimum. The occurrence of the retention minimum shifts to higher fluences as the flux decreases. Although the near-surface helium accumulation spreads deeper into the material with decreasing flux and increasing fluence, the spread appears to saturate at depths between 80 and 100 nm. Finally, we present a detailed analysis of the influence of helium flux on the size and depth distribution of total helium and helium bubbles.
The synergism of transmutant helium and irradiation defects is critically important to understand the performance of SiC when in service in nuclear environments. In this study, we exposed 3C-SiC in contact with a 2 μm Ni foil to neutron irradiation at 500 °C to 29 dpa in the high flux isotope reactor to simulate the simultaneous introduction of helium and irradiation defects into SiC. The combination of TEM observations and thermal desorption measurements helps to elucidate the impact of helium on the microstructural evolution in 3C-SiC. Thermal helium desorption spectra from ion-implanted and neutron-irradiated SiC showed a completely different desorption behavior in terms of the peak position and helium desorption flux. The identification of possible helium trapping sites was attempted by applying the first order dissociation model to the measured thermal helium desorption spectra. TEM observations of the neutron-irradiated samples showed that the presence of helium facilitated the stabilization of the defect clusters and promoted the formation of visible helium bubbles in SiC subject to simultaneous neutron irradiation and in-situ He implantation. Following the subsequent heat treatment during the thermal desorption measurements, large faceted helium bubbles were found along grain boundaries in the neutron-irradiated SiC with in-situ He implantation, while the helium bubbles in the grain interior were relatively smaller than those found in the neutron-irradiated sample without in-situ He implantation due to the reduced mobility of He-defect clusters. In conclusion, the He contents contained in the neutron-irradiated samples were quantified based on the experimental data from TEM and thermal desorption measurements.
Helium has profound effects on the microstructure and mechanical property of materials used in nuclear power systems. Here to understand the interaction between helium and irradiation-induced defects and its effect on bubble formation, single-beam and dual-beam irradiations on nickel were performed with 16 keV helium ions and 1 MeV krypton ions under in-situ TEM observation. For 1 MeV krypton single-beam irradiation, voids were observed at 600ºC but not at 500ºC and 700ºC. For helium single-beam irradiation at 500ºC, helium moved freely and formed bubble-loop complexes. For dual-beam irradiation at 500ºC, helium was trapped by vacancies and formed bubbles homogeneously. The structure of dislocation loops was also affected by the presence of helium. For krypton single-beam irradiation at 500ºC, the microstructure was {100} periodic wall of faulted and perfect dislocation loops. For dual-beam irradiation at 500ºC, periodic wall was not observed, and dislocation loops were predominantly perfect. For single-beam helium at 500ºC, dislocation loops were produced by athermal process of SIAs ejection or loop punching mechanisms. The interplay between helium and vacancies played an important role on the evolution of helium bubbles and defect microstructures.
Helium gas accumulation from alpha decay during extended storage of spent fuel has potential to compromise the structural integrity the fuel. We report results obtained with surrogate nickel particles which suggest that alumina formed by atomic layer deposition can serve as a low volume-fraction, uniformly-distributed phase for retention of helium generated in fuel particles such as uranium oxide. Thin alumina layers may also form transport paths for helium in the fuel rod, which would otherwise be impermeable. Micron-scale nickel particles, representative of uranium oxide particles in their low helium solubility and compatibility with the alumina synthesis process, were homogeneously coated with alumina approximately 3–20 nm by particle atomic layer deposition (ALD) using a fluidized bed reactor. Particles were then loaded with helium at 800 °C in a tube furnace. Subsequent helium spectroscopy measurements showed that the alumina phase, or more likely a related nickel/alumina interface structure, retains helium at a density of at least 10 17 atoms/cm 3 . High resolution transmission electron microscopy revealed that the thermal treatment increased the alumina thickness and generated additional porosity. Results from Monte Carlo simulations on amorphous alumina predict the helium retention concentration at room temperature could reach 10 21 atoms/cm 3 at 400 MPa, a pressure predicted by others to be developed in uranium oxide without an alumina secondary phase. This concentration is sufficient to eliminate bubble formation in the nuclear fuel for long-term storage scenarios, for example. Finally, measurements by others of the diffusion coefficient in polycrystalline alumina indicate values several orders of magnitude higher than in uranium oxide, which then can also allow for helium transport out of the spent fuel.
Abstract not provided.
Transition-metal ultra-high temperature ceramics are promising materials for nuclear structural applications. However, an understanding of their response to high-temperature irradiation and helium is vastly limited. This paper presents a study of helium effects in zirconium carbide (ZrC) by performing room temperature 3 MeV 3 He + ion irradiations up to 5 × 10 20 ions m –2 (~1.8 at.% at peak) and high-temperature annealing (1273–1873 K), coupled with state-of-art characterization using transmission electron microscopy (TEM), scanning electron microscopy (SEM) and nuclear reaction analysis (NRA). We reveal that ZrC is susceptible to irradiation damage in terms of helium bubble formation. After annealing at 1373 K, tiny bubbles (1–2 nm) formed aligned clusters which were highly over-pressurized, producing strain contrast in TEM. At 1773 K, significant bubble growth occurred. Additionally, at 1773 K, a combined TEM/SEM analysis revealed dramatic matrix damage due to helium-induced surface blistering. Underneath blister caps, the microstructure evolved into ultra-fine nano-scale grains similar to high burn-up structures observed in nuclear fuels, but consisting of numerous nano-cracks. We hypothesize that such structures are formed due to high gas pressure build-up and its subsequent release. This phenomenon initiated at the grain boundaries. Blister top surface consisted of inter-granular and trans-granular cracks. NRA depth profiling revealed that helium was present as double peaks with major portion lying at the end-of-the-range (EOR) and the rest as TEM invisible clusters in a shoulder extending to the surface. ZrC started releasing helium after 1373 K. As a result, helium release increased significantly at higher temperatures, with majority helium loss occurring from EOR rather than from near-surface regions.
A cryo-quenched 70 wt % Fe-15 wt% Cr-15 wt% Ni single-crystal alloy with fcc (face centered cubic), bcc (body centered cubic), and hcp (hexagonal close packed) phases was implanted with 200 keV He+ ions up to 2 × 1017 ions·cm -2 at 773 K. Surface-relief features were observed subsequent to the He+ ion implantation, and transmission electron microscopy was used to characterize both the surface relief properties and the details of associated “swelling effects” arising cumulatively from the austenitic-to-martensitic phase transformation and helium ion-induced bubble evolution in the single-crystal ternary alloy. The bubble size in the bcc phase was found to be larger than that in the fcc phase, while the bubble density in the bcc phase was correspondingly lower. The phase boundaries with misfit dislocations formed during the martensitic transformation and reversion processes served as helium traps that dispersed the helium bubble distribution. Swelling caused by the phase transformation in the alloy was dominant compared to that caused by helium bubble formation due to the limited depth of the helium ion implantation. The detailed morphology of helium bubbles formed in the bcc, hcp, and fcc phases were compared and correlated with the characters of each phase. The helium diffusion coefficient under irradiation at 773 K in the bcc phase was much higher (i.e., by several orders of magnitude) than that in the fcc phase and led to faster bubble growth. Moreover, the misfit phase boundaries were shown to be effective sites for the diffusion of helium atoms. This feature may be considered to be a desirable property for improving the radiation tolerance of the subject, ternary alloy.
Helium evolution from the transfer of helium-saturated propellant in space is quantified to assess its impacts from creating two-phase gas/liquid flow from the supply tank, gas injection into the receiving tank, and liquid discharge from the receiving tank. Propellant transfer takes place between two similar tanks whose maximum storage capacity is approximately 2.55 cubic meters each. The maximum on-orbit propellants transfer capability is 9000 lbm (fuel and oxidizer). The transfer line is approximately 1.27 cm in diameter and 6096 cm in length and comprised of the fluid interconnect system (FICS), the orbiter propellant transfer system (OPTS), and the International Space Station (ISS) propulsion module (ISSPM). The propellant transfer rate begins at approximately 11 liter per minute (lpm) and subsequently drops to approximately 0.5 lpm. The tank nominal operating pressure is approximately 1827 kPa (absolute). The line pressure drops for Monomethy1hydrazine (MMH) and Nitrogen tetroxide (NTO) at 11.3 lpm are approximately 202 kPa and 302 kPa, respectively. The pressure-drop results are based on a single-phase flow. The receiving tank is required to vent from approximately 1827 kPa to a lower pressure to affect propellant transfer. These pressure-drop scenarios cause the helium-saturated propellants to release excess helium. For tank ullage venting, the maximum volumes of helium evolved at tank pressure are approximately 0.5 ft3 for MMH and 2 ft3 for NTO. In microgravity environment, due to lack of body force, the helium evolution from a liquid body acts to propel it, which influences its fluid dynamics. For propellant transfer, the volume fractions of helium evolved at line pressure are 0.1% by volume for MMH and 0.6 % by volume for NTO at 11.3 lpm. The void fraction of helium evolved varies as an approximate second order power function of flow rate.
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The use of ultrafine and nanocrystalline materials is a proposed pathway to mitigate irradiation damage in nuclear fusion components. Here, we examine the radiation tolerance of helium bubble formation in 85 nm (average grain size) nanocrystalline-equiaxed-grained tungsten and an ultrafine tungsten-TiC alloy under extreme low energy helium implantation at 1223 K via in-situ transmission electron microscope (TEM). Helium bubble damage evolution in terms of number density, size, and total volume contribution to grain matrices has been determined as a function of He+ implantation fluence. The outputs were compared to previously published results on severe plastically deformed (SPD) tungsten implanted under the same conditions. Large helium bubbles were formed on the grain boundaries and helium bubble damage evolution profiles are shown to differ among the different materials with less overall damage in the nanocrystalline tungsten. Compared to previous works, the results in this work indicate that the nanocrystalline tungsten should possess a fuzz formation threshold more than one order of magnitude higher than coarse-grained tungsten.
Interactions of satellite-speed helium atoms with practical satellite surfaces were investigated experimentally, and spatial distributions of satellite-speed helium beams scattered from four different engineering surfaces were measured. The 7000 m/sec helium beams were produced using an arc-heated supersonic molecular beam source. The test surfaces included cleaned 6061-T6 aluminum plate, anodized aluminum foil, white paint, and quartz surfaces. Both in-plane (in the plane containing the incident beam and the surface normal) and out-of-plane spatial distributions of reflected helium atoms were measured for six different incidence angles (0, 15, 30, 45, 60, and 75 deg from the surface normal). It was found that a large fraction of the incident helium atoms were scattered back in the vicinity of the incoming beam, particularly in the case of glancing incidence angles. This unexpected scattering feature results perhaps from the gross roughness of these test surfaces. This prominent backscattering could yield drag coefficients which are higher than for surfaces with either forward-lobed or diffusive (cosine) scattering patterns.
The paper describes a helium heater to be used to conduct non-nuclear demonstration tests of the complete power conversion loop for a direct-cycle gas-cooled nuclear reactor power plant. Requirements for the heater include: heating the helium to a 1500 F temperature, operating at a 1000 psia helium pressure, providing a thermal response capability and helium volume similar to that of the nuclear reactor, and a total heater system helium pressure drop of not more than 15 psi. The unique compact heater system design proposed consists of 18 heater modules; air preheaters, compressors, and compressor drive systems; an integral control system; piping; and auxiliary equipment. The heater modules incorporate the dual-concentric-tube 'Variflux' heat exchanger design which provides a controlled heat flux along the entire length of the tube element. The heater design as proposed will meet all system requirements. The heater uses pressurized combustion (50 psia) to provide intensive heat transfer, and to minimize furnace volume and heat storage mass.
The dynamics of helium shell convection driven by nuclear burning establish the conditions for runaway in the sub-Chandrasekhar-mass, double-detonation model for SNe Ia, as well as for a variety of other explosive phenomena. We explore these convection dynamics for a range of white dwarf core and helium shell masses in three dimensions using the low Mach number hydrodynamics code MAESTRO. We present calculations of the bulk properties of this evolution, including time-series evolution of global diagnostics, lateral averages of the 3D state, and the global 3D state. We find a variety of outcomes, including quasi-equilibrium, localized runaway, and convective runaway. Our results suggest that the double-detonation progenitor model is promising and that 3D dynamic convection plays a key role.
Austenitic stainless steels are the standard materials for containment of hydrogen and tritium because of their resistance to mechanical property degradation in those environments. The mechanical performance of the primary containment material is critical for tritium handling, processing, and storage, thus comprehensive understanding of the processes of tritium embrittlement is an enabling capability for fusion energy. This work describes the investigation of the effects of low levels of tritium-decay-helium ingrowth on 304 L tubes. Long-term aging with tritium leads to high helium contents in austenitic stainless steels and can reduce fracture toughness by 95 %, but the details of behavior at low helium contents are not as well characterized. Here, we present results from tensile testing of tritium pre-charged 304 L tube specimens with a variety of starting microstructures that all contain a low level of helium. The results of the tritium exposed-and-aged materials are compared to previously reported results on similar specimens tested in an unexposed condition as well as the hydrogen precharged condition. Tritium precharging and aging for a short duration resulted in increased yield strengths, ultimate tensile strengths and slightly increased elongation to failure, comparable to higher concentrations of hydrogen precharging.
Helium is generally understood to be chemically inert and this is due to its extremely stable closed-shell electronic configuration, zero electron affinity and an unsurpassed ionization potential. It is not known to form thermodynamically stable compounds, except a few inclusion compounds. Here, using the ab initio evolutionary algorithm USPEX and subsequent high-pressure synthesis in a diamond anvil cell, we report the discovery of a thermodynamically stable compound of helium and sodium, Na 2 He, which has a fluorite-type structure and is stable at pressures >113 GPa. We show that the presence of He atoms causes strong electron localization and makes this material insulating. This phase is an electride, with electron pairs localized in interstices, forming eight-centre two-electron bonds within empty Na 8 cubes. As a result, we also predict the existence of Na 2 HeO with a similar structure at pressures above 15 GPa.