Fusion Reactor Components Research Articles

Validation of cyanide copper electrodeposited layer on test coupons for anti-seizing and outgassing in Tokamak vacuum vessel

A vacuum vessel which is the first confinement barrier of Tokamak fusion reactor should have numerous in-vessel components. Since those components are to be exposed to the plasma under ultrahigh vacuum condition and should be maintained during Tokamak operation, a threaded connection of vacuum vessel and in-vessel components is necessary. The threaded connection of stainless steel with the components causes difficulties during assembly or disassembly due to the large friction coefficient of stainless steel. To resolve this issue, anti-seize coating that does not affect the mechanical property, thermal stability, and outgassing rate of the threaded insert needs to be developed. In this study, adhesion, porosity, outgassing, and friction wear tests for a cyanide copper electrodeposited surface on 316 L austenitic stainless steel test coupons were performed to validate the high level of confidence in the performance of a cyanide copper electrodeposited threaded insert. The validation tests results showed that cyanide copper electrodeposition for anti-seizing could be applied to the 316 L stainless steel threaded insert in the vacuum vessel and in-vessel components of the Tokamak fusion reactor.

Characterising the impact of castellations on the efficiency of induction heating during testing in the HIVE facility

High heat flux testing is a vital part of engineering component validation for fusion technology. The Heat by Induction to Verify Extremes (HIVE) facility is designed to provide a practical avenue for this aspect of component testing. It provides fast turnaround for smaller concepts and a cost-effective approach by utilising induction heating within a small vacuum vessel. Due to the potential complexity of induced current paths in an induction heating system, the extent and homogeneity of power coupled to a component is difficult to model. This uncertainty increases where components have geometrical features such as the grid pattern castellations that are often present on plasma facing components in fusion reactors. This project investigates the influence of various castellation patterns on the coupling characteristics of HIVE. It shows for example that as the grid density of castellations is increased, the applied heat flux increases from 4.5 MW/m2 to 6.93 MW/m2 for an input power of 30kW over 2 s, due to an improvement in the efficiency of the inductive coupling from 13.4% to 20.8%. Additional experimental factors affecting efficiency and homogeneity of heating are also discussed.

New WC-Cu composites for the divertor in fusion reactors

The requirements for the divertor components of future fusion reactors are challenging and therefore a stimulus for the development of new materials. In this paper, WC-Cu composites are studied for use as thermal barrier between the plasma facing tungsten tiles and the copper-based heat sink of the divertor. Composite materials with 50% vol. WC were prepared by hot pressing and characterized in terms of microstructure, density, expansion coefficient, elastic modulus, Young's modulus and thermal diffusivity. The produced materials consisted of WC particles homogeneously dispersed in a Cu matrix with densifications between 88% and 98%. The sample with WC particles coated with Cu evidenced the highest densification. The thermal diffusivity was significantly lower than that of pure copper or tungsten. The sample with higher densification exhibits a low value of Young's modulus (however, it is higher compared to pure copper), and an average linear thermal expansion coefficient of 13.6 × 10−6 °C-1 in a temperature range between 100 °C and 550 °C. To estimate the behaviour of this composite in actual conditions, a monoblock of the divertor in extreme conditions was modelled. The results predict that while the use of WC-Cu interlayer leads to an increase of 190 °C on the temperature of the upper part of the monoblock when compared to a pure Cu interlayer, the composite will improve and reduce significantly the cold-state stress between this interlayer and the tungsten.

Production of tungsten-fibre reinforced tungsten composites by a novel continuous chemical vapour deposition process

The brittleness below the ductile-to-brittle transition temperature and the embrittlement during operation are the main drawbacks for the use of pure tungsten in plasma facing components of fusion reactors. Tungsten fibre-reinforced tungsten composites overcome this problem by utilizing extrinsic mechanisms to improve the toughness. Dense samples (>99%) have been successfully produced by a step-wise chemical deposition process of single layers of equally spaced fibres. The sequential deposition process with intermediate vents for the fibre placement leads, however, to inhomogeneities and internal interfaces in the deposited tungsten matrix. A new set-up utilizing a continuous deposition process was developed and used for the fabrication of first samples. These samples were examined by microstructural analysis and compared to material produced by the standard technique. It is shown that this advanced set-up in combination with tungsten fabrics allows the productions of tungsten-fibre reinforced tungsten composite in a more controlled and considerably faster way.

Fabrication of thin-walled fusion blanket components like flow channel inserts by selective laser melting

New manufacturing methods for the production of key components for nuclear fusion reactors by selective laser melting (SLM) are currently under investigation at Karlsruhe Institute of Technology. SLM offers great potential compared to conventional manufacturing methods. Within the framework of feasibility studies, complex 3D structures, such as a thin- and double-walled flow channel inserts (FCIs) for dual-coolant lead lithium (DCLL) blankets, were successfully manufactured and subjected to preliminary tests. The paper shows that fabrication of thin-walled FCIs by the SLM technique is feasible in principle. The complexity of the fabrication process is outlined and application of the SLM technique for production of other fusion blanket sub-components, such as thin-walled cooling plates with internal cooling channels, is addressed.

Conceptual design of cryogenic system for comprehensive research facility for key fusion reactor core systems

To break the bottlenecks of key components of fusion reactors, the Chinese government have included the construction of a Comprehensive Research Facility for Key Fusion Core System in its recent National Grand Project plans. It encompasses two synergistic research platforms, one for superconductor and magnets capable of a maximum field of 15T, and the other for divertor physics, material and component capable of a maximum plasma particle flux of 1024/m2s. The cryogenic system provides the 4.5K, 3K and 1.8K cooling power to all the magnets, superconductor components and cryopumps. The total average cooling capacity is 7.2kW/4.5K (1kW/3K, 250W/1.8K). This paper describes the conceptual design of the system including the helium refrigerators and cryogenic research platform. The features of the cryogenic system and the project schedule are also described.

Important Thermodynamic Parameters of Lithium–Tin Alloys from the Point of View of Their Use in Tokamaks

Along with other liquid metals, lithium–tin alloys are considered an alternative to the use of solid materials in the development of plasma-facing intrachamber components for future fusion reactors. The possibility of the practical use of these alloys in tokamaks is determined by their thermodynamic parameters, their ability to retain hydrogen isotopes in the melt, and the properties that determine the principles by which hydrogen isotopes can be extracted from a liquid Li–Sn alloy. The Sieverts’ constants for dilute hydrogen solutions in the melts of this system are calculated with the use of equations of the coordination-cluster model. The results of theoretical calculations are compared with previously published experimental data for two alloys of the Li–Sn system. A calculation method based on the use of the Butler equation, together with the experimentally obtained concentration dependences of the thermodynamic potentials of the components of this binary system, is applied to find the composition of the liquid-alloy surface.

Validation of Multi-Physics Integrated Procedure for the HCPB Breeding Blanket

The wide range of requirements and constraints involved in the design of nuclear components for fusion reactors makes the development of multi-physics analysis procedures of utmost importance. In the framework of the European DEMO project, the Karlsruhe Institute of Technology (KIT) is dedicating several efforts to the development of a multi-physics analysis tool allowing the characterization of breeding blanket design points which are consistent from the neutronic, thermal-hydraulic and thermal-mechanical points of view. In particular, a procedure developed at KIT is characterized by the implementation of analysis software only. A preliminary step for the validation of such a procedure has been accomplished using a dedicated model of the DEMO Helium Cooled Pebble Bed Blanket 4th outboard module. A global model representative of nuclear irradiation in DEMO and two local models have been set up. Nuclear power deposition and the spatial distribution of its volumetric density have been calculated using Monte Carlo N-Particle transport code for the aforementioned models and compared in order to validate the procedure set up. The outcomes of this comparative study are herein presented and critically discussed.

Molecular dynamics study of trapping and detrapping process of hydrogen in tungsten vacancy

Tungsten (W) alloys and W-based alloys are the primary candidate materials for plasma-facing components in future fusion reactors (e.g. ITER and CFETR). One of the critical issues still to be clarified in the design of the fusion reactor materials is the retention of hydrogen (H) isotopes in W, when the plasma-facing materials are supposed to sustain high-flux plasma and high-energy neutron. The dynamical behaviours of H in W with radiation defects (e.g. vacancy) are of serious concerns for understanding the mechanism of H capture, retention and permeation in W. In this work, a new model to extract the effective capture radius (ECR) and dissociation coefficient simultaneously is presented through coupling the trapping process and detrapping process of H in W vacancy. In the new model, the quantity ratio of vacancy to H atom in vacancy-H complex (VH<sub><i>x</i>+1</sub>) in the molecular dynamics (MD) simulations is described as a function of time, while the exact occurrence time of corresponding event is not required. This new model, combined with extensive MD calculations, enables the simultaneous determining of the ECR and dissociation coefficient of H in W vacancy. It is found that the parameters are dependent not only on the event type but also on temperature. The dissociation energy of H from vacancy-H complex decreases gradually with the increase of the trapped number of H atoms in the vacancy-H complex. It is also found that the common assumption (i.e. the ECR is equal to one lattice constant and the pre-exponential factor is equal to 10<sup>13</sup> s<sup>–1</sup>) in the long-term simulation methods (e.g. kinetic Monte Carlo and rate theory) is not always valid, since these calculated dynamical parameters are dispersive. The new model to obtain more reliable results with lower cost of computing resources can be easily extended into the other similar kinetic processes (e.g. H/He trapping and detrapping processes in other materials systems). These calculated dynamical parameters should be potentially helpful in supplying the initial input parameters for the long-term simulation methods.

Hardening and Strain Localisation in Helium-Ion-Implanted Tungsten

Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. In service, bombardment with energetic neutrons will create collision cascades that leave behind lattice defects. Helium, injected from the plasma and produced by transmutation, strongly interacts with these defects, modifying their behaviour and retention. Helium-ion-implantation provides an effective tool for examining helium-defect interactions and their effect on the properties of tungsten armour. We use nano-indentation to probe the mechanical properties of the shallow helium-ion-implanted layer. Comparison of spherical indents in unimplanted and helium-implanted regions of the same single crystal shows a large increase in hardness and substantial pile-up in the implanted material. The complex lattice distortions beneath indents are probed non-destructively using 3D-resolved synchrotron X-ray micro-diffraction. Reduced lattice rotations and indentation-induced residual strains in the ion-implanted material indicate a more confined plastic zone. This is confirmed by HR-EBSD and TEM observations. Together our results suggest that dislocation motion is initially obstructed by helium-induced defects. The obstacle strength of these defects is reduced by the passage of dislocations, leading to a strain-softening and slip channel formation. A constitutive law for 3D crystal plasticity finite element simulations is developed based on this hypothesis. Simulations of the indentation process successfully capture the helium-implantation induced changes in deformation behaviour observed in experiments. Importantly, the effects we observe are markedly different from previous observations for self-ion-implanted tungsten, highlighting that the exact nature of the implantation damage plays a critical role in determining mechanical property change.

A tritium permeation ‘short cut’ for plasma-facing components of fusion reactors

In ITER and other fusion reactors, plasma-facing components (PFCs) will be subjected to intense plasma/neutral fluxes and tritium (T) permeation issue will be raised. In the present work, first-of-a-kind experiments on hydrogen isotope (D) transport through ITER-like PFC mock-ups have been demonstrated in a linear plasma facility. 8 mm thick W tiles with 0.5 mm gaps are installed on structural materials and exposed to D plasmas. The angle between the gaps and the magnetic field line is set to be 5° to ensure the structural materials behind W are not exposed to charged D ions from plasma. When plasma is started up, the sample mock-ups are found to be penetrated by D within tens of seconds, suggesting that D can penetrate through heat sink of PFCs without passing through the thick W armor, i.e. the gaps act as a hydrogen isotope leakage ‘short cut’. To make a straightforward comparison of permeation flux for these materials, direct D plasma-driven permeation (PDP) experiments through W and structural materials are performed as well. The measured permeated D fluxes through structural materials are found to be orders of magnitude higher than that of W, indicating T transport from the gaps of PFCs may contribute a remarkable fraction of total leakage in reactors. In the latter part of this paper, a solution to address the permeation issue is proposed.

Helium-implantation-induced lattice strains and defects in tungsten probed by X-ray micro-diffraction

Tungsten is the main candidate material for plasma-facing armour components in future fusion reactors. Bombardment with energetic fusion neutrons causes collision cascade damage and defect formation. Interaction of defects with helium, produced by transmutation and injected from the plasma, modifies defect retention and behaviour. Here we investigate the residual lattice strains caused by different doses of helium-ion-implantation into tungsten and tungsten‑rhenium alloys. Energy and depth-resolved synchrotron X-ray micro-diffraction uniquely permits the measurement of lattice strain with sub-micron 3D spatial resolution and ~10−4 strain sensitivity. Increase of helium dose from 300 appm to 3000 appm increases volumetric strain by only ~2.4 times, indicating that defect retention per injected helium is ~3 times higher at low helium doses. This suggests defect retention is not a simple function of implanted helium dose, but strongly depends on material composition and presence of impurities. Conversely, analysis of W-1 wt% Re alloy samples and of different crystal orientations shows that both the presence of rhenium, and crystal orientation, have a comparatively small effect on defect retention. These insights are key for the design of armour components in future reactors where it will be essential to account for irradiation-induced dimensional change when predicting component lifetime and performance.

Refractory metals as structural materials for fusion high heat flux components

Tungsten is the favoured armour material for plasma facing components for future fusion reactors, but studies examining the use of tungsten or other refractory metals in the underlying cooled structures have historically excluded them, leaving current concepts heavily dependent on copper alloys such as copper chrome zirconium. This paper first outlines the challenge of selecting an appropriate alternative material for this application, with reference to historical selection methodology and design solutions, and then re-examines the use of refractory metals in the light of current design priorities and manufacturing techniques.The rationale for considering refractory alloys as structural materials is discussed, showing how this is the result of relatively small changes to the logic previously applied, with a greater emphasis on high temperature operation, a re-evaluation of current costs, a relaxation of absolute activation limits, and the availability of advanced manufacturing techniques such as additive manufacturing. A set of qualitative and quantitative assessment criteria are proposed, drawing on the requirements detailed in the first section; including thermal and mechanical performance, radiation damage tolerance, manufacturability, and cost and availability. Considering these criteria in parallel rather than sequence gives a less binary approach to material selection and instead provides a strengths and weaknesses based summary from which more nuanced conclusions can be drawn.Data on relevant material properties for a range of candidate materials, including elemental refractory metals and a selection of related alloys are gathered from a range of sources and collated using a newly developed set of tools written in the python language. These tools are then used to apply the aforementioned assessment criteria and display the results. The lack of relevant data for a number of promising materials is highlighted, and although a conclusive best material cannot be identified, refractory alloys in general are proposed as worthy of further investigation.

Effect of the magnetic field and current orientation on the splashing of liquid metal free surface of fusion reactor PFCs

In future tokamak nuclear fusion devices, the development of suitable plasma facing materials is considered to be one of the great challenges. Liquid metal, instead of conventional solid materials, has been proposed as a potential solution to plasma facing components (PFCs) for future nuclear fusion reactors, and it can solve the problems that most fusion reactors are facing. The splashing of liquid metal into the fusion plasma is a big problem because of the implementation of liquid metal PFCs in fusion reactors. An experiment involving the splashing of liquid metal has been developed under a strong magnetic field and electric field to imitate the interaction between plasma electric current and PFCs of liquid metal in fusion reactors. The current direction is perpendicular or parallel to the magnetic field direction. The influences of Kelvin–Helmholtz (KH) instability are experimentally observed and quantitatively analyzed on the magneto-electricity-driven liquid metal free surface. Processes of liquid metal splashing in the rectangular chamber are recorded by a high-speed camera. When the Lorentz force is upward and the external magnetic field and liquid thickness are constant, the liquid metal splashing from the liquid metal free surface when the external current exceeds the critical value. Critical current presents as almost inverse proportional functions with the magnetic field variation for different liquid thicknesses. In the experiment, the instability is characterized by the dimensionless numbers Bond number, Bd, determined from the importance of gravitational forces compared to surface tension forces, and the Hartmann number, Ha, determined from the intensities of the imposed magnetic field. Multiple linear regression models of the critical current density are summarized which indicates that the magnetic field and current’s influence determine the splashing critical point of liquid metal under the condition of multi-field coupling.

Thermal evolution of irradiation defects in ferritic/martensitic steel during isochronal annealing

The aim of investigating the annealing effect induced by helium/hydrogen ion irradiation on the micro-defects of ferritic/martensitic (FM) steel is to gain a basic understanding on the development of fusion reactor components. In this study, 250 keV He2+ and 130 keV H+ were used for irradiation. Then, micro-defects were annealed isochronally between 423 K and 673 K, and were monitored by the slow positron beam Doppler broadening technique. The results revealed that the He ions implanted into the steels combined with vacancies to form HenVm clusters, and were likely to grow and form overpressured HenVm clusters by thermal activation. The S-W line of the irradiated specimen deviated the line segment, which indicates the formation of overpressured HenVm clusters in the He-irradiated steel. Moreover, the ΔS-E curves indicate residual vacancies that were induced by He-irradiation at 723 K and later migrated toward the bulk of the sample at 423 K. Subsequently, the vacancies tended to migrate to the sample surface as the annealing temperature increased. The S parameter of the He+H-irradiated sample was much larger than that of all other samples, owing to the synergy effect of H and He. After annealing at 423 K, the H-V clusters decomposed and the hydrogen atoms in the He+H-irradiated sample escaped from the surface leaving a large number of vacancies.

Review of the Development of the CFETR Divertor with Additional Electromagnetic Studies

As one of the key components in tokamak fusion reactor, the divertor is often exposed to tritium environment, high heat flux, and neutron radiation which are harmful for human beings and the divertor itself, which is damaged easily. So maintenance for the divertor is necessary. However, due to neutron damage and activation, maintenance for the divertor should be done in a remote way rather than by personnel directly, especially the process for installing the divertor into the inner and outer rails in the vacuum vessel from the outside. As an important compatible structure for the divertor remote handling (RH) process, inner and outer supports should be considered for ensuring the RH process goes well. Thus this paper introduces the China Fusion Engineering Test Reactor divertor support design mechanism, including inner support and outer support, which are based on the requirements of RH. Then the electromagnetic forces generated by Halo current and Eddy current due to plasma disruption are analyzed in order to validate the design of the newly designed support mechanism.

ODS ferritic steels obtained from gas atomized powders through the STARS processing route: Reactive synthesis as an alternative to mechanical alloying

Oxide Dispersion Strengthened Ferritic Stainless Steels (ODS FS) are candidate materials for structural components in fusion reactors. Their ultrafine microstructure and the presence of a very stable dispersion of Y-Ti-O nanoclusters provide reasonable fracture toughness, high mechanical and creep strength, and resistance to radiation damage at the operation temperature, up to about 750 °C.An innovative route to produce ODS FS with composition Fe-14Cr-2W-0.3Ti-0.3Y2O3 (wt.%), named STARS (Surface Treatment of gas Atomized powder followed by Reactive Synthesis), is presented. This route avoids the mechanical alloying (MA) of the elemental or prealloyed powders with yttria to dissolve the yttrium in the ferritic matrix.In this study, starting powders containing Ti and Y are obtained by gas atomization at laboratory and industrial scale. Then, a metastable Cr- and Fe- rich oxide layer is formed on the surface of the powder particles. During consolidation by HIP the metastable oxide layer at Prior Particle Boundaries (PPBs) dissociates, the oxygen diffuses towards saturated solutions or metallic Ti- and Y-rich particles, and Y-Ti-O nano-oxides (mainly Y2TiO5) precipitate in the ferritic matrix.Detailed Microstructural characterization by X-ray Photoelectron Spectroscopy (XPS), X-ray Absorption Spectroscopy (XAS), Scanning Electron Microscopy (SEM) and Transmission Electron Microscopy (TEM) of powders and consolidated materials is presented and correlated with mechanical behaviour.

Fabrication and Characterization of Li<sub>4</sub>SiO<sub>4</sub> Ceramic Pebbles Doped with Y<sub>2</sub>O<sub>3 </sub>and Nb<sub>2</sub>O<sub>5</sub>

Tritium breeder blanket has been considered as one of the most important components of thermonuclear fusion reactor. Recently, Li4SiO4has been regarded as one of the favored ceramics breeders as it exhibits excellent tritium release properties. In order to further improve the comprehensive properties of Li4SiO4pebbles, the grain size and crush load of Li4SiO4pebbles doped with Y2O3and Nb2O5were systematically studied in this work. The results showed that the relative density of Li4SiO4pebbles was increased to 87.2% and the crush load was increased from 43N to 48N when Y2O3doping was used. In view of inhibiting grain growth and increasing the crush load and relative density of Li4SiO4pebbles, the effect of Nb2O5doping was better than that of Y2O3doping.

Formation of gradient microstructure and mechanical properties of hot-pressed W-20 wt% Cu composites after sliding friction severe deformation

W-based alloys are currently considered promising candidates for high heat flux components in future fusion reactors. In this paper, hot pressed W-20 wt%Cu composites were treated at room temperature using a sliding friction severe deformation (SFD) process, with a moving speed of 0.2 m/s and an applied load of 500 N. Microstructural evolution of composites after the SFD treatment was evaluated and compared with that of the untreated composites. Results showed that there was a gradient structure generated and an obvious refinement in tungsten particles size in the surface layer after the SFD process. The average particle size of tungsten in the SFD treated composites was 2.60 μm, whereas it was 4.5 μm for tungsten in the untreated composites. Fracture surfaces of the composites indicated that the SFD treatment destroyed the W skeleton and changed fracture mode from predominant inter-granular one to trans-granular one due to the decrease in contact area of W-W inter-particles. Yield strength and ultimate tensile strength of composites after the SFD treatment were 308 MPa and 553 MPa, respectively. The treated composites exhibited micro-hardness values with an average reading of about 308 HV. Analysis of the facture microstructures clearly suggested that the tungsten particles in the treated composites are consisted of dislocations and boundaries as well as dislocation tangles. The electrical conductivity of the composites was decreased from 33 IACS% to 28.5 IACS% after the SFD treatment, mainly due to loss or squeezing of copper into the inner surface.

Surface morphology of Tungsten-F82H after high-heat flux testing using plasma-arc lamps

F82H reduced activation steel coated with vacuum plasma sprayed (VPS) tungsten is a candidate as a plasma facing material for main chamber components in future fusion reactors. Due to different coefficients of thermal expansion (CTE), significant thermal stresses are expected in these bimetallic materials. Thus, a major uncertainty in the performance of W/F82H components during the operation under high-heat fluxes is the effect of CTE mismatch. In this study, a high intensity plasma-arc lamp was used for high-heat flux cycling tests of W/F82H specimens. While no surface damage was observed for specimens tested for 100–200 cycles at a heat flux of 1.4 MW/m2 pulse when the backside surface temperature was maintained below 550 °C, significant cracking occurred at higher temperatures. A simple analytical model for bimetallic materials indicated that the stress in the VPS-W layer is likely to exceed its failure stress solely due to the bilayer thermal stress. A finite element analysis of the state of stress and deformation confirmed that a significant stress also would occur at the W surface due to the rigid-body like constraint imposed by the clamp, which can be the main cause of the cracking.