Development and testing of an additively manufactured lattice for DEMO limiters

In the conceptual design of EU-DEMO, damage to plasma-facing components under disruption events is planned to be mitigated by specific sacrificial limiter components. A new limiter concept has been proposed using lattice structures fabricated with tungsten powder by additive manufacturing techniques. The major potential benefits of using a lattice structure for limiters are the possibility to customise the thermal conductivity and structural compliance of the structure to manage temperatures and stress within material limits and lower sensitivity to crack propagation. This paper presents the results of the first investigations into the production, characterisation, and high heat flux testing of the lattices to assess their suitability for DEMO limiters. First stage prototypes have been manufactured from tungsten and tungsten tantalum mixed powder with two distinct laser power bed fusion processes, namely pulsed laser and continuous laser with heated bed. The samples are characterised in terms of mass, volume, density, extent of microcracks and voids, level of un-melted or partially melted particulates, texture and grain size, as well as tantalum segregation when applicable. High transient (0.25ms) heat load testing, with hydrogen plasma of energy density up to ~3 MJm-2 was carried out at KIPT on the QSPA Kh-50. These tests have shown that the energy absorbed by latticed targets preheated at 500°C is close to that absorbed by solid tungsten, suggesting that they may be used for limiter applications with the added advantage of adjustment of the heat transfer and stiffness performance by geometry design or material properties.

Study on onset of nucleate boiling with screw tube for fusion reactor application

The onset of nucleate boiling (ONB) is the point at which the heat transfer mechanism in fluids changes and is one of the thermo-hydraulic factors that must be considered when establishing a cooling system operation strategy. Because the high heat flux of several MW/m2, which is loaded within a tokamak, is applied under a one-side heating condition, it is necessary to determine a correlative relation that can predict ONB under special heating conditions. In this study, the ONB of a one-side-heated screw tube was experimentally analyzed via a subcooled flow boiling experiment. The helical nut structure of the screw tube flow path wall allows for improved heat transfer performance relative to smooth tubes, providing a screw tube with a 53.98% higher ONB than a smooth tube. The effects of the system parameters on the ONB heat flux were analyzed based on the changes in the heat transfer mechanism, with the results indicating that the flow rate and degree of subcooling are proportional to the ONB heat flux because increasing these factors improves the forced convection heat transfer and increases the condensation rate, respectively. However, it was observed that the liquid surface tension and latent heat decrease as the pressure increases, leading to a decrease in the ONB heat flux. An evaluation of the predictive performance of existing ONB correlations revealed that most have high error rates because they were developed based on ONB experiments on micro-channels or smooth tubes and not under one-side high heat load conditions. To address this, we used dimensional analysis based on Python code to develop new ONB correlations that reflect the influence of system parameters.

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.

Lifetime measurement of a particular deep crack failure on a flat-type divertor mockup under cyclic high heat flux loading conditions

Plasma facing components (PFCs) are key to enduring high heat flux (HHF) loading from high-temperature plasma in nuclear fusion reactors. Understanding their thermal-mechanical behavior and cracking failure mechanisms related to structural designs and fabrication technologies during high heat flux loading is of great significance for improving their servicing performance and R&D (Research and Development) levels. In this study, a particular deep cracking failure process on the tungsten layer of a flat-type divertor mockup during 1800 cycles of 10 MW m-2 HHF loadings is completely monitored and measured with a special improved digital image correlation (DIC) technique. It is found that the DIC measurement under the HHF loading environment is improved successfully to capture fine deformation and strain fields with a spatial resolution less than 0.35 mm so that field strain on a 1 mm thick copper interlayer and deep crack initiation at several microns scale on the tungsten layer are measured out. Based on both full field and local strain and displacement measurements of the target divertor mockup, the thermal mechanical behaviors from deformation to crack initiation and propagation are successfully measured and traced. It is revealed that for the baseline copper interlayer design of a flat-type divertor mockup, the accumulation of plastic strain in the copper interlayer during ratcheting damage induces enough tensile stress on the tungsten layer during HHF cycles, leading to cracking and fracture failures even in its elastic state earlier than the copper LCF lifetime. Current SDC-IC rules fail to cover this kind of ratcheting cracking failure mode in the design stage. New design models or mechanical validation rules to resolve this design blind spot should be established in the future.

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.

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.