SOLPS-ITER simulations of a vapour box design for the linear device Magnum-PSI

A vapour box (VB) is a physical device currently being considered to reduce the high heat and particle fluxes typically impacting the divertor in tokamaks. This system usually consists of a series of boxes that retains neutral particles to increase the amount of collision events with the impacting plasma. The neutral particles come from recycling and recombination of the plasma, gas puffing inside the box and by the evaporation of a liquid metal, typically Li or Sn. Currently, an VB is being constructed for testing in the linear plasma generator Magnum-PSI, operated at DIFFER. Its modular design will allow for open (not enclosing the target) and closed (enclosing the target) configurations, as well as evaporating a liquid metal to create a vapour cloud inside the box. The experiments carried out with this device will investigate its capabilities to reduce the plasma flux towards the target. This work presents a numerical study performed with SOLPS-ITER about the effectiveness of the current VB design in its open configuration to retain neutrals and its effect on the plasma beam properties. This is a first step before validation against experiments and studying closed configurations to ensure that the VB can successfully operate in a wide range of plasma parameters. Simulations show that the VB is capable of retaining neutrals and reducing fluxes to the target without requiring additional gas puffing in High and Low plasma flux scenarios. When lithium is evaporated from inside the box, the hydrogen plasma is completely extinguished and replaced by a low temperature Li plasma with lower flux. The fraction of Li and Li+ transported upstream the VB is three orders of magnitude below the amount evaporated form the central box, as most of the lithium is condensed in the side boxes and another small portion (two orders of magnitude below the amount evaporated) is deposited on the target. The VB design in its open configuration can mitigate incoming plasma peak heat flux by 0.6 M W m − 2 , which represents a fraction of 75 % and 81 % for the High and Low flux scenarios. This effect is expected to be higher when a closed configuration is employed, which could result in a significant reduction of heat fluxes on the divertor of tokamaks once that this design is extrapolated to the toroidal geometry, with just a minimal amount of Li and Li+ reaching the core.</p

Nuclear data for fusion: Validation of typical pre-processing methods for radiation transport calculations

AbstractNuclear data form the basis of the radiation transport codes used to design and simulate the behaviour of nuclear facilities, such as the ITER and DEMO fusion reactors. Typically these data and codes are biased towards fission and high-energy physics applications yet are still applied to fusion problems. With increasing interest in fusion applications, the lack of fusion specific codes and relevant data libraries is becoming increasingly apparent. Industry standard radiation transport codes require pre-processing of the evaluated data libraries prior to use in simulation. Historically these methods focus on speed of simulation at the cost of accurate data representation. For legacy applications this has not been a major concern, but current fusion needs differ significantly. Pre-processing reconstructs the differential and double differential interaction cross sections with a coarse binned structure, or more recently as a tabulated cumulative distribution function. This work looks at the validity of applying these processing methods to data used in fusion specific calculations in comparison to fission. The relative effects of applying this pre-processing mechanism, to both fission and fusion relevant reaction channels are demonstrated, and as such the poor representation of these distributions for the fusion energy regime. For the natC(n,el) reaction at 2.0MeV, the binned differential cross section deviates from the original data by 0.6% on average. For the 56Fe(n,el) reaction at 14.1MeV, the deviation increases to 11.0%. We show how this discrepancy propagates through to varying levels of simulation complexity. Simulations were run with Turnip-MC and the ENDF-B/VII.1 library in an effort to define a new systematic error for this range of applications. Alternative representations of differential and double differential distributions are explored in addition to their impact on computational efficiency and relevant simulation results

High Flux Helium Irradiation of Dispersion-Strengthened Tungsten Alloys and Effects of Heavy Metal Impurity Layer Deposition

Tungsten has been chosen as the plasma-facing material (PFM) for the divertor region in ITER and also a candidate PFM for future plasma-burning nuclear fusion reactors. During fusion device operation, PFMs will be exposed to low-energy He irradiation at high temperatures, resulting in sub-surface bubbles and surface morphology changes such as pores and fuzz. Carbide dispersion-strengthened W materials may enhance the ductility of W, but their behavior under high flux He irradiation remains unclear. In this work, the response of dispersion-strengthened tungsten materials to high flux, low energy He irradiation at high temperature is examined. Tungsten alloyed with 1, 5, or 10 wt. % tantalum carbide or titanium carbide exposed to these conditions result in surface pores, coral-like feature growth and sub-surface helium bubbles. Reactor-relevant helium irradiation (5x10 26_ m-2_ fluence) combined with high powered laser pulses to simulate off-normal reactor events does not significantly alter the surface morphology, as the surface nanostructures appear stable and cracks are only observed on a localized region of one sample. However, specimens show the development of an impurity layer on the surface, likely impurity deposition from the sample holder during irradiation, resulting in a mixed material layer on the surface. Helium bubbles exist in this impurity layer, and obscure conclusions about helium interactions with the carbide dispersoids. Nonetheless, it is clear that the dispersoid microstructure limits He bubble formation and subsequent surface nanostructuring, attributed to the dispersoid composition.</p

Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices

Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m-2. The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature.</p

The EU strategy for solving the DEMO exhaust problem

Exhaust of power and particles is crucial for the DEMO device and the EU has developed a strategy to address the challenges. This strategy consists of a conventional approach based on extrapolation of the ITER solution (detached lower single null divertor) as well as the development of alternatives as risk mitigation. These comprise alternative magnetic divertor geometry, liquid metal targets and intrinsically ELM-free operational scenarios. On the experimental side, the EUROfusion programme has initiated both upgrades to existing linear and toroidal devices as well as plans to engage in new devices presently under construction in the EU. In parallel, the theory and modelling efforts are ramped up in a targeted effort to obtain the necessary understanding for safe extrapolation to DEMO. This is especially important for the alternatives, which cannot be tested in ITER.</p

Performance of smoothing by spectral dispersion (SSD) with frequency conversion on the Beamlet Laser for the National Ignition Facility

Simulations and ongoing measurements indicate that SSD results in small degradation to the near field beam quality. The measured effect of SSD bandwidth on conversion to the third harmonic and smoothing of the target illumination will also be described

B2.5-Eunomia simulations of Magnum-PSI detachment experiments: I. Quantitative comparisons with experimental measurements

Detachment experiments have been carried out in the linear plasma device Magnum-PSI by increasing the gas pressure near the target. In order to have a proper detailed analysis of the mechanism behind momentum and power loss in detachment, a quantitative match is pursued between B2.5-Eunomia solutions and experimental data. B2.5 is a multi fluid plasma code and Eunomia is a Monte Carlo solver for neutral particles, and they are coupled together to provide steady-state solution of the plasma and neutral distribution in space. B2.5-Eunomia input parameters are adjusted to produce a close replication of the plasma beam measured in the experiments without any gas puffing in the target chamber. Using this replication as an initial condition, the neutral pressure near the plasma beam target is exclusively increased during simulation, matching the pressures measured in the experiments. Reasonable agreement is found between the electron temperature of the simulation results with experimental measurements using laser Thomson scattering near the target. The simulations also reveal the effect of increased gas pressure on the plasma current, effectively reducing the current penetration from the plasma source. B2.5-Eunomia is capable of reproducing detachment characteristics, namely the loss of plasma pressure along the magnetic field and the reduction of particle and heat flux to the target. The simulation results for plasma and neutrals will allow future studies of the exact contribution of individual plasma-neutral collisions to momentum and energy loss in detachment in Magnum-PSI.</p

Plasma detachment study of high density helium plasmas in the Pilot-PSI device

We have investigated plasma detachment phenomena of high-density helium plasmas in the linear plasma device Pilot-PSI, which can realize a relevant ITER SOL/Divertor plasma condition. The experiment clearly indicated plasma detachment features such as drops in the plasma pressure and particle flux along the magnetic field lines that were observed under the condition of high neutral pressure; a feature of flux drop was parameterized using the degree of detachment (DOD) index. Fundamental plasma parameters such as electron temperature (Te) and electron density in the detached recombining plasmas were measured by different methods: reciprocating electrostatic probes, Thomson scattering (TS), and optical emission spectroscopy (OES). The Te measured using single and double probes corresponded to the TS measurement. No anomalies in the single probe I–V characteristics, observed in other linear plasma devices [16, 17, 36], appeared under the present condition in the Pilot-PSI device. A possible reason for this difference is discussed by comparing the different linear devices. The OES results are also compared with the simulation results of a collisional radiative (CR) model. Further, we demonstrated more than 90% of parallel particle and heat fluxes were dissipated in a short length of 0.5 m under the high neutral pressure condition in Pilot-PSI

Response of tungsten surfaces to helium and hydrogen plasma exposure under ITER relevant steady state and repetitive transient conditions

The effect of helium (He) plasma exposure, and associated surface modifications, on the thermal shock resistance of tungsten (W) under ITER relevant steady state and transient heat and particle loads was studied. W samples were exposed to steady state and pulsed He plasmas at surface base temperatures from 670 to 1170 K. The same exposures were repeated in hydrogen (H) to allow a direct comparison of the role of the ion species on the thermal shock resistance. Exposure to He plasma pulses caused the formation of fine cracking network on W samples which occurred at a higher density and smaller depths compared to H pulsed plasma irradiation. The peak temperature reached during an ELM-like plasma pulse increased by a factor  1.45 over the 100 s of He plasma exposure, indicating a deterioration of the thermal properties. Transient loading experiments were also performed using a high power pulsed laser during He plasma exposure, showing a significant modification of the target thermal response caused by the surface damage. The effect of He-induced morphology changes on the thermal response modification was found to be very small compared to that of transient-induced damage