The JET 2019-2020 scientific and technological programme exploited the results of years of concerted scientific and engineering work, including the ITER-like wall (ILW: Be wall and W divertor) ...installed in 2010, improved diagnostic capabilities now fully available, a major neutral beam injection upgrade providing record power in 2019-2020, and tested the technical and procedural preparation for safe operation with tritium. Research along three complementary axes yielded a wealth of new results. Firstly, the JET plasma programme delivered scenarios suitable for high fusion power and alpha particle (α) physics in the coming D-T campaign (DTE2), with record sustained neutron rates, as well as plasmas for clarifying the impact of isotope mass on plasma core, edge and plasma-wall interactions, and for ITER pre-fusion power operation. The efficacy of the newly installed shattered pellet injector for mitigating disruption forces and runaway electrons was demonstrated. Secondly, research on the consequences of long-term exposure to JET-ILW plasma was completed, with emphasis on wall damage and fuel retention, and with analyses of wall materials and dust particles that will help validate assumptions and codes for design and operation of ITER and DEMO. Thirdly, the nuclear technology programme aiming to deliver maximum technological return from operations in D, T and D-T benefited from the highest D-D neutron yield in years, securing results for validating radiation transport and activation codes, and nuclear data for ITER.
In magnetic confinement nuclear fusion devices, Tungsten (W) stands out as the most promising material for plasma-facing components, especially for divertor targets exposed to high neutron loads and ...high heat flux. However, the harsh operational conditions subject W undergoes significant degradation due to irradiation-induced defect production and irradiation-induced recrystallization, resulting in a modification of the original microstructure. These irradiation effects lead to a consequential deterioration in mechanical properties and component durability.
This study implements a neutron irradiation-induced recrystallization (NIIR) modeling framework to quantify the recrystallization fraction in the W section of the plasma-facing components (PFCs). The simulation results reveal that the rate of neutron irradiation damage significantly impacts NIIR, leading to varying durations for achieving complete recrystallization in W at elevated temperatures. Furthermore, the half recrystallization time of W experiences a substantial reduction under neutron irradiation, resulting in a thicker recrystallized layer in W-PFCs compared to those without neutron irradiation but at the equivalent heat flux conditions.
•Cluster dynamic and modified JMAK models are employed for W-PFMs neutron irradiation-induced recrystallization simulation.•Varying damage rates affect grain boundary mobility and driving force, leading to different recrystallization times.•Neutron irradiation accelerates recrystallization in the W divertor target, forming a thicker recrystallized layer.
A DEMOnstration fusion reactor (EU-DEMO) is currently being developed by the EUROfusion consortium, in line with the Horizon Europe research framework programme. This fusion device will be the first ...on a large scale to generate net electricity. The EU-DEMO reactor is anticipated to undergo a pulsed duty cycle under normal operating conditions, which could shorten the qualified lifetime of the main equipment due to the inevitable induced mechanical and thermal cycling. Furthermore, the plasma control strategy envisaged for the EU-DEMO reactor foresees the potential occurrence of planned and unplanned plasma over-power transients, which might harm the plasma-facing components structure. In light of this, it is essential to dispose of reliable means to predict the thermal–hydraulic performance of the Primary Heat Transport Systems (PHTSs). Given this background, the University of Palermo, in partnership with the DEMO Central Team (DCT), has embarked on a research programme to evaluate the thermal–hydraulic response of the DIVertor Plasma-Facing Components (DIV PFC) PHTS under normal and upset conditions. To this purpose, in order to capture all the relevant geometric, hydraulic and thermal features associated with both ex-vessel and in-vessel components, a detailed finite-volume model has been developed. Next, the analysis of the thermal–hydraulic performance of the DIV PFC PHTS has been conducted both under hypothetical steady-state conditions and during the standard DEMO power cycle. The scope of the activity has been to verify whether the current design of the PHTS is capable of withstanding the pulsating loads it is expected to undergo under nominal conditions. The study has followed a theoretical–computational methodology founded on the use of the thermal–hydraulic system code TRACE. The hypotheses, models and results of the study are presented and commented.
•Study on the behaviour of the EU-DEMO Divertor Plasma Facing Components cooling system.•Development of a finite-volume model, based on the requirements of the TRACE system code.•Comparison of model and analytical results and parametric analysis of the ramps.•Validation of the set-up TRACE model to catch the main thermal–hydraulic phenomena.
Supported by the world magnetic fusion research community, a series of International Symposia on Liquid metals Applications for fusion (ISLA) have been held biannually since 2010. The 7th edition ...(ISLA-7) was held for the period from 12 December through 16 December 2022, at Chubu University located in Kasugai, Aichi, Japan. For the first time in the history of this series of symposia, ISLA-7 was held in a hybrid fashion, due to the COVID-19 situation. The total number of the participants was 60, 34 out of whom attended the symposium in person, and the rest participated online. As to the presentation statistics, 29 papers were presented in person, whereas 21 presentations were delivered online but real-time by the presenters in China, Spain, the UK, and the USA. Both of the presentations delivered in person and online were recorded, and the video has been shared by all participants. These participants represent 11 countries: China, Czech, Italy, Japan, Latvia, Netherlands, Russia, Thailand, the UK, and the USA. All these numbers are among the largest in this series of symposia. Covered by these presentations are; in session-2, program overviews and liquid metal research review; in session-3, liquid metal flows, and MHD issues; in session-4, liquid metal facilities; in sessions-5 and 6, liquid metal experiments and modeling; in session-7, divertor physics and heat flux mitigation; in session-8, plasma and liquid metals interactions; in session-9 liquid metal plasma-facing components, erosion, and wettability. In addition, there were an opening session whereby several opening addresses were delivered and also a closing session where all technical session summaries were presented by the respective session chairs.
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•Li pellet injection provides uniform toroidal coverage compared to Li evaporation.•Li evaporation deposits large amount of Li in close proximity to the evaporator.•Hydrogenation of ...fresh graphite surfaces observed after He-fueled plasma exposure.•Li2CO3 and LiC6 observed in C 1 s and Li 1 s HR-XPS spectra of certain samples.
Lithium wall conditioning has improved the performance of many magnetic fusion devices. Li conditioning in the RFX-mod device was performed by: (1) a single Li pellet injector and (2) a multi-Li pellet injector during He plasma discharges; (3) a Li evaporator after He glow discharge cleaning and He plasma discharge exposure. This report compares the spatial and depth distributions of Li deposited on polycrystalline graphite witness samples at different locations in RFX-mod and the elemental and chemical compositions of the resulting surfaces. The sample surfaces were analyzed ex situ using secondary ion mass spectrometry (SIMS) and high-resolution X-ray photoelectron spectroscopy (HR-XPS). The results showed that Li pellet injection provided a relatively uniform toroidal coverage while Li evaporation produced highly localized Li deposition. A Li 1s HR-XPS peak at 56.5 eV binding energy (BE) characteristic of lithium-intercalated graphite was only observed with the sample exposed to Li evaporation. All of the samples exhibited a HR-XPS C 1s peak at 285.1–285.2 eV BE that is largely attributed to hydrogenated graphite. This finding suggests that hydrogenation of fresh graphite occurs during He plasma discharge exposures. Our results have implications for density control and the selection of Li conditioning techniques in magnetic fusion devices.
In this work we present the results of the LIBS surface mapping and depth profiling of the Frascati Tokamak Upgrade (FTU) tiles of the poloidal limiter, made of a Molybdenum alloy (TZM) and removed ...from the reactor after its last experimental campaign with fusion plasma (late 2019). The measurements were performed by using a spectrometer array that covers the range 200–780 nm and a high-resolution spectrometer (0.1 Å at 500 nm) covering a spectral window of 10 nm. By applying a depth profiling procedure, the contamination of the tiles by B, Ca, Cr, Fe, K, Li, Mn, Mo, Ni, Ti was observed in their surface layers. Contamination with H, O, N was instead observed in deeper layers after some laser shots. B was attributed by the preliminary boronization procedure of the first wall (FW), carried out with deuterated diborane (D2B6), although D was not detected on the surface. Li was detected as deposited layer coming from previous experiments with liquid metals limiters in FTU. Ca and K are believed to come from the manipulation of the tiles after FTU decommissioning or as residual impurities of the deposited together with Li. H, O and N, from exposure of the tiles to the environmental gases, Mo and Ti were detected as main constituents of the tiles but also as eroded material during exposure to fusion plasma. Cr, Fe, Mn and Ni as eroded materials from the vacuum vessel (VV). In all cases, the surface concentration and thickness of the contaminated layer strongly varies across the TZM tile, showing a tendency of major impurity accumulations in areas more distant from the reactor's equatorial section.
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•LIBS analysis of the internal components of the Frascati Tokamak Upgrade.•Influence of previous experiments with liquid Lithium limiter are explained•Interpretation of the origin of the chemical elements detected•Quantification of the superficial Lithium deposited on the limiter tiles
A Volumetric Neutron Source (VNS) is proposed as a compact tokamak to serve as a qualification testbed for fusion in-vessel components, breeding blanket in particular. The main goal is to determine ...the minimum size of the machine while maximizing the Neutron Wall Load (NWL). This paper discusses the design parameters and constraints involved in specifying the magnetic cage for VNS. The focus of the study is on sizing the TF (toroidal field) coils and related structures with the plasma magnetic field, as well as assessing how various design constraints influence the machine performance at higher magnetic field strengths. In particular, the trade-off between TF neutron heat load and dose and shielding efficiency. A very preliminary (and still evolving) design point for VNS is presented and discussed.
Experiments have been conducted in the DIII-D tokamak to explore the in-situ growth of silicon-rich layers as a potential technique for real-time replenishment of surface coatings on plasma-facing ...components (PFCs) during steady-state long-pulse reactor operation. Silicon (Si) pellets of 1 mm diameter were injected into low- and high-confinement (L-mode and H-mode) plasma discharges with densities ranging from 3.9-7.5×1019 m-3 and input powers ranging from 5.5-9 MW. The small Si pellets were delivered with the impurity granule injector (IGI) at frequencies ranging from 4-16 Hz corresponding to mass flow rates of 5-19 mg/s (1-4.2×1020 Si/s) at cumulative amounts of up to 34 mg of Si per five-second discharge. Graphite samples were exposed to the scrape-off layer and private flux region plasmas through the divertor material evaluation system (DiMES) to evaluate the Si deposition on the divertor targets. The Si II emission at the sample correlates with silicon injection and suggests net surface Si-deposition in measurable amounts. Post-mortem analysis showed Si-rich coatings containing silicon oxides, of which SiO2 is the dominant component. No evidence of SiC was found, which is attributed to low divertor surface temperatures. The in-situ and ex-situ analysis found that Si-rich coatings of at least 0.4-1.2 nm thickness have been deposited at 0.4-0.7 nm/s. The technique is estimated to coat a surface area of at least 0.94 m2 on the outer divertor. These results demonstrate the potential of using real-time material injection to form Si-enriched layers on divertor PFCs during reactor operation.
•Updated thermal-fluids correlations and performance curves for HEMJ and flat design•Thermal-fluids and thermal-structural performance curves based on numerical simulations•Flat design divertor and ...HEMJ have comparable thermal-fluids performance•Flat design with K-doped W-3%Re pressure boundary meets structural criteria, unlike HEMJ
Over the past decade, our group has investigated the thermal-fluid performance of the helium-cooled modular divertor with multiple jets (HEMJ) and a simplified “flat” design of the HEMJ for long-pulse magnetic fusion energy (MFE) reactors. Experimental studies were performed in a helium (He) loop at the prototypical pressure of 10 MPa, nearly prototypical He temperatures and incident heat fluxes using test sections made from stainless steel and tungsten alloys. Correlations for average Nusselt numbers and pressure loss coefficients were developed from the data and are used to validate computational fluid dynamics (CFD) models. This work presents updated thermal-fluids performance curves based on these correlations that estimate the maximum heat flux that can be accommodated by the plasma-facing surface and coolant pumping power requirements at prototypical operating conditions. Thermal-structural performance curves developed from ITER structural design criteria are introduced, which include protection against ductile and non-ductile failure, ratcheting fatigue, and creep fatigue. The performance design curves for these finger-type divertors demonstrate that the “flat” design, with a significantly less complicated geometry than the HEMJ, has thermal-fluid and thermal-structural performance comparable to the original HEMJ concept, and may be superior in terms of non-ductile failure criteria.
The provision of a particle and power exhaust solution which is compatible with first-wall components and edge-plasma conditions is a key area of present-day fusion research and mandatory for a ...successful operation of ITER and DEMO. The work package plasma-facing components (WP PFC) within the European fusion programme complements with laboratory experiments, i.e. in linear plasma devices, electron and ion beam loading facilities, the studies performed in toroidally confined magnetic devices, such as JET, ASDEX Upgrade, WEST etc. The connection of both groups is done via common physics and engineering studies, including the qualification and specification of plasma-facing components, and by modelling codes that simulate edge-plasma conditions and the plasma-material interaction as well as the study of fundamental processes. WP PFC addresses these critical points in order to ensure reliable and efficient use of conventional, solid PFCs in ITER (Be and W) and DEMO (W and steel) with respect to heat-load capabilities (transient and steady-state heat and particle loads), lifetime estimates (erosion, material mixing and surface morphology), and safety aspects (fuel retention, fuel removal, material migration and dust formation) particularly for quasi-steady-state conditions. Alternative scenarios and concepts (liquid Sn or Li as PFCs) for DEMO are developed and tested in the event that the conventional solution turns out to not be functional. Here, we present an overview of the activities with an emphasis on a few key results: (i) the observed synergistic effects in particle and heat loading of ITER-grade W with the available set of exposition devices on material properties such as roughness, ductility and microstructure; (ii) the progress in understanding of fuel retention, diffusion and outgassing in different W-based materials, including the impact of damage and impurities like N; and (iii), the preferential sputtering of Fe in EUROFER steel providing an in situ W surface and a potential first-wall solution for DEMO.