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UKAEA-CCFE-PR(23)1882022
In 2019, the JET-ILW was equipped with a Shattered Pellet Injector (SPI) system with a wide capability to allow studies on the efficacy of shattered pellets in reducing the electro-magnetic and the thermal loads during disruptions and the avoidance/suppression of the formation of runaway electrons. The fully commissioned system became operationa…
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UKAEA-CCFE-PR(22)312022
Unusually high power loads on the Beryllium limiter caused by Neutral Beam re-ionisation, and much cooler divertor target surfaces were observed during the recent JET tritium campaign. As both phenomena are driven by scrape-off layer (SOL) physics, the SOL features of 72 tritium H-mode discharges and their deuterium references have been studied. …
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UKAEA-CCFE-CP(23)392021
A key aim of the 2021 JET deuterium-tritium (D-T) experiments was to demonstrate steady high fusion power (10-15MW) with the ITER-like Be/W first wall. Plasmas were developed using D, repeated with T to investigate and mitigate isotope effects, and run with D-T to maximise fusion power. Compared with high current (q95~3) ‘baseline’ plasmas, …
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UKAEA-CCFE-CP(21)072021
The disruption mitigation system at ITER will include four shattered pellet injectors (SPI), which will be dedicated to the mitigation of electro-magnetic loads (EML), thermal loads and the avoidance and suppression of runaway electrons. Recently the JETILW was equipped with an SPI with a wide capability. Specifically: pellet diameter d = [4.57, 8.…
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UKAEA-CCFE-CP(21)052021
Plasma Event TRiggering for Alarms (PETRA) is a flexible new system at JET for detecting plasma events to trigger exception handling responses. It provides a platform for all existing and new techniques which involve processing JET real-time data to identify when to change the trajectory of a plasma shot, and reduces the technical cost and risk in …
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UKAEA-CCFE-PR(20)892020
With a lack of plasma disruptions and current-driven instabilities, stellarators are potentially an attractive option for a fusion power plant. Previous system studies have been performed to optimise a HELIAS (HELIcal-axis Advanced Stellarator) 5-B power plant using PROCESS, however these have been based around a single design point. In reality …
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UKAEA-CCFE-CP(19)452019
Spherical Tokamaks offer a number of potential advantages for a future fusion power plant. They have a high ratio of thermal to magnetic field pressure (beta) and strong flows, either of which could result in reduced turbulence. Fewer Toroidal Field (TF) coils and a different geometry offers the potential for new methods of remote maintenance …
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UKAEA-CCFE-CP(18)122018
RACE has been developing a concept design for the remote maintenance system for the EUROfusion DEMO powerplant. Within the DEMO tokamak, tritium breeding blankets will require periodic replacement which is currently designed to utilize the upper vertical ports at the top of the vacuum vessel. This operation will be challenging due to the scale of t…
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UKAEA-CCFE-PR(19)072018
Ray-tracing techniques are applied to divertor filtered imaging, a diagnostic that has long been plagued by polluting reflected light features in metal walled fusion machines. A physically realistic surface reflection model is developed from a Cook-Torrence microfacet BRDF model. Camera calibration images of in-vessel point lights at JET are use…
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UKAEA-CCFE-PR(18)522018
Partial detachment is the desired regime for the baseline burning plasma scenario in ITER and next-step devices, as it allows to disspate the majority of the energy carried by charged particles through the scrape-off-layer (SOL) and thus avoids localised heat flux deposition in the divertor region. The COMPASS tokamak is equipped with an open diver…
Showing 1 - 10 of 10 UKAEA Paper Results
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