Simulations of liquid metal flows over plasma-facing component edges and application to beryllium melt events in JET

Navier-Stokes simulations of liquid beryllium flows over the straight edge of plasma-facing components are carried out in conditions emulating upper dump plate melting observed experimentally in JET. The results demonstrate the existence of three main hydrodynamic regimes featuring various degrees of downstream flow attachment to the underlying solid surface. Transitions between these regimes are characterized by critical values of the Weber number, which quantifies the relative strength of fluid inertia and surface tension, thereby providing a general stability criterion that can be applied to any instance of transient melt events in fusion devices. The predictive capabilities of the model are tested by comparing numerical output with JET data regarding the morphology of the frozen melt layers and the location of beryllium droplets splashed onto nearby vacuum vessel surfaces as a result of disruption current quench plasmas interacting with the solid beryllium tiles protecting the upper main chamber regions. Simulations accounting for the coupling between fluid flow and heat transfer confirm the key role played by re-solidification as a stabilizing process, as previously found through macroscopic melt dynamics calculations performed with the MEMOS-U code. The favourable agreement found between the simulations and the general characteristics of the JET beryllium upper dump plate melt splashing give confidence that the same approach can be applied to estimate the possibility of such mechanisms occurring during disruptions on ITER.

One-point calibration laser-induced breakdown spectroscopy for the quantitative analysis of EAST-like plasma-facing materials

Quantitative in-situ elemental analyses of the impure deposited layer on the plasma-facing components (PFCs) of magnetically confined fusion devices such as EAST are crucial and challenging. Laser-induced breakdown spectroscopy (LIBS)...

Real-Time Detection of Overloads on the Plasma-Facing Components of Wendelstein 7-X

Wendelstein 7-X (W7-X) is the leading experiment on the path of demonstrating that stellarators are a feasible concept for a future power plant. One of its major goals is to prove quasi-steady-state operation in a reactor-relevant parameter regime. The surveillance and protection of the water-cooled plasma-facing components (PFCs) against overheating is fundamental to guarantee a safe steady-state high-heat-flux operation. The system has to detect thermal events in real-time and timely interrupt operation if it detects a critical event. The fast reaction times required to prevent damage to the device make it imperative to automate fully the image analysis algorithms. During the past operational phases, W7-X was equipped with inertially cooled test divertor units and the system still required manual supervision. With the experience gained, we have designed a new real-time PFC protection system based on image processing techniques. It uses a precise registration of the entire field of view against the CAD model to determine the temperature limits and thermal properties of the different PFCs. Instead of reacting when the temperature limits are breached in certain regions of interest, the system predicts when an overload will occur based on a heat flux estimation, triggering the interlock system in advance to compensate for the system delay. To conclude, we present our research roadmap towards a feedback control system of thermal loads to prevent unnecessary plasma interruptions in long high-performance plasmas.

Short-term retention in metallic PFCs: modelling in view of mass spectrometry and LIBS

Based on the conventional model of hydrogen retention in plasma-facing components, the question of hydrogen outgassing during and after plasma exposure is addressed in relation to mass spectrometry and laser-induced breakdown sprectroscopy (LIBS) measurements. Fundamental differences in retention and release data acquired by LIBS and by mass spectrometry are described analytically and by modelling. Reaction-diffusion simulations are presented that demonstrate possible thermal outgassing effects caused by LIBS. Advantages and limitations of LIBS as a tool for analysis of short term retention are discussed.

Activities in Divertor Reflector and Linear Plates Using WCLL and HCPB Breeding Blanket Concepts

In fusion devices, such as European Demonstration Fusion Power Reactor (EU DEMO), primary neutrons can cause material activation due to the interaction between the source particles and the targeting material. Subsequently, the reactor’s inner components become activated. For safety and safe performance purposes, it is necessary to evaluate neutron-induced activities. Activities results from divertor reflector and liner plates are presented in this work. The purpose of liner shielding plates is to protect the vacuum vessel and magnet coils from neutrons. As for reflector plates, the function is to shield the cooling components under plasma-facing components from alpha particles, thermal effects, and impurities. Plates are made of Eurofer with a 3 mm layer of tungsten, while the water is used for cooling purposes. The calculations were performed using two EU DEMO MCNP (Monte Carlo N-Particles) models with different breeding blanket configurations: helium-cooled pebble bed (HCPB) and water-cooled lithium lead (WCLL). The TENDL–2017 nuclear data library has been used for activation reactions cross-sections and nuclear reactions. Activation calculations were performed using the FISPACT-II code at the end of irradiation for cooling times of 0 s–1000 years. Radionuclide analysis of divertor liner and reflector plates is also presented in this paper. The main radionuclides, with at least 1% contribution to the total value of activation characteristics, were identified for the previously mentioned cooling times.

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.

First plasma exposure of a pre-damaged ITER-like plasma facing unit in the WEST tokamak: procedure for the PFU preparation and lessons learnt

The evaluation of the impact of plasma-facing components (PFCs) damage on subsequent plasma operation is an important issue for ITER. During the first phase of operation of WEST, a few ITER like divertor plasma-facing units (PFUs) have been installed on the lower divertor. One PFU was pre-damaged under electron beam gun thermal loading, before its installation in WEST, and the subsequent evolution of the damage was studied after the WEST plasma exposure. This paper presents the procedure followed to get the pre-damaged PFU. It consists in the characterization of the response of tungsten samples representative of WEST PFU under high heat flux (HHF) loading, the selection of damage (namely small cracks, crack network, crack network and W melt droplets). Finally, according to the WEST plasma loading conditions, the blocks with damage within the PFU and the position of the pre-damaged PFU on the WEST lower divertor are attributed. The first results obtained after an initial plasma exposure in WEST lead to assess, as expected with regard to the heat loading conditions, that no major surface aspect modification was found. This result emphasized the possibility to implement as pre-damage some small local droplets of melted tungsten in a high heat loaded zone for a future WEST experimental campaign.

Recent progress in the assessment of irradiation effects for in-vessel fusion materials: Tungsten and copper alloys

The present contribution highlights results of the recent irradiation campaigns applied to screen mechanical properties of advanced tungsten and copper-based materials – main candidates for the application in the plasma-facing components (PFC) in the European DEMO, which has also been presented at 28th IAEA Fusion Energy Conference. The main challenges in the formulated irradiation programme were linked to: (I) assessment of the ductile-to-brittle transition temperature (DBTT) of newly developed tungsten-based materials; (ii) investigation of an industrial pure tungsten grade under high temperature irradiation, reflecting operational conditions in the high flux divertor region; (iii) assessment of the high temperature strength of CuCrZr-based alloys and composites developed to enable the extension of the operational window for the heat sink materials. The development and choice of the advanced materials is driven naturally by the need to extend the operation temperature/fluence window thereby enlarging the design space for PFCs. The obtained results helped identifying the prospective tungsten and copper-based material grades as well as yielded a number of unexpected results pointing at severe degradation of the mechanical properties due to the irradiation. The results are discussed along with the highlights of the microstructural examination. An outlook for near future investigations involving in-depth post-irradiation examination and further irradiation campaigns is provided.

SOLPS-ITER simulations of high power exhaust for CFETR divertor with full drifts

One of the major challenges for the GW-class Chinese Fusion Engineering Testing Reactor (CFETR) is to efficiently handle huge power fluxes on plasma-facing components (PFCs), especially the divertor targets. This work investigates the effects of two candidate radiation impurity species, argon (Ar) and neon (Ne), with two different divertor geometries (baseline and long leg divertor geometry) on the reduction of steady-state power load to divertor targets in CFETR by using the SOLPS-ITER code package with full drifts and kinetic description of neutrals. The modeling results show clearly that increasing the seeding rate of Ar or Ne with fixed fueling gas D2 injection rate reduces the target electron temperature and heat flux density for the baseline divertor geometry, which can be reduced further by higher D2 injection rate. With a high impurity seeding rate, partial detachment with steady-state power load at the divertor target below the engineering limit of 10 MWm-2 is demonstrated. In addition, the radiation efficiency for Ar is better than that for Ne. Increasing the divertor leg length reduces the electron temperature and heat load at the targets. This modeling, therefore, suggests that a long leg divertor design with Ar seeding impurity is appropriate to meet the CFETR divertor requirements.