Fusion Plasma Facing Research Articles

Numerical Study of Heat Transport in Static Liquid Metal Exposed to Plasma with Magnetic Field

Suitable management of excessive heat in liquid metal is one of key roles to accomplish the proper liquid metal usage as future alternative fusion plasma facing components. In this work, theoretical heat transfer in liquid metal in contact with magnetized plasma was carried out, and a numerical solver for investigating temperature and induced velocity by j×B force has been developed. The study suggests the trends: 1. lower bulk temperature in liquid metal at which magnetic field is presented, compared to no magnetic field, by additional convection from j × B, stronger than natural convection; and 2. asymmetric temperature distribution in liquid metal along j × B.

Irradiation embrittlement in pure chromium and chromium-tungsten alloy in a view of their potential application for fusion plasma facing components

Design of plasma-facing components (PFC) for DEMO divertor unravels new challenges to be met by the in-vessel materials. Embrittlement induced by 14 MeV neutrons in the baseline first wall material - tungsten (W) endangers structural integrity of PFCs. Chromium (Cr) and/or Cr-W alloy has been considered as a candidate material in the design of mid heat flux PFCs as structural body of the monoblock. In our preceding work, the potential of the vacuum arc melting for the fabrication of pure Cr and Cr-W solid solution (with 10 at.%) has been demonstrated. The pure Cr exhibits ductile deformation above 50°C, while the ductile to brittle transition in Cr-10%W is at ~300°C. The particular advantage of the alloying was seen in the strong increase of the work hardening capacity as well as much higher fracture strength compared to pure Cr. Moreover, the solid solution formation as a result of alloying might improve the resistance of the structural materials against irradiation, as was shown for the case of iron-chromium alloys/steels. In this work, the mechanical properties of two pure Cr grades and Cr-10at.%W alloy are assessed after neutron irradiation performed in the conditions relevant for the application of PFCs. The mechanical tests are performed over a wide temperature range to investigate the strength at the irradiation temperature as well as to clarify the role of limited defect annealing. The neutron irradiation resulted in remarkable degradation of mechanical properties making all the studied materials brittle at the applied irradiation temperature. In the case of Cr-10at.%W, the irradiation caused such severe modifications that the fracture strength has become lower than in non-irradiated state. The fracture surface analysis is made to demonstrate the nature of the brittle fracture and to discuss possible reasons for the observed brittleness.

Monte Carlo simulations of chromium target under proton irradiation of 17.9, 22.3 MeV

Chromium material is commonly used for fusion plasma facing applications because of the low neutron activation property. The Monte Carlo method is one of the useful ways to investigate the ion-target interactions. In this study, Chromium target irradiated by protons was investigated using Monte Carlo based simulation tools. In this context, the calculations of radiation damage on Chromium material irradiated with protons at 17.9 and 22.3 MeV energies were carried out using GEANT4 and SRIM codes. Besides, the cross sections for proton interaction with Chromium target were calculated by the TALYS 1.9 code using CTM + FGM, BSFGM, and GSFM level densities. As a result, GEANT4, SRIM and TALYS 1.9 codes provide a suitable tool for the predictions of radiation damage and cross cross section with proton irradiation.

An ab-initio study of hydrogen trapping energetics at BCC tungsten metal-noble gas interfaces

Density functional theory (DFT) calculations have been performed to assess the trapping and segregation strength of hydrogen (H) to noble gas interfaces in tungsten (W). These calculations consist of a supercell containing a slab of BCC W and an initial lattice of FCC noble gas atoms. The interfaces included noble gases of helium, neon, and argon, with densities in the range of 1-4 atoms-per-vacancy (V), and W surface orientations of (100), (110), and (111). We report on the binding energy of H to these interfaces as well as the modification to the migration barriers in the W slab, which together provide information on the segregation strength and de-trapping energy for H at noble gas bubbles. These calculations indicate that the binding energy of H to W-noble gas interfaces varies with surface orientation and decreases with increasing gas density; whereas the H migration energy is sensitive to the noble gas density, surface orientation, and diffusion pathway, and typically increases with gas density. Together, the de-trapping energy of H to these interfaces is shown to depend less significantly on noble gas density. These DFT calculations provide valuable first-principles energetics necessary for mesoscale models, and provide insight into the temperatures required to de-trap tritium from He bubbles in fusion plasma facing components.

Helium ions irradiation analysis of W0.5(TaTiVCr)0.5 for application as a future fusion plasma-facing material

The irradiation resistance of a reduced activation tungsten-containing high-entropy alloy-based material W0.5(TaTiVCr)0.5 was analyzed using He ions irradiation. Mirror-polished samples were exposed to 200 keV He+ ions at a maximum fluence of 1 × 1021 ions/m2. The nano-indentation hardness of W0.5(TaTiVCr)0.5 increased from 11.6 GPa to 14.9 GPa (at 200 nm depth) following He ion irradiation. Examination of the sample's internal microstructure, using transmission electron microscopy, revealed some randomly dispersed black spots and precursors of He bubbles and dislocations causing irradiation hardening. Relatively better irradiation resistance of W0.5(TaTiVCr)0.5 compared with W, under the given experimental conditions suggests its potential fusion plasma facing applications.

Recent accomplishments of the fusion safety program at the Idaho National Laboratory

The Idaho National Laboratory (INL) Fusion Safety Program (FSP) is the Office of Fusion Energy Sciences’ (FES) lead laboratory for Magnetic Fusion Energy (MFE) Safety. Our mission is to assist the US and international fusion communities in developing the inherent safety and environmental potential of fusion power by: 1) Developing fusion licensing data and analysis tools, 2) Participating in national and international collaborations and design studies, and 3) Assisting the US and international fusion community in licensing activities and guidance in operational safety. To achieve our mission, the FSP maintains core competencies in the several areas, two of which are: fusion safety code development and tritium retention and permeation in fusion plasma facing component (PFC) and blanket materials. This article details recent accomplishments of our program in these areas and future directions for fusion safety research and development at INL.

Fabrication of Tungsten-Vanadium Hybrid Material with Sufficient Toughness for High-Temperature Applications by Diffusion Bonding

The design of fusion plasma-facing components is challenging, as their materials have to meet rigorous requirements in terms of low activation and high-temperature strength. At the same time, sufficient ductility is required even in the low-temperature range. Unfortunately, these properties are not found in conventional materials. To solve this problem, a hybrid material that combines the high strength of one material with the high ductility of the other material was developed. This paper presents the hybrid material, which consists of thin tungsten and vanadium layers. This hybrid material was produced by means of diffusion bonding at relatively low temperature in a vacuum chamber. Microstructural investigations and nanoindentation tests indicated no cracks, no delamination, and no brittle intermetallic phases along the bond interfaces. Investigations of the mechanical properties of the hybrid material by instrumented Charpy impact tests revealed a relatively low ductile-to-brittle transition temperature (DBTT) at 124°C (compared to the DBTT of polycrystalline tungsten of >441°C) with an absorbed Charpy impact energy of 4.53 J [kleinst (KLST)-specimen]. Additionally, the tested Charpy impact specimens were found to be not fractured thoroughly even at room temperature.

Interfacial microstructures and hardness distributions of vacuum plasma spraying W-coated ODS ferritic steels for fusion plasma facing applications

In the present study, interfacial microstructures and hardness distributions of W-coated ODS steels as plasma facing structural materials were investigated. A vacuum plasma spraying (VPS) technique was employed to fabricate a W layer on the surface of the ODS ferritic steel substrates. The microstructural observations revealed that the VPS-W has very fine grains aligned toward the spraying direction, and a favorable interface between W and ODS ferritic steels by a mechanical inter-locking without an intermetallic layer. However, crack-type defects were found in VPS-W. Because a brittle inter-diffused layer does not exist at the joint interface, the hardness was gradually distributed in the joint region. After neutron irradiation, irradiation hardening significantly occurred in the VPS-W. However, the hardening of VPS-W was less than that of bulk W irradiated at 773K. Thus, the VPS is considered to be one of the promising ways to join dissimilar materials between W and ODS steels, which can avoid the formation of an interfacial intermetallic layer and create favorable irradiation hardening resistance on the W coated layer.

Nanocrystalline diamond coating of fusion plasma facing components

Nanocrystalline diamond layers were deposited by hot-filament CVD on several substrates including molybdenum tiles and foils, graphite blocks and silicon wafers. The structural characterisation of the deposited layers, achieved by SEM, AFM, Raman and X-ray photoelectron spectroscopies, showed the typical features of nanocrystalline diamond. A nanocrystalline diamond coated molybdenum tile was exposed to a fusion grade deuterium plasma in the MAST tokamak, with a plasma current of 700 kA, core electron temperature of about 0.6 keV, and total exposure time of about 1 s. The diamond coating showed good resistance to the extremely harsh conditions of the fusion plasma edge.

High flux ion beam acceleration at the 100-eV level for fusion plasma facing material studies

For the investigation of interactions between the low energy ions in fusion edge plasmas and plasma facing materials, new electrodes for beam extraction have been developed to achieve a 100-eV level, high-flux and steady state operation with large irradiation area in the super low energy ion source (SLEIS) facility. The structure of these electrodes has a multi-aperture triode electrodes, whose size is smaller than conventional ones. Namely, these electrodes with the thickness of 0.5 mm have a multi-aperture of 0.9 mm in diameter. The gap distance between electrodes is 0.5 mm. Ion fluxes have been achieved from 1.7 × 10 20 to 5.3 × 10 20 H/m 2 s with the energy from 32 to 102 eV/H. This paper also presents a study about the temperature dependence of blister formation of tungsten by using this new hydrogen ion beam.

Vacuum Arc Deposited Boron Carbide Films for Fusion Plasma Facing Components

Boron carbide is an ideal coating for radio-frequency antennas in magnetic fusion energy, due to a combination of desirable properties: high hardness at high temperature, high melting point, low Z and high thermal conductivity. In this paper, the feasibility of using vacuum arc technology for coating antennas and other magnetic fusion energy plasma facing components is explored. This technique has the potential of producing much denser film than plasma spray and substantially higher deposition rates than magnetron sputtering. In addition, the use of hyper-thermal species may result in the formation of high thermal conductivity crystalline phase at lower deposition temperatures than would otherwise be expected. Finally, the compatibility of the vacuum arc with ultra-high vacuum conditions raises the possibility of in situ repair of components in a fusion reactor. Initial deposition studies are presented, which produced primarily amorphous film, but with the correct stoichiometry and a high deposition rate (>10nm/s). The properties of this film are presented in this paper. Some of the properties of the vacuum arc discharge, the first to be operated successfully with a sintered boron carbide cathode, are also presented.