Tungsten Armor Research Articles

Designing tungsten armoured plasma facing components to pulsed heat loads in magnetic fusion machines

A possible design rule for preventing surface damage from thermal transients to solid tungsten armour is proposed and formulated for the plasma facing components (divertor, first wall) of magnetic fusion machines. The rule is based on combined results from laboratory experiments and operating fusion machines, and fundamental engineering principles such as the heat flux factor (FHF) and fatigue usage fraction (FUF). As an example, the rule would allow 2.104 transient heat loads cycles at a FHF of 10MJm-2s-½ before the lifetime is considered exhausted. The formulation of the rule using engineering principles allows combining loads of different magnitudes and various number of cycles. A practical example of the rule usage is provided, illustrating loads combination and how the rule may contribute to the component geometrical design. The proposed rule is only valid for surface loading conditions, hence is not usable for volumetric loading conditions such as runaway electrons. Setting a budget lifetime and a design rule does not preclude actual plasma operation beyond the design lifetime. It is actually normal that experimental devices explore a larger domain than the one defined at the time of the design.

High heat flux components with beryllium armour: From the small-scale mock-ups to the full-scale prototype of the ITER first wall panel

The Russian Federation shall supply 179 enhanced heat flux First Wall Panels (FWP) to the ITER construction site. These panels have about 30 types and consist of a steel supporting structure (FW Beam) with multilayer plasma-facing units (PFU) attached onto it. To start the production of the panels, the manufacturing readiness must be confirmed by the ITER Organization (IO). The JSC “NIIEFA” responsible for manufacturing of a part of the components, final assembly and testing of the panels has come a long way from the experimental work on small-scale mock-ups, through the development of the design and key manufacturing technologies to the production and testing of the full-scale prototype (FSP) which is the final step before the confirmation for the manufacture. The manufactured FSP has successfully passed most of the acceptance tests including hydraulic pressure and leak tightness tests. The high heat flux tests have been successfully carried out on the PFUs with beryllium armour. In parallel to the tests of the beryllium prototype, the tungsten armour has been analyzed and tested which is due to the IO's intention to change the armour material.

Deuterium retention in cyclic transient heat loaded tungsten with increasing cycle numbers

Surface damage and microscopic defect evolution of tungsten (W) armor under transient heat loads are key factors for fuel retention in fusion reactors. In this work, experiments were conducted to investigate the effects of cyclic thermal shocks on deuterium (D) retention and surface blistering in W. Thermal shock experiments were conducted on recrystallized W using an electron beam with a power density of 0.15 GW m−2 across 100–1500 cycles, followed by D plasma exposure with high-fluence (∼1 × 1026 D m−2). The results demonstrate that samples subjected to 500 and 1500 cycles exhibit a significant presence of sub-grains within 90 μm. Notably, the inhibition of blistering induced by thermal shock leads to a substantial reduction in D retention (5.45 × 1019 D m−2) at lower cycle numbers (100 cycles) compared to the reference sample (2.35 × 1020 D m−2) which was only exposed to D plasma. When cycle numbers increase to 500 and 1500, D retention reaches 1.98 × 1020 D m−2 and 4.56 × 1020 D m−2, respectively. Based on the tritium migration analysis program, we propose that total D retention is a consequence of the competition between defects reduced by thermal shock-induced suppression of blistering and defects generated by plastic deformation induced by thermal stress. D retention initially decreases with the increase in cycle numbers, followed by a subsequent rise, with the inflection point slightly higher than 500 cycles. Additionally, due to the extensive scope of thermal stress, an escalated exposure period will result in substantial D captured by heat-induced defects, consequently intensifying the D retention. Whether there exists an upper limit to D retention induced by the increasing thermal shock cycles necessitates further experimental analysis. Nonetheless, it is evident that thermal shock significantly contributes to D retention within a profoundly deep bulk region under high cycles.

Thermal and structural analysis of JT-60SA actively cooled divertor target submitted to high heat flux

JT-60SA is a joint international fusion experiment being built and operated in Japan under the framework of the Broader Approach Agreement and Japanese National Fusion Programme, aiming at an early realization of fusion energy by conducting supportive and complementary work for the ITER project, all towards supporting the basis for DEMO. In this context, a collaboration has been established via EUROfusion between Fusion for Energy and CEA to develop actively cooled targets of the JT-60SA divertor.The first Actively Cooled Divertor (ACD) target, planned to be operated in 2029, has graphite as armour material and TZM as heat sink material. To be DEMO relevant, ACD target with tungsten armour is planned to be operated in 2033, with either CuCrZr or TZM as heat sink material.This paper focusses on the choice of the design (material, dimensions) of these targets. Results of thermal and structural analysis of these targets are presented. The calculations, carried out in the environment of Ansys 2021R2 code, have taken into account the heat loading under plasma operation (with a peak heat flux of 10 MW/m2) and the electromagnetic loads from a representative plasma disruption (VDE 30 ms). In order to check the compliance with nuclear requirements, RCC-MRx 2015 has been used for the mechanical assessment of metallic components, while for the graphite the comparison between the maximum stress and the yield stress has been considered. The performed analyses show that the thermal load on target during the plasma heat loading drives target designs due to differential expansion at the interface between the heat sink and the flat tiles.Additionally, on-going assessments on these divertor targets are presented, focussing on experimental High Heat Flux testing, as well as computational thermal-hydraulic studies on enhanced target configurations.

Analysis of the nuclear loads on Chromium monoblock divertor target for DEMO

The Plasma Facing Components (PFCs) of DEMO divertor play a fundamental role for the heat removal and particle exhaust functions, during fusion plasma. This implies a very harsh operation environment characterized by the combination of complex loading and stress conditions, intense particle bombardment, high heat flux deposition and a significant neutron irradiation. In the framework of the European DEMO divertor project (WPDIV), several target design concepts alternative to the baseline ITER-like technology are under evaluation, such as Chromium block design. It consists of chromium block, tungsten armor tile and copper alloy cooling pipe, and it is very attractive from an activation reduction, waste management and safety point of view. Therefore, the assessment of the most impacting nuclear loads for the PFC design is a pivotal aspect for DEMO divertor R&D program and this study is fully devoted to the comprehension of the spatial distributions of the main nuclear quantities (radiation damage, He-production and nuclear heating density) on the overall divertor targets. Neutronics analyses have been performed with MCNP5 Monte Carlo code using both DEMO MCNP Water Cooled Lithium Lead (WCLL) and Helium Cooled Pebble Bed (HCPB) blanket models, with a fully heterogeneous representation of the divertor cassette and Chromium PFCs geometry. The results of this analysis show that nuclear heating density and radiation damage distributions are higher with WCLL blanket than HCPB, except for He-production which seems not significantly affected by the blanket configuration. For each nuclear quantity, the peak values are achieved in correspondence of the baffle region and especially on the external units of the targets, as expected, being this region the most exposed to irradiation.

Temperature gradient based annealing methodology for tungsten recrystallization kinetics assessment

In future thermonuclear fusion reactors, like ITER and DEMO, plasma-facing components with tungsten armor material will have to sustain high thermal fluxes (10 MW/m² in steady state and 20 MW/m² in quasi-steady-state). For such extreme conditions, tungsten may reach temperatures higher than 1000 °C and, consequently can be prone to recrystallize, which can limit the lifetime of the divertor targets under cycling thermal loadings. Characterization of recrystallization kinetics involves the use of different methods and devices (furnace, laser heating) depending on the annealing temperature regime to be analyzed. For these methods, the heating and the subsequent characterization of around 10 samples (5 × 4 × 4 mm) per annealing temperature is needed. In this article, a method is proposed to quantify the recrystallization kinetics between 1300 °C and 1600 °C using a limited number of samples and heating conditions. The basic idea is to induce a steady-state temperature gradient in tungsten rods by heating one side of the rod using a laser heating system. Tungsten rods (50 × 4 × 5 mm) were annealed with this method. The resulting microstructures are characterized using optical microscopy and hardness measurements. The different stages of tungsten softening are clearly evidenced. These first results can be considered as a proof of concept for the use of such a methodology to assess tungsten recrystallization at high temperature.

Analysis of high heat flux tested tungsten mono-blocks by hardness and microstructural profiling

The integrity of tungsten plasma facing component mockups was investigated by microstructure and hardness measurement before and after high heat flux (HHF) test. The mockups consisting of five ITER-like tungsten mono-blocks were fabricated employing gas pressure casting and hot radial pressing (HRP) method. The tungsten blocks were machined from rolled W plates of two commercial suppliers. The tested mockups showed discernible results in response to different heat flux conditions of which increased heat flux was of importance. Post-mortem analysis of the tungsten armor of the HHF tested mockups were conducted through microstructure and hardness profiling from surface along the depth. The Vickers hardness profile measured from the surface into the depth showed notable decrease around 300 μm ∼ 1600 μm depth after HHF test compared to that before the test. The depth and magnitude of hardness change was dependent on the heat loading condition and the position in the tungsten block. The behavior of hardness indicating the materials property of tungsten showed correlation with the microstructural change measured by electron back scatter diffraction (EBSD).

High Heat Flux Testing of Graded W-Steel Joining Concepts for the First Wall

The realization of the first wall (FW), which is composed of a protective tungsten (W) armor covering the structural steel material, is a critical challenge in the development of future fusion reactors. Due to the different coefficients of thermal expansion (CTE) of W and steel, the direct joining of them results in cyclic thermal stress at their bonding seam during the operation of the fusion reactor. To address this issue, this study benchmarks two joining concepts. The first concept uses an atmospheric plasma sprayed graded interlayer composed of W/steel composites with a varying content of W and steel to gradually change the CTE. The second concept uses a spark plasma sintered graded interlayer. Furthermore, in order to benchmark these concepts, a directly bonded W-steel reference joint as well as a W-steel joint featuring a vanadium interlayer were also tested. These joints were tested under steady-state high heat flux cyclic loading, starting from a heat flux of 1 MW/m2 up to 4.5 MW/m2, with stepwise increments of 0.5 MW/m2. At each heat flux level, 200 thermal cycles were performed. The joints featuring a sintered graded interlayer survived only until 1.5 MW/m2 of loading, while the joint featuring plasma sprayed graded interlayer and V interlayer survived until 3 MW/m2.

Processing and Properties of Tungsten-Steel Composites and FGMs Prepared by Spark Plasma Sintering

Tungsten is the prime candidate material for the plasma-facing components of fusion reactors. For the joining of tungsten armor to the cooling system or support structure, composites or graded interlayers can be used to reduce the stress concentration at the interface. These interlayers can be produced by several technologies. Among these, spark plasma sintering appears advantageous because of its ability to fabricate fully dense parts at lower temperatures and in a shorter time than traditional powder metallurgy techniques, thanks to the concurrent application of temperature, pressure, and electrical current. In this work, spark plasma sintering of tungsten-steel composites and functionally graded layers (FGMs) was investigated. As a first step, pure tungsten and steel powders of different sizes were sintered at a range of temperatures to find a suitable temperature window for fully dense compacts. Characterization of the sintered compacts included structure (by SEM); porosity (by the Archimedean method and image analysis); thermal diffusivity (by the flash method) and mechanical properties (microhardness and flexural strength). Compacts with practically full density and fine grains were obtained; while the temperature needed to achieve full sintering decreased with decreasing powder size (down to about 1500 °C for the 0.4 μm powder). For fully sintered compacts, the hardness and thermal diffusivity increased with decreasing powder size. Composites with selected tungsten/steel ratios were produced at several conditions and characterized. At temperatures of 1100 °C or above, intermetallic formation was observed in the composites; nevertheless, without a detrimental effect on the mechanical strength. Finally, the formation of graded layers and tungsten-steel joints in various configurations was demonstrated.

Rationale Behind EU-DEMO Limiter’s Plasma-Facing Component Design Under Material Phase Change

The protection strategy adopted for the European DEMOnstration (EU-DEMO) fusion power reactor foresees the use of sacrificial components–referred to as limiters–dealing with plasma-wall contacts. Their aim is to protect the first wall (FW) against the huge amount of plasma energy (up to GigaJoules) released in a few milliseconds during disruptive events, which might lead to melting and/or vaporization of the foreseen plasma-facing tungsten armor. The current limiter design concepts rely on actively water-cooled plasma-facing components (PFCs) made of tungsten. As water is not allowed inside the main chamber, limiter’s PFCs must be designed to preserve the cooling system integrity under any scenario, therefore the estimate of the thickness of material undergoing any phase change is crucial. Given the initial assessment of plasma magnetic configurations during off-normal events, this article describes the procedure followed for designing the limiter’s PFC, which includes a novel approach for estimating the molten thickness of material under high heat flux. A simplified 1-D model has been implemented in MATLAB, which deals with multiphase moving boundary problems, hence its name Thermal Analysis foR Tracking InterFaces under meLting&vaporizaTion-induced plasma Transient Events (TARTIFL&TTE). This model takes its inspiration from the way phase change interface tracking problems are tackled in food industry (i.e., freeze-drying processes) and solute concentration diffusion-controlled problems. It overcomes the complexity of solving a strongly coupled non-linear system of partial and ordinary differential equations in moving spatial domains by adopting a change of coordinate system based on the Landau transformation. As a result, an equivalent and fixed spatial coordinate system is defined where the spatial domain boundaries of the different phases are constrained, and the systems of governing equations are easy to solve. Both the solid and liquid phases are modeled, while the vapor is assumed to be removed once formed. Its benchmark against computational results found in the literature has shown a very good agreement, which paves the way to further development of it. For phase change interface tracking problems in 2-D/3-D and more complex geometries, a commercial Multiphysics software adopting the Lagrangian–Eulerian approach in moving mesh frames will be used for tackling problems where material is removed following a phase change. Although the vapor domain is not simulated, a set of gas kinetics boundary conditions couples the interface between vapor and liquid phases, driving its position over time. This will be detailed described in a future companion article.

Electrodeposition of Functionally Graded Brazing Interlayers for Enhanced Joint Strength between Fusion Plasma-Facing Materials and Heat Sinks

In response to the significant plasma science and engineering breakthroughs witnessed over the past decade, the fusion research community has developed an ambitious roadmap to achieve large-scale, energy-positive fusion reactors that will enable economically competitive fusion electricity. Critical to the success of commercial fusion energy is the development of materials able to serve as plasma facing components (PFCs) that can withstand the unprecedented high heat, particle, and neutron fluxes within the fusion reactor. Tungsten is a leading candidate for the plasma-facing portion of these PFCs because of its high melting point, high sputtering resistance, and low tritium retention. Copper or copper-based alloys (e.g., copper-chromium-zirconium alloy C18150) and reduced activation ferritic/martensitic (RAFM) steels have been proposed for the heat sink materials behind the tungsten armor due to their high thermal conductivity. Brazing is considered a promising method for joining the tungsten armor and heat sink layer. However, the extreme mismatch in the coefficient of thermal expansion (CTE) between tungsten and these potential heat sink materials makes directly joining these two layers challenging. This thermal stress can be reduced by incorporating a joining layer of functionally graded materials between the tungsten and heat sink that imparts gradual changes in CTE. Copper/tungsten or copper/tungsten carbide composites are leading candidates for these functionally graded materials. For these composites, gradual changes in the CTE are achieved by varying the volume fraction of tungsten/tungsten carbide in the composite, from tungsten-rich at the tungsten armor to tungsten-poor at the heat sink.This talk will discuss the development of a scalable, electrochemical approach for fabricating interlayers with functionally graded CTE that can decrease the thermal mismatch between tungsten PFCs and copper or RAFM-based heat sinks. In this approach, the copper/tungsten or copper/tungsten carbide composite are electrodeposited onto the tungsten PFC. This component is then bonded to the heat sink via brazing. A key objective of the work is to achieve compositional control during electrodeposition through the use of highly tunable pulse and pulse-reverse electric fields. Engineering of the electric field enables control of the mass transport and crystallization phenomena in the deposition process and thus enables control of composite properties such as composition, porosity, and surface roughness. Results from electrodeposition trials and mechanical testing of brazed joints under ambient conditions as an initial screen for interlayer performance will be discussed, in anticipation of high-heat flux testing of joint performance under fusion-relevant conditions.

Crack formation in the tungsten armour of divertor targets under high heat flux loads: A computational fracture mechanics study

In the framework of EUROfusion, the thermally induced stress and the consequential fracture of the design of DEMO-divertor with copper alloy heat sink have been analyzed, to assess the structural integrity of tungsten armor of the plasma facing component (PFC). With finite element method (FEM), the influence of mesh and plasticity of tungsten has been evaluated. Due to the lack of material data, especially those of irradiated tungsten under operation conditions, conservative assumptions have been made in terms of plasticity, fracture strength and fracture toughness. The crack initiation and propagation with assumed plasticity and fracture toughness for heat fluxes ranging from 10MW/m2 up to 30 MW/m2 have been analyzed. Thermal stress analysis has been performed with various plasticity of tungsten, to locate the peak stress, which leads to the highest possibility of crack initiation. Extended finite element method (XFEM) analysis has been performed afterwards with various fracture strength toughness. The influence of plasticity and fracture toughness from almost zero up to Kic = 20MPa∙m1/2 has been evaluated. The predicted crack propagation has been verified by results of J-integral calculation.

Preliminary exploration of a WTaVTiCr high-entropy alloy as a plasma-facing material

With great power comes great challenges. For nuclear fusion, the holy grail of energy, taming the flame of a miniature star in a solid container remains one of the most fundamental challenges. A tungsten armour for the solid container marks a temporary triumph—a solution adopted by the world’s largest fusion experiment, ITER—but may be insufficient for future challenges. High-entropy alloys (HEAs), which are characteristic of a massive compositional space, may bring new solutions. Here, we explore their potential as plasma-facing materials (PFMs) with a prototype W57Ta21V11Ti8Cr3 HEA that was designed by exploiting the natural-mixing tendency among low-activation refractory elements. Revealed by x-ray diffraction analysis and energy-dispersive x-ray spectroscopy, it predominantly consists of a single bcc-phase but with V, Ti, and Cr segregation to grain boundaries and at precipitates. Its yield strength improves ∼60% at room temperature and oxidation rate reduces ∼6 times at 1273 K, compared with conventionally used W. The Ti–V–Cr rich segregations and the formed CrTaO4 compound contribute to the improved oxidation resistance. However, the Ti–V–Cr rich segregations, along with the decreasing valence-electron concentration of the matrix by the addition of Ta, V and Ti elements, considerably increase the deuterium retention of the W57Ta21V11Ti8Cr3 HEA to ∼675 multiples of recrystallized W. Moreover, its thermal conductivity decreases, being ∼40% of W at 973 K. However, the maximum tolerable steady-state heat load is still ∼84% of W because of its exceedingly high yield strength at elevated temperatures. Overall, despite being preliminary, we expect HEAs to play an important role in the development of advanced PFMs, for their disadvantages are likely to be compensated by their advantages or be overcome by composition optimization.

Optimization of a WCCB blanket module with detached first wall for CFETR

State-of-the-art blanket designs in fusion reactor engineering tend to use a first wall to protect the plasma facing components from high surface heat flux. This first wall is prone to get damaged, and thus should better be easily replaceable for convenient maintenance. The wall should also have a tungsten cladding of reasonable thickness to balance the blanket's tritium breeding and plasma shielding capabilities. In this regard, optimization of a CFETR water-cooled ceramic breeder (WCCB) blanket module is proposed. Compared to the present design, the optimized module features a detached first wall bolted to the back part of the blanket, with thickened tungsten armor covered on the surface. The new design not only allows convenient replacement and maintenance by remote handling, but is expected to provide improved protection against plasma surface loads. This paper presents the detailed design of the proposed blanket module, as well as feasibility analysis results from thermo-mechanical and thermal-hydraulics aspects.

Developing solid-surface plasma facing components for pilot plants and reactors with replenishable wall claddings and continuous surface conditioning. Part A: concepts and questions

It is estimated that pilot plants and reactors may experience rates of net erosion and deposition of solid plasma facing component (PFC) material of 103–105 kg yr−1. Even if the net erosion (wear) problem can be solved, the redeposition of so much material has the potential for major interference with operation, including disruptions due to so-called ‘unidentified flying objects (UFOs)’ and unsafe dust levels. The potential implications appear to be no less serious than for plasma contact with the divertor target: a dust explosion or a major UFO-disruption could be as damaging for an actively-cooled deuterium-tritium (DT) tokamak as target failure. It will therefore be necessary to manage material deposits to prevent their fouling operation. This situation appears to require a fundamental paradigm shift with regard to meeting the challenge of taming the plasma–material interface: it appears that any acceptable solid PFC material will in effect be flow-through, like liquid–metal PFCs, although at far lower mass flow rates. Solid PFC material will have to be treated as a consumable, like brake pads in cars. ITER will use high-Z (tungsten) armor on the divertor targets and low-Z (beryllium) on the main walls. The ARIES-AT reactor design calls for a similar arrangement, but with SiC cladding on the main walls. Non-metallic low-Z refractory materials such as ceramics (graphite, SiC, etc) used as in situ replenishable, relatively thin—of order mm—claddings on a substrate which is resistant to neutron damage could provide a potential solution for the main walls, while reducing the risk of degrading the confined plasma. Separately, wall conditioning has proven essential for achieving high performance. For DT devices, however, standard methods appear to be unworkable, but recently powder droppers injecting low-Z material ∼continuously into discharges have been quite effective and may be usable in DT devices as well. The resulting massive generation of low-Z debris, however, has the same potential to seriously disrupt operation as noted above. Powder droppers provide a unique opportunity to carry out controlled studies on the management of low-Z slag in all current tokamaks, independent of whether their protection tiles use low-Z or high-Z material.

Gas pressure casting technique for manufacturing of W/OFHC-Cu monoblock

The important engineering part of the divertor vertical target is the quality of joint between armor materials and heat sink. To reduce the joint interface stresses, a pure copper (Cu) interlayer was provided between the tungsten(W) armor and the CuCrZr cooling tube. Gas pressure casting method was developed to reduce the porosity and improve the mechanical properties of the Cu interlayer of W/Cu monoblock. In the vacuum casting sample without pressure, porosities were found in Cu. These porosities become defects at the interface between W and Cu and decrease the mechanical properties of the Cu interlayer. In the cooling process of casting, argon gas was pressurized into a porous tube inserted in the Cu and tungsten monoblock applying perpendicular force on the joint surface. This method was effective to remove the porosity in Cu generated during the cooling process and to increase the bonding strength between Cu and tungsten. The quality of bonding interface, depth of the Cu diffusion, and hardness of Cu were compared by SEM, EDS, and Vickers hardness measurement. In the samples made by gas pressure casting method, the porosity was decreased in Cu, no defects on the interface between tungsten and Cu, Cu was diffused into tungsten with a depth of 2∼3 μm and the hardness of Cu was increased about 15%.Small mock-ups were manufactured by hot radial pressing (HRP) with gas pressure casted W/Cu monoblock and CuCrZr cooling tube and underwent high heat flux (HHF) of 10 MW/m2 up to 5,000 cycles. The ultrasonic test (UT) results before and after the HHF test showed no defects on interfaces and was confirmed by observation of SEM on the cutting surface of the monoblock mock-up after the HHF test.

Limiters for DEMO wall protection: Initial design concepts & technology options

Off-normal macroscopic plasma instabilities which possibly occur in a fusion power plant represent the most critical operational accidents and pose a serious risk to economically viable exploitation of fusion power. During such an instability, a huge amount of plasma energy is rapidly deposited onto the plasma-facing first wall in form of extreme heat flux. Such heavy thermal shocks can cause severe damages or a fatal failure of the first wall even by a single event. This issue is a serious design concern raising the requirement of a wall protection strategy. To cope with this issue, the European DEMO project adopted the limiter concept. The limiters equipped with tungsten sacrificial armor shall be installed at those wall areas where plasma contact is expected to be probable. To implement the limiter concept, comprehensive engineering studies have been conducted including conceptual design of the plasma facing component and technology R&D for armor material and joining. In this overview paper, the background of the wall protection issue and its implication on design and materials are explained and the outcomes from the recent project period (2019-20) are reported. The focus is placed on the technology achievement related to the development of novel lattice-type tungsten armor material fabricated by an innovative additive manufacturing process. Interim results from the extreme and medium heat flux tests and the joining trials for tungsten-to-heat sink material (copper or steel) are also addressed. The design rationale and the interpretation of the test data are supported by computational simulations.

Effect of tungsten armor on electromagnetic characteristics of blanket for the fusion device under plasma disruption

Tungsten armors are arranged on the plasma side of the blankets to protect the first wall in a fusion device, such as China Fusion Engineering Test Reactor (CFETR) and European Demonstration Power Plant (EU DEMO). Tungsten armors will produce a huge eddy current and Electromagnetic (EM) load under plasma disruption due to their high conductivity and strong magnetic field environment, which will cause the blankets to be subjected to a substantial eddy current and thus heat load, thermal stress and mechanical stress, that potential could damage blankets. To evaluate the effect of the tungsten armor on the electromagnetic characteristics of blanket, electromagnetic calculation method verification is carried out. And the electromagnetic finite element model of blanket is established by using ANSYS, and the eddy current distribution in the first wall is calculated. Effects of the tungsten armor, split size and gap size of the tungsten armor on the eddy current of blanket first wall and effect of the tungsten armor on the EM force of blanket are researched. And the influence of different tungsten armor design on the thermal stress distribution of blanket first wall is also studied. The study results will provide an important reference for the blanket design of the fusion device.

Conceptual design of a liquid-metal divertor for the European DEMO

Liquid metal (LM) divertors are considered for the European DEMO reactor, because they may offer improved performance compared to the tungsten monoblock concept. The goal of this work is to provide a concept design, and explore the limitations of liquid metal divertors. To this end, a set of design requirements was formulated in close collaboration with the EUROfusion Power Plant Physics and Technology team (responsible for the design of the EU-DEMO). Tin was chosen as the preferred liquid metal, because unacceptable Tritium retention issues arise when lithium is used in DEMO. A concept design was then chosen that consists of water cooled pipes that are square on the outside and round on the inside, a corrosion barrier, and a 3D-printed porous tungsten armor layer filled with liquid tin. The porous armor layer acts as a Capillary Porous System (CPS). The design was analyzed using thermo-mechanical FEM simulations for various armor thicknesses and heat sink materials: Densimet, W/Cu composites, and CuCrZr. The highest heat loading capability achieved is 26.5 MW/m2 in steady state (18.9 MW/m2 when taking into account a safety margin of 1.4). This is achieved using a CuCrZr pipe, with a 1.9 mm thick armor. When increasing the armor layer to 3 mm thick, more than 80 MW/m2 can be withstood during slow transients thanks to vapor shielding, but at the same time the steady-state capability is reduced to 18 MW/m2. Resilience against disruptions cannot yet be proven, but is deemed within the realm of possibility based on estimates regarding the behavior of vapor shielding. This should be further investigated. Overall, the concept is considered a significant improvement compared to the original specifications (which are also the specifications to the tungsten monoblocks: 10 MW/m2 in steady state, and ∼20 MW/m2 during slow transients). Moreover, the possibility of withstanding disruptions is regarded as a potentially major improvement.

Design and Analysis of an Advanced Three-Point Bend Test Approach for Miniature Irradiated Disk Specimens

Candidate tungsten armor materials in a magnetic confinement fusion device must be able to withstand thermal variation that leads to internal stresses caused by the impinging heat load. In addition, the thermomechanical properties of these materials are degraded by irradiation-induced defect accumulation. Fission reactor–based irradiation data are used to predict the fusion neutron damage and property change. This study examines the motivation and design of a custom-designed three-point bend test for neutron-irradiated disk specimens that are 3 mm in diameter to be able to define the flexural strength of advanced tungsten materials, alloys, and composites—and to the extent that embrittlement occurs after neutron irradiation. The theory provided shows a calculation for the flexural deflection and shear deflection due to the small-geometry constraints. A finite element deformation analysis is performed to evaluate the mechanical stress field of disk bend specimens. The stress values above 80% of the maximum stress are concentrated in 2.4 mm of the 3.0-mm length of the centerline across the tungsten disk diameter. A bend test fixture has been designed and fabricated to enable testing of these specimens with precisely engineered tolerance and minimal machine compliance. This fixture will be able to be placed inside a universal testing frame at elevated temperatures for the mechanical property evaluation of future neutron-irradiated disk specimens.