A5068847143- Fusion materials and technologies
- Magnetic confinement fusion research
- Nuclear Materials and Properties
- Metal and Thin Film Mechanics
- Nuclear reactor physics and engineering
- Superconducting Materials and Applications
- Diamond and Carbon-based Materials Research
- Nuclear Physics and Applications
- Laser-Plasma Interactions and Diagnostics
- Plasma Diagnostics and Applications
- Ionosphere and magnetosphere dynamics
- Ion-surface interactions and analysis
- Semiconductor materials and devices
- Particle accelerators and beam dynamics
- Dust and Plasma Wave Phenomena
- Magnetic properties of thin films
- High-Energy Particle Collisions Research
- Silicon Carbide Semiconductor Technologies
- High-pressure geophysics and materials
- Semiconductor materials and interfaces
- High-Temperature Coating Behaviors
- Atomic and Subatomic Physics Research
- Metallurgical Processes and Thermodynamics
- Intermetallics and Advanced Alloy Properties
- Electrohydrodynamics and Fluid Dynamics
National Institute for Laser Plasma and Radiation Physics
2015-2024
Culham Science Centre
2019-2024
Culham Centre for Fusion Energy
2018-2024
United Kingdom Atomic Energy Authority
2021-2024
Royal Military Academy
2020
Laboratoire de Physique des Plasmas
2019
Institutul de Fizică Atomică
2017
Yahoo (United Kingdom)
2010
The provision of a particle and power exhaust solution which is compatible with first-wall components edge-plasma conditions key area present-day fusion research mandatory for successful operation ITER DEMO. work package plasma-facing (WP PFC) within the European programme complements laboratory experiments, i.e. in linear plasma devices, electron ion beam loading facilities, studies performed toroidally confined magnetic such as JET, ASDEX Upgrade, WEST etc. connection both groups done via...
As a licensed nuclear facility, ITER must limit the in-vessel tritium (T) retention to reduce risks of potential release during accidents, inventory being set at 1 kg. Simulations and extrapolations from existing experiments indicate that T-retention in will mainly be driven by co-deposition with beryllium (Be) eroded first wall, co-deposits forming divertor region but also possibly on wall itself. A pulsed Laser-Induced Desorption (LID) system, called Tritium Monitor, is designed locally...
Data on erosion and melting of beryllium upper limiter tiles, so-called dump plates (DP), are presented for all three campaigns in the JET tokamak with ITER-like wall. High-resolution images wall show clear signs flash ridge roof-shaped tiles. The melt layers move poloidal direction from inboard to outboard tile, ending last DP tile an upward going waterfall-like structure. Melting was caused mainly by unmitigated plasma disruptions. During ILW campaigns, around 15% 12376 pulses were...
Abstract After the second Deuterium–Tritium Campaign (DTE2) in JET tokamak with ITER-Like Wall (ILW) and full tritium campaigns that preceded followed after DTE2, a sequence of fuel recovery methods was applied to promote removal from wall components. The started several days baking main chamber walls at 240 °C 320 °C. Subsequently, superimposed Ion-Cyclotron Conditioning (ICWC) Glow Discharge (GDC) cleaning cycles deuterium. Diverted plasma operation deuterium different strike point...
Abstract The paper reports the first demonstration of in situ laser-induced desorption — quadrupole mass spectrometry (LID-QMS) application on a large scale fusion device performed summer 2023. LID-QMS allows direct measurements fuel inventory plasma facing components without retrieving them from device. diagnostic desorbs retained gases by heating 3 mm diameter spot wall using 1 ms long laser pulse and detects QMS. Thus, it can measure gas content at any position accessible to laser....
Abstract This paper presents a study on the dependence of ion temperature stiffness plasma main isotope mass in JET ITER-like wall and C discharges. To this aim, database H, D T shots is analyzed, including new dedicated shots, comparing experiments with lower higher power injected by NBI system. In order to characterize turbulence mass, three these discharges (two one D) same external heating scheme are studied detail interpreted gyrokinetic linear nonlinear simulations. The analysis...
Abstract We present an overview of results from a series L–H transition experiments undertaken at JET since the installation ITER-like-wall (JET-ILW), with beryllium wall tiles and tungsten divertor. Tritium, helium deuterium plasmas have been investigated. Initial in tritium show ohmic transitions low density power threshold for ( P LH ) is lower than ones densities, while we still lack contrasted data to provide scaling high densities. In there notable shift which minimum <?CDATA...
Abstract The recent deuterium–tritium campaign in JET-ILW (DTE2) has provided a unique opportunity to study the isotope dependence of L-H power threshold an ITER-like wall environment (Be and W divertor). Here we present results from dedicated transition experiments at JET-ILW, documenting tritium plasmas, comparing them with matching deuterium hydrogen datasets. From earlier it is known that as plasma isotopic composition changes deuterium, through varying deuterium/hydrogen concentrations,...
Abstract This paper reports the first experiment carried out in deuterium–tritium addressing integration of a radiative divertor for heat-load control with good confinement. Neon seeding was time D–T plasma as part second campaign JET its Be/W wall environment. The technical difficulties linked to re-ionisation heat load are reported T and D–T. compares impact neon on plasmas their D counterpart detachment, localisation radiation, scrape-off profiles, pedestal structure, edge localised modes global
Abstract JET returned to deuterium-tritium operations in 2023 (DTE3 campaign), approximately two years after DTE2. DTE3 was designed as an extension of JET's 2022-2023 deuterium campaigns, which focused on developing scenarios for ITER and DEMO, integrating in-depth physics understanding control schemes. These were evaluated with mixed D-T fuel, using the only remaining tritium-capable tokamak until its closure 2023. A core-edge-SOL integrated H-mode scenario developed tested D-T, showing...
Self-ion irradiation of pure tungsten with 2 MeV W ions provides a way simulating microstructures generated by neutron in components fusion reactor. Transmission electron microscopy (TEM) has been used to characterize defects formed samples ion irradiation. It was found that irradiated 0.85 dpa at relatively low temperatures develops characteristic microstructure dominated dislocation loops and black dots. The density size distribution these were estimated. Some the exposed self-ion then...
Abstract An assessment of the tritium (T) inventory in plasma facing components (PFC) during JET T and deuterium-tritium (DT) operations is presented based on most comprehensive ex situ fuel retention data set PFCs from 2015-2016 ILW3 operating period presented. The global 4.19 × 10 23 D atoms, 0.19% injected fuel. inner divertor remains region highest (46.5%). at end calculated as 7.48 22 atoms informative for accountancy, clean-up efficacy waste liability assessments. accumulation rate...
Abstract The Joint European Torus (JET) fusion reactor was upgraded to the metallic wall configuration in 2011 which consisted of bulk beryllium (Be) tiles main chamber and tungsten (W) W-coated CFC divertor (Matthews G.F. et al Phys. Scr. T148 014001). During each campaign, a series damages were observed; on upper dump plates (UDP) positioned top part vessel walls inner wall—mainly affecting guard limiters (IWGL). In both cases, it concluded that causes these unmitigated plasma disruptions....
Since 2011 the JET tokamak has been operated with a metal ITER-like wall (JET-ILW) including castellated beryllium limiters and lamellae-type bulk tungsten tiles in divertor.This allowed for large scale test of plasma-facing components (PFC).Procedures sectioning into single blocks castellation have developed.This facilitated morphology studies surfaces inside grooves after experimental campaigns 2011-2012 2013-2014.The deposition 0.4-0.5 mm wide is 'shallow'.It reaches 1-2 12 deep...
The fuel outgassing efficiency from plasma-facing components exposed in JET-ILW has been studied at ITER-relevant baking temperatures. Samples retrieved the W divertor and Be main chamber were annealed 350 240 °C, respectively. Annealing was performed with thermal desoprtion spectrometry (TDS) for 0, 5 15 h to study deuterium removal effectiveness nominal remained fraction determined by emptying samples fully of heating up 1000 775 °C,respectively. Results showed deposits having an...
The JET ITER-Like Wall experiment, with its all-metal plasma-facing components, provides a unique environment for plasma and plasma-wall interaction studies. These studies are of great importance in understanding the underlying phenomena taking place during operation future fusion reactor. Present work summarizes reports fuel retention divertor resulting from two first experimental campaigns Wall. deposition pattern after second campaign shows same trend as was observed campaign: highest...
Melting is one of the major risks associated with tungsten (W) plasma-facing components (PFCs) in tokamaks like JET or ITER. These are designed such that leading edges and hence excessive plasma heat loads deposited at near normal incidence avoided. Due to high stored energies ITER discharges, shallow surface melting can occur under insufficiently mitigated disruption so-called edge localised modes—power load transients. A dedicated program was carried out study physics consequences W...
The manuscript presents an overview of the erosion and deposition data in inner outer JET divertor observed during first three ITER-like wall campaigns (JET-ILW1, JET-ILW2, JET-ILW3). Erosion were studied using core samples cut out from tiles. For a similar general pattern was all campaigns: More than 60% total occurred upper region on tiles 0 1, where Be transported deposited scrape-off layer. High only tile 5. In JET-ILW2 3, together with high power fluxes at bottom 7. Additionally, peaks...
Abstract A sequence of fuel recovery methods was tested in JET, equipped with the ITER-like beryllium main chamber wall and tungsten divertor, to reduce plasma deuterium concentration less than 1% preparation for operation tritium. This also a key activity regard refining clean-up strategy be implemented at end 2nd DT campaign JET (DTE2) assess tools that are envisaged mitigate tritium inventory build-up ITER. The began 4 days baking 320 °C, followed by further which Ion Cyclotron Wall...
Abstract The required heating power, <mml:math xmlns:mml="http://www.w3.org/1998/Math/MathML" overflow="scroll"> <mml:mrow> <mml:msub> <mml:mi>P</mml:mi> <mml:mi>LH</mml:mi> </mml:mrow> </mml:msub> </mml:math> , to access the high confinement regime (H-mode) in tritium containing plasmas is investigated JET with ITER-like wall at a toroidal magnetic field of <mml:mi>B</mml:mi> <mml:mi>t</mml:mi> <mml:mo>=</mml:mo> <mml:mn>1.8</mml:mn> T and plasma current <mml:mi>I</mml:mi>...
JET components are removed periodically for surface analysis to assess material migration and fuel retention. This paper describes issues related handling procedures preparing samples analysis; in particular a newly developed procedure cutting beryllium tiles is presented. Consideration also given the hazards likely due increased tritium inventory activation from 14 MeV neutrons following planned TT DT operations (DTE2) 2017. Conclusions drawn as feasibility of post DTE2.