Fusion Divertor Research Articles

Observation of tungsten emission spectra up to W46+ ions in the Large Helical Device and contribution to the study of high-Z impurity transport in fusion plasmas

Spectroscopic studies of emissions released from tungsten ions combined with a pellet injection technique have been conducted in Large Helical Device for contribution to the tungsten transport study in tungsten divertor fusion devices and for expansion of the experimental database of tungsten line emissions. The spectral intensities of W5+, W24+–W28+, W37+, W38+, W41+–W43+, W45+, and W46+ emission lines were measured simultaneously over a wide wavelength range from x-ray to visible. Time evolutions of the various tungsten line spectra indicate that the tungsten confinement time depends on the electron density of the plasma and is long in high density plasmas, on the order of seconds, and short in low density plasmas, on the order of sub-seconds. When the confinement time was long, the tungsten ions remained in the plasma until the end of the discharge, changing their dominant charge with the change in electron temperature. When the confinement time was short, the tungsten ions rapidly decreased in all charge states and disappeared. Space-resolved EUV and visible spectroscopy measurements have revealed that tungsten ions stayed in the core region of the plasma with changing their dominant charge state depending on the electron temperature in the discharges with the long confinement time. Detailed analysis of soft x-ray emission suggested that the confinement time increases with density and becomes saturated when the central electron density exceeds 2 × 1013 cm−3.

Experimental study of boiling heat transfer characteristics of multi-elbow cooling system under one-sided heating with long section for fusion divertor cooling

In this study, heat transfer experiments were conducted under one-sided heating conditions using water as the coolant and an electron gun as the heat source. To evaluate the boiling heat transfer characteristics of the swirling flow formed by a multi-elbow piping, heat flux by the electron gun is deposited on the region of 60 mm-section in the flow direction corresponding to 0.5 D (inner diameter of the flow channel) to 6.5 D from the elbow outlet, and then temperature distributions in the flow, circumferential and radial directions were measured. The experimental results showed that the wall temperature cooled by the swirling flow generated downstream of the multi-elbow was lower than that obtained in the case of turbulent flow in the straight pipe when irradiated with the same value of heat flux. Thereby, the heat fluxes at the boiling start and DNB points obtained in the multi-elbow case were higher than those in the straight pipe case. These results imply the superior cooling performance of multi-elbow system.

Algorithmic Differentiation for adjoint sensitivity calculation in plasma edge codes

In the framework of fusion energy research, divertor design and model calibration based on plasma edge codes currently rely either on manual iterative tuning of parameters, or alternatively on parameter scans. This, combined with the complex and computationally expensive nature of plasma edge codes, makes these procedures extremely cumbersome. Gradient-based optimization methods can significantly reduce this effort, but require an efficient strategy for sensitivity calculation. Algorithmic Differentiation (AD) offers an efficient and accurate solution, allowing semi-automatic sensitivity computation in complex, continuously developed codes. The adjoint AD mode is especially attractive, as its cost is independent of the number of input parameters. In this paper, adjoint AD sensitivity calculation is deployed for the first time in plasma edge codes, applying the TAPENADE AD tool to SOLPS-ITER. Adjoint AD results are verified to be machine precision accurate compared to tangent AD mode, and up to 10−9 compared to finite differences. Scalings of AD computational efforts prove the advantages of adjoint compared to tangent AD, while memory requirements rapidly increase for adjoint, showing the need for an improved checkpointing strategy. With the new tool, a wealth of information becomes accessible to the modeler. An adjoint sensitivity analysis for a COMPASS density scan identifies the input parameters with largest impact on the solution, and 2D sensitivity maps show their spatial dependence.

Heat Transfer Characteristics Evaluation of Multi-Elbow Cooling System for Fusion Divertor under One-Sided Heating

Heat Transfer Characteristics Evaluation of Multi-Elbow Cooling System for Fusion Divertor under One-Sided Heating

Unusual ablation behaviors of tungsten graded network reinforced copper composites synthesized by 3D printing and infiltration sintering at ultra-high temperatures

Tungsten network reinforced functionally graded copper composites were prepared using an integrated 3D printing and infiltration technology to improve high temperature ablation properties of nuclear fusion divertor materials. The composites' ablation behavior was systematically investigated using an argon arc welding torch. Results showed that mass loss of composites was increased firstly but then decreased with the increase of ablation time. During ablation, evaporation of the melted copper reduces the local temperature of tungsten skeleton, whose high melting point also restrict the further evaporation of copper. These mechanisms significantly improve the ablative resistance of CuW gradient composite. Tungsten twinning structures were identified to form during ultra-high temperature (at 6273.15 K) ablation process in a cold environment in air. The formation of distinctive W twinning structures and gradient distribution of copper and tungsten skeleton were identified as the key mechanisms for the significantly enhanced ablation performance of the graded composite.

Parallel transport modeling of linear divertor simulators with fundamental ion cyclotron heating *

The Material Plasma Exposure eXperiment (MPEX) is a steady state linear device with the goal to perform plasma material interaction studies at future fusion reactor relevant conditions. A prototype of MPEX referred as ‘Proto-MPEX’ is designed to carry out research and development related to source, heating and transport concepts on the planned full MPEX device. The auxiliary heating schemes in MPEX are based on cyclotron resonance heating with radio frequency (RF) waves. Ion cyclotron heating (ICH) and electron cyclotron heating in MPEX are used to independently heat the ions and electrons and provide fusion divertor conditions ranging from sheath-limited to fully detached divertor regimes at a material target. A hybrid particle-in-cell code- PICOS++ is developed and applied to understand the plasma parallel transport during ICH in MPEX/Proto-MPEX to the target. With this tool, evolution of the distribution function of MPEX/Proto-MPEX ions is modeled in the presence of (a) Coulomb collisions, (b) volumetric particle sources and (c) quasi-linear RF-based ICH. The code is benchmarked against experimental data from Proto-MPEX and simulation data from B2.5 EIRENE. The experimental observation of ‘density-drop’ near the target in Proto-MPEX and MPEX during ICH is demonstrated and explained via physics-based arguments using PICOS++ modeling. In fact, the density drops at the target during ICH in Proto-MPEX/MPEX to conserve the flux and to compensate for the increased flow during ICH. Furthermore, sensitivity scans of various plasma parameters with respect to ICH power are performed for MPEX to investigate its role on plasma transport and particle and energy fluxes at the target.

Be surface structures on W(110) and W(211): A DFT study

Beryllium and Tungsten are promising candidates for use as plasma facing materials (PFMs) in upcoming fusion reactors. Many complex and competing phenomena however complicate the understanding and development of materials that operate in the harsh conditions of the reactor. In particular, redeposition of Be onto the W divertor must be considered due to the expected erosion of the Be first wall under exposure to the plasma. It is known that a build up of Be on W allows for the formation of BeW alloys which can harm the longevity and performance of the fusion divertor. In an effort to understand the interaction of Be with W surfaces, a study of Be structures on W(110) and W(211) as a function of Be coverage has been carried out using Density Functional Theory. We have found that both W surfaces develop monolayers of Be characterized by a densely packed hexagonal structure. Below this monolayer coverage, Be structures are found to have two motifs of coverage, a densely packed hexagonal pattern punctuated by areas of low-density coverage. The structures found here produce work function values and trends in good agreement with experimental measurements.

Experimental study on CHF of one-side heated rectangular heat sink with regard to fusion divertor cooling

When the critical heat flux (CHF) is reached, the heat transfer efficiency rapidly decreases. Therefore, CHF must be considered as a potential hazard for a stable cooling system operation. In particular, securing the thermal margin of plasma-facing-components inside the tokamak loaded with ultra-high heat flux is directly related to system safety. According to this necessity, in this study, CHF of flat rectagular channel under one-side high-heat-load condition was experimentally explored. As a result of analyzing the effect of system parameters on CHF, it was observed that mass flow rate and sub-cooling had a proportional relationship with CHF, whereas pressure had an inverse relationship with CHF. This is because a high mass flow rate and high sub-cooling promote forced convection heat transfer and quickly remove vapors from the flow path, but an increase in pressure reduces liquid surface tension and latent heat, causing early boiling. In addition, the predicton performance of the existing sub-cooled flow boiling CHF correlations was evaluated under the one-side high heat load condition. However, the large difference in experimental conditions and the structural difference between circular tube and rectangular channel made them under-predict the experimental values. Accordingly, we developed new optimized CHF correlations using a Python code grafted with machine learning technology. The developed correatlion can be utilized when evaluating the thermal safety margin of the future tokamak, and can be a great help when establishing a cooling system operation strategy.

Heat Transfer Potential of Unidirectional Porous Tubes for Gas Cooling under High Heat Flux Conditions

To discuss a suitable porous structure for helium gas cooling under high heat flux conditions of a nuclear fusion divertor, we first evaluate effective thermal conductivity of sintered copper-particles in a simple cubic lattice by direct numerical heat-conduction simulation. The simulation reveals that the effective thermal conductivity of the sintered copper-particle highly depends on the contacting state of each particle, which leads to the difficulty for the thermal design. To cope with this difficulty, we newly propose utilization of a unidirectional porous tube formed by explosive compression technology. Quantitative prediction of its cooling potential using the heat transfer correlation equation demonstrates that the heat transfer coefficient of the helium gas cooling at the pressure of 10 MPa exceeds 30,000 W/m2/K at the inlet flow velocity of 25 m/s, which verifies that the unidirectional porous copper tubes can be a candidate for the gas-cooled divertor concept.

Predicting the onset of nucleate boiling heat flux of one-side heated flat heat sink in sub-cooled flow conditions for fusion divertor cooling

For the cooling system of the future, nuclear fusion tokamak, to operate stably and continuously, it is important to identify potential hazards that may occur in the system in advance. Among the various hazards associated with the nuclear fusion tokamak, the onset of nucleate boiling (ONB) is a point at which the heat-transfer mechanism changes dramatically and is a crucial factor that must be addressed. In particular, the equipment inside the tokamak is loaded with a heat flux of several MW m−2 under single-side heating conditions, and it is important to predict the ONB under these special heating conditions. Therefore, in this study, the ONB of a flat heat sink was experimentally investigated under highly subcooled flow conditions. Based on the physical understanding of the thermo-hydraulic aspect of the ONB, the wall temperature gradient change point, which is mainly used in the subcooled flow condition, was selected as the ONB detection criterion. Trends in the ONB heat flux change were analyzed as representative system parameters that can be tuned in the cooling system, such as subcooling, mass flow rate, and pressure. In addition, the ONB correlations developed in the previous studies were evaluated for predicting the performance under one-side high heat load conditions. However, the large difference in the experimental conditions (range of system parameters and heating conditions) and the fact that the influence of system parameters was not reflected in the correlation resulted in high error rates. Therefore, the ONB correlation that can be used in the tokamak heat flux condition was newly developed through a dimensional analysis that can effectively reflect the influences in the correlation through dimensionless numbers. The developed correlation can be of great help in designing a diverter or blanket cooling system and establishing an operational strategy.

Factors Affecting the Coarsening and Agglomeration of Alumina Particles in Cu–Al2O3 Composites Prepared with Internal Oxidation

Alumina dispersion strengthened copper (DS‐Cu) shows excellent properties regarding mechanical behavior, thermal stability, thermal conductivity, and radiation resistance, making it a promising candidate of heat sink materials for fusion divertor applications. Compared to mechanical alloying, traditional internal oxidation route has merits of reduced contamination and capability of mass production. However, the tailoring of oxide particles, either intra‐ or intergranular, is essential to further improve the properties of DS‐Cu fabricated through internal oxidation route. Herein, the origins of coarsening and agglomeration of alumina particles are uncovered through a dedicated combining investigation between powder metallurgy and Rhines‐pack method, and contributing factors, such as grain boundaries, diffusion distance, concentration of Al, atomization, and the existence of water vapor, are identified. The effects of Ti doping on tailoring the size, shape, and distribution of intergranular and intragranular particles are demonstrated also. This study is of significance for the further improvement of properties of DS‐Cu fabricated with the combination of powder metallurgy and internal oxidation.

Steady-State Mechanical Analysis for Target Assembly in the Material Plasma Exposure eXperiment Facility

The Material Plasma Exposure eXperiment (MPEX) steady-state linear plasma facility is currently under design at Oak Ridge National Laboratory to expose target specimens to fusion divertor regimes. The neutron-irradiated target is actively cooled and remote handled in the MPEX facility for conducting plasma-material–interaction (PMI) experiments. In this study, the steady-state stresses in the target and target assembly system are investigated using two-dimensional (2-D) and three-dimensional (3-D) models to provide expected stresses/strains under the heat loads to which various system components would be exposed during MPEX operation. The calculated temperatures from the 2-D axisymmetric mechanical model were found to be in excellent agreement with those from the full 3-D thermohydraulic model, providing a strong model validation. Numerical simulation results for the steady-state mechanical model indicate nonuniform distributions for the temperature, stress, and deformation within the critical components. For the initial design, the deformation results indicate possible gap openings between contacting surfaces below the plasma-facing materials. To reduce the possibility of interfacial gap opening, the target assembly was slightly changed and evaluated using the 2-D stress model. Numerical simulation results indicate that the interfacial gap openings can be minimized without drastically changing the entire target assembly. The stress-strain conditions for the target will be further used to assess the appropriate operation during MPEX experiments and gain insight into materials science phenomena during PMI.

Simultaneous Observation of Tungsten Spectra of W0 to W46+ Ions in Visible, VUV and EUV Wavelength Ranges in the Large Helical Device

Spectroscopic studies for emissions released from tungsten ions have been conducted in the Large Helical Device (LHD) for contribution to the tungsten transport study in tungsten divertor fusion devices and for expansion of the experimental database of tungsten line emissions. Tungsten ions are distributed in the LHD plasma by injecting a pellet consisting of a small piece of tungsten metal wire enclosed by a carbon tube. Line emissions from W0, W5+, W6+, W24+–W28+, W37+, W38+, and W41+–W46+ are observed simultaneously in the visible (3200–3550 Å), vacuum ultraviolet (250–1050 Å), and extreme ultraviolet (5–300 Å) wavelength ranges and the wavelengths are summarized. Temporal evolutions of line emissions from these charge states are compared for comprehensive understanding of tungsten impurity behavior in a single discharge. The charge distribution of tungsten ions strongly depends on the electron temperature. Measurements of emissions from W10+ to W20+ are still insufficient, which is addressed as a future task.

Design uncertainty for a HELIAS 5-B stellarator fusion power plant

With a lack of plasma disruptions and current-driven instabilities, stellarators are potentially an attractive option for a fusion power plant. Previous system studies have been performed to optimise a HELIAS (HELIcal-axis Advanced Stellarator) 5-B power plant using the systems code process, however these have been based around a single design point. In reality there is a lot of uncertainty extrapolating from present day devices and understanding. In this paper we study how this will affect the design by identifying eight uncertainty distributions on the input. We then perform parameter studies and Monte-Carlo based analysis to look at the impact on fusion power and divertor heat load. We find that the two uncertainties that have the largest impact on the fusion power are the helium primary coolant mechanical pumping power and the energy multiplication in the blanket and shield. Eighty-three per cent of our solutions are within a tolerable divertor heat load, however this is additionally influenced by the tungsten impurity levels. In order to stay below the density cut-off limit for Electron Cyclotron Resonance Heating, the confinement time needs to be enhanced relative to the ISS04 scaling relation to produce acceptable performance. By identifying the highest impact design parameters, we are able to highlight that research into the blankets should be prioritised to reduce overall design uncertainty.

Ion heating in the PISCES-RF liquid-cooled high-power, steady-state, helicon plasma device

Radio frequency (RF) driven helicon plasma sources are commonly used for their ability to produce high-density argon plasmas (n > 1019 m−3) at relatively moderate powers (typical RF power < 2 kW). Typical electron temperatures are <10 eV and typical ion temperatures are <0.6 eV. A newly designed helicon antenna assembly (with concentric, double-layered, fully liquid-cooled RF-transparent windows) operates in steady-state at RF powers up to 10 kW. We report on the dependence of argon plasma density, electron temperature and ion temperature on RF power. At 10 kW, ion temperatures >2 eV in argon plasmas are measured with laser induced fluorescence, which is consistent with a simple volume averaged 0D power balance model. 1D Monte Carlo simulations of the neutral density profile for these plasma conditions show strong neutral depletion near the core and predict neutral temperatures well above room temperatures. The plasmas created in this high-power helicon source (when light ions are employed) are ideally suited for fusion divertor plasma-material interaction studies and negative ion production for neutral beams.

Experimental Study on Critical Heat Flux of One-Side Heated Rectangular Channel with Regard to Fusion Divertor Cooling

Experimental Study on Critical Heat Flux of One-Side Heated Rectangular Channel with Regard to Fusion Divertor Cooling

Heat and particle load test facility using an applied-field MPD thruster for studying fusion divertor technology

A test facility for high heat and particle loads relevant to a fusion divertor has been developed using an applied-field magnetoplasmadynamic (AF-MPD) thruster. The AF-MPD thruster can provide high heat and particle fluxes to a target at a relatively low pressure with a relatively strong magnetic field. Several diagnostics for heat and particle fluxes such as a calorimeter and Langmuir probe were utilized to confirm that a heat flux of 4.0 MW m−2 and particle flux of 4.2 × 1022 m−2 s−1 were achieved using a 0.17 T permanent magnet. The electron temperature and electron density were estimated to be 4 eV and 1013 cm3, respectively, using optical emission spectroscopy based on the collisional-radiative model.

Elemental Characterization of Neutron-Irradiated Tungsten Using the GD-OES Technique

In reactor-relevant fusion divertor conditions, tungsten (W) will be used as an armor material due to its excellent thermal properties. It will be exposed to impurities from numerous sources, including ion implantation and mixing, neutron transmutation, low-Z plasma-facing-component (PFC) redeposition and codeposition of deuterium and tritium fuel, and trapped helium bubbles. The impurity plasma material–interaction effects are a concern because they can cause gradual degradation of the material and of plasma performance due to dust formation, fuel retention, and even changes to the thermal and mechanical properties of the W armor. It is crucial to measure the amount of impurities in W, and the glow discharge–optical emission spectroscopy (GD-OES) technique is exceptionally well suited for analysis of irradiated samples. GD-OES can measure a sample’s elemental composition by sputtering the surface of the sample, ionizing the eroded material, and measuring the optical emission of the excited atoms. In order for the GD-OES technique to be applied to neutron-irradiated tungsten samples, a mounting system for miniature samples was designed. The sample mounting and centering procedure was successful in measuring the depth distribution of control W and W alloy sample elemental concentrations. These control depth spectra will be used as elemental references for postirradiated samples. The residence time of surface layers was measured, a comparison of signals from different anodes was completed, and the influence of initial surface roughness or nonuniformity was understood. The depth distribution of an arc-welded W-0.4% rhenium (Re) alloy was measured to have a stable Re signal that was distributed evenly in the W matrix. The methods developed here will allow for quantification of impurities and transmutation amounts in neutron-irradiated W. GD-OES is a powerful tool but requires calibration and careful optimization of the parameters to obtain meaningful results.

Comparison of X-ray and neutron tomographic imaging to qualify manufacturing of a fusion divertor tungsten monoblock

Within a tokamak fusion energy device, the performance and lifespan of a divertor monoblock under high heat flux cycles is of particular interest. Key to this is the quality of manufacture, especially the material joining interfaces. Presented here is a comparative study between X-ray and neutron tomography to investigate the quality of manufactured monoblocks. Tungsten is a high attenuator of X-rays, thus X-ray tomography was performed on ‘region of interest’ samples where the majority of the tungsten armour was removed to reduce the attenuation path. Neutron tomography was performed on the full monoblock samples for non-destructive testing and on the ‘region of interest’ samples for direct comparison. Both techniques were shown to be capable of imaging the samples but having their own advantages and disadvantages relating to image accuracy and logistical feasibility. The techniques discussed are beneficial for either the research and development cycle of fusion component design or in quality assurance of manufacturing.

CFD-based shape optimization on cross-section of monoblock fusion divertor cooling channel for minimizing local heat flux

The vertical targets in fusion divertors are the main component which are exposed to very high heat flux generally exceeding 10 MW/m2. In addition, since the heat is non-uniformly imposed to the cooling channels by unique one-side heating, the local heat flux can easily reach up to 20 MW/m2. It is a very challenging engineering target even when considering the latest cooling technology. In this study, CFD-based shape optimization is applied in order to find the optimal channel cross-sectional shape for minimizing local heat flux on the cooling surface. For simplicity, 2-D thermal analysis is adopted as the first step and it estimates preliminary optimized shapes. In this 2-D analysis, heat transfer model is simply reflected on the boundary conditions by analytic method using heat transfer correlations. Then, the optimized shapes are validated using detailed 3-D CFD simulations based on Eulerian two-phase flow models with/without boiling heat transfer. Maximum heat flux on the cooling channel surface is defined as the objective function in the optimization process. Finally, it is found that the maximum heat flux and temperature in the cooling channel can be significantly reduced compared to the original circular channel shape by simple modification to the ellipse-like cross-section. Effects of boiling heat transfer and manufacturability are also discussed in this paper.