Experimental study on liquid metal free surface flow under magnetic and electric field for nuclear fusion

In the future application of nuclear fusion, the liquid metal flows are considered to be an attractive option of the first wall of the Tokamak which can effectively remove impurities and improve the confinement of plasma. Moreover, the flowing liquid metal can solve the problem of the corrosion of the solid first wall due to high thermal load and particle discharge. In the magnetic confinement fusion reactor, the liquid metal flow experiences strong magnetic and electric, fields from plasma. In the present paper, an experiment has been conducted to explore the influence of electric and magnetic fields on liquid metal flow. The direction of electric current is perpendicular to that of the magnetic field direction, and thus the Lorentz force is upward or downward. A laser profilometer (LP) based on the laser triangulation technique is used to measure the thickness of the liquid film of Galinstan. The phenomenon of the liquid column from the free surface is observed by the high-speed camera under various flow rates, intensities of magnetic field and electric field. Under a constant external magnetic field, the liquid column appears at the position of the incident current once the external current exceeds a critical value, which is inversely proportional to the magnetic field. The thickness of the flowing liquid film increases with the intensities of magnetic field, electric field, and Reynolds number. The thickness of the liquid film at the incident current position reaches a maximum value when the force is upward. The distribution of liquid metal in the channel presents a parabolic shape with high central and low marginal. Additionally, the splashing, i.e., the detachment of liquid metal is not observed in the present experiment, which suggests a higher critical current for splashing to occur.

Improved scaling law for the prediction of deuterium retention in beryllium co-deposits

Hydrogen isotope co-deposition with Be eroded from the first wall is expected to be the main fusion fuel retention mechanism in ITER. Since good fuel accounting is crucial for economic and safety reasons, reliable predictions of hydrogen isotope retention are needed. This study builds upon the well-established empirical De Temmerman scaling law [1] that predicts D/Be ratios in co-deposited layers based on deposition temperature, deposition rate, and deuterium particle energy. Expanding the data used in the original development of the scaling law with an additional dataset obtained with more recent measurements using a different technique to the original De Temmerman approach, allows us to obtain new values for free parameters and improve the prediction capabilities of the new scaling law. In an effort to improve the model even further, scaling with D2 background pressure was included and a new two-term model derived, describing D retention in low- and high-energy traps separately.

Beryllium erosion and redeposition in ITER H, He and D-T discharges

The Monte-Carlo code ERO2.0 was used to simulate steady-state erosion and transport of beryllium (Be) in the ITER main chamber. Various plasma scenarios were tested, including a variation of the main species (hydrogen, deuterium, helium), plasma conditions (density, temperature, flow velocity) and magnetic configurations. The study provides valuable predictions for the Be transport to the divertor, where it is expected to be an important contributor to dust formation and fuel retention due to build-up of co-deposited layers. The Be gross and net erosion rates provided by this study can help identifying first wall regions with potentially critical armour lifetime.

Post-Test Numerical Analysis of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions with the MELCOR Fusion Code

The validation of numerical tools employed in the analysis of incidental transients in a fusion reactor is a topic of main concern. KIT is taking part in this task providing both experimental data and by performing numerical analysis in support of the main codes used for the safety analyses of the Helium Cooled Pebble Bed (HCPB) blanket concept. In recent years, an experimental campaign has been performed in the KIT-HELOKA facility to investigate the behavior of a First Wall Mock-Up (FWMU) under Loss Of Flow Accident (LOFA) conditions. The aim of the experimental campaign was twofold: to check the expected DEMO thermal-hydraulics conditions during normal and off-normal conditions and to provide robust data for code validation. The present work is part of these validation efforts, and it deals with the analysis of the LOFA experimental campaign with the system code MELCOR 1.8.6 for fusion. A best-estimate methodology has been used in support of this analysis to ease the distinction between user’s assumptions and code limitations. The numerical analyses are here described together with their goals, achievements, and lesson learnt.

Experimental Investigation of a Helium-Cooled Breeding Blanket First Wall under LOFA Conditions and Pre-Test and Post-Test Numerical Analysis

The experimental investigation of a prototypical set-up simulating a loss of flow accident in a helium-cooled breeding blanket first wall mock-up under typical heat load conditions is presented. The experimental campaign reproduces the expected DEMO thermal-hydraulics conditions during normal and off-normal situations and aims at providing some insight into the fast transients associated with the loss of flow in the blanket first wall. The experimental set-up and the definition of the experimental matrix are discussed, including pre-test analysis performed in support of these activities. The major experimental results are discussed, and a procedure of using the acquired data for validating and calibrating the RELAP-3D model of the mock-up is introduced. All these activities contributed to the creation of a relevant theoretical and practical experience that can be used in further studies concerning incidental transients in real-plant scenarios in the framework of DEMO plant fusion safety activities.

Analysis of the beryllium stability under standard and critical operation in a fusion reactor

The paper provides data on the peculiarity of change in the structure, structural phase changes and destructions in beryllium resulting from interaction with a near-wall plasma of fusion facilities. Beryllium resistance under conditions of ITER operation was evaluated, which considers factors leading to possible partial melting and erosion of panels of the ITER first wall. It presents the modelling of a heat s distribution in element (”finger”) of the first wall at ”normal” and ”increased” heat flux of the ITER operation.

Technological Processes for Steel Applications in Nuclear Fusion

Plasma facing components for energy conversion in future nuclear fusion reactors require a broad variety of different fabrication processes. We present, along a series of studies, the general effects and the mutual impact of these processes on the properties of the EUROFER97 steel. We also consider robust fabrication routes, which fit the demands for industrial environments. This includes heat treatment, fusion welding, machining, and solid-state bonding. Introducing and following a new design strategy, we apply the results to the fabrication of a first-wall mock-up, using the same production steps and processes as for real components. Finally, we perform high heat flux tests in the Helium Loop Karlsruhe, applying a few hundred short pulses, in which the maximum operating temperature of 550 °C for EUROFER97 is finally exceeded by 100 K. Microstructure analyses do not reveal critical defects or recognizable damage. A distinct ferrite zone at the EUROFER/ODS steel interface is detected. The main conclusions are that future breeding blankets can be successfully fabricated by available industrial processes. The use of ODS steel could make a decisive difference in the performance of breeding blankets, and the first wall should be completely fabricated from ODS steel or plated by an ODS carbon steel.

EXPERIMENTAL RESEARCH AND WORK DEVELOPMENT ON THE HELIUM IONS LINAC

The paper provides a brief summary of the experimental research carried out at the present time at the materials science complex developed at the NSC KIPT. The main directions for the development of the work carried out have been determined. The complex is based on a linear accelerator of helium ions. The features and advantages of the accelerating structure of the accelerator, which is based on the principle of APF, are described. A technique was de-veloped and irradiation of candidate materials for the divertor and the first wall of the TNR was carried out. The damageability of the irradiated samples could create from 3 to 80 dpa.

Comparative Analysis of Lithium First Wall Concepts for Tokamak with Reactor Technologies

When developing the stationary fusion reactor, an unresolved issue is the design of its intra-chamber plasma-facing elements. It has now become obvious that among the materials conventionally used for intra-chamber elements, there are no solid structural materials that would meet the requirements for the long-term operation under the effect of the flux of fusion neutrons (14 MeV) with a density of ~1014 cm–2 s–1 and the heat flux with a power density of 10–20 MW/m2. An alternative solution to this problem is the use of liquid metals as a plasma-facing materials, and, first of all, the use of lithium, which has a low atomic number (low charge number Z). Other easily-melting metals are also considered, which have higher Z number, but lower saturation vapor pressure than lithium. This will make it possible to create the long-lived, heavy-to-damage and self-renewing surface of the intra-chamber elements, which will not contaminate the plasma. The main ideas of the alternative concept of the intra-chamber elements can be formulated based on the comprehensive analysis of the problems and requirements arising during the development of intra-chamber elements of the stationary reactor, for example, the DEMO-type reactor. The article presents the analysis of the possible design of the lithium-coated intra-chamber elements and discusses the main ideas of the lithium first wall concept for the tokamak with reactor technologies.