Operating a full tungsten actively cooled tokamak: overview of WEST first phase of operation

WEST is an MA class superconducting, actively cooled, full tungsten (W) tokamak, designed to operate in long pulses up to 1000 s. In support of ITER operation and DEMO conceptual activities, key missions of WEST are: (i) qualification of high heat flux plasma-facing components in integrating both technological and physics aspects in relevant heat and particle exhaust conditions, particularly for the tungsten monoblocks foreseen in ITER divertor; (ii) integrated steady-state operation at high confinement, with a focus on power exhaust issues. During the phase 1 of operation (2017–2020), a set of actively cooled ITER-grade plasma facing unit prototypes was integrated into the inertially cooled W coated startup lower divertor. Up to 8.8 MW of RF power has been coupled to the plasma and divertor heat flux of up to 6 MW m−2 were reached. Long pulse operation was started, using the upper actively cooled divertor, with a discharge of about 1 min achieved. This paper gives an overview of the results achieved in phase 1. Perspectives for phase 2, operating with the full capability of the device with the complete ITER-grade actively cooled lower divertor, are also described

Towards a maintainable and high efficiency neutral beam system for future fusion reactors

International audienceAchievement of an economic fusion reactor imposes a high level of unprecedented requirements for the Neutral Beam (NB) systems; the first one is the nuclear safety constraints which imposes prerequisite on the ease of access of all injector components to provide a preventive (and curative) maintenance by remote handling while maintaining reactor operation. In addition, the challenge is to develop high power injectors with very high wall-plug efficiency (above 60 %) able to operate in stable conditions over several months. There is a significant gap to bridge with respect to the present NB systems, which are handicapped by a low efficiency and by complex and long maintenance operations. Evidence that this injector concept does not offer adaptations to cope with the reactor requirements makes it clear that a new concept has to be addressed. An injector concept with modular sources at ground voltage is proposed. The concept makes remote maintenance of the injector components easier where each source module can be replaced by a new one without breaking the vacuum and affecting injector conditioning. With the grounded and modular ion source, photoneutralization associated with energy recovery appears as the best route capable of attaining the reactor requirements. This concept of maintainable NB system would provide a high heating power with a wall-plug efficiency above 70% and unprecedented features such as the capacity of producing temporal and spatial modulation of the beam power for a better control of the plasma stability. Up to now, photoneutralization feasibility studies already carried out on reduced-scale prototypes have not highlighted any showstoppers. Continuation of R&D in the years to come can pave the way towards the achievement of a first full-scale high power cavity in the 1 MW range, and to the realization of a multi-amperes (~10 A) thin blade-like D-beam

TEMPERATURE EFFECTS ON THE OPTICAL-PROPERTIES OF AMORPHOUS HYDROGENATED SILICON NITROGEN ALLOYS PREPARED BY RF SPUTTERING

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Prediction Model for Identifying Computational Phenotypes of Children with Cerebral Palsy Needing Neurotoxin Treatments

Factors associated with neurotoxin treatments in children with cerebral palsy (CP) are poorly studied. We developed and externally validated a prediction model to identify the prognostic phenotype of children with CP who require neurotoxin injections. We conducted a longitudinal, international, multicenter, double-blind descriptive study of 165 children with CP (mean age 16.5 ± 1.2 years, range 12–18 years) with and without neurotoxin treatments. We collected functional and clinical data from 2005 to 2020, entered them into the BTX-PredictMed machine-learning model, and followed the guidelines, “Transparent Reporting of a Multivariable Prediction Model for Individual Prognosis or Diagnosis”. In the univariate analysis, neuromuscular scoliosis (p = 0.0014), equines foot (p peri/postnatal causes, p = 0.05) were linked with neurotoxin treatments. In the multivariate analysis, upper limbs (p p = 0.02), the presence of spasticity (p = 0.01), dystonia (p = 0.004), and hip dysplasia (p = 0.005) were strongly associated with neurotoxin injections; and the average accuracy, sensitivity, and specificity was 75%. These results have helped us identify, with good accuracy, the clinical features of prognostic phenotypes of subjects likely to require neurotoxin injections

Design and integration of femtosecond fiber Bragg gratings temperature probes inside actively cooled ITER-like plasma-facing components

International audienceMeasuring the temperature in plasma-facing components (PFCs) provides information both on plasma parameters in the divertor region and on the thermal stress experienced by PFCs. Fiber Bragg gratings (FBGs) are interesting candidates for this application because they are immune to electromagnetic interferences and their ability to be multiplexed allows an extended spatial coverage. Four fibers, each of them including eleven regenerated Bragg gratings, have been embedded in tungsten-coated graphite components and operated up to their signal-collapsing limit at 800°C. Extending the measurement range towards higher temperatures increases the sensitivity to plasma parameters and allows withstanding higher energy experiments. To overcome thermal limitations, the system is upgraded using femtosecond laser inscribed fibers. In addition to their outstanding thermal stability, femtosecond FBGs benefit from higher signal-to-noise ratios than regenerated FBGs. The paper addresses femtosecond FBGs design and issues relative to their integration inside the actively cooled ITER-like PFCs of the WEST tokamak. Gratings period and length is designed to increase the number of measurement spots to fourteen gratings per fiber, regularly distributed over 17cm, while ensuring robust detection even with strong thermal gradients (no overlapping or deformation of Bragg peaks). The system operates up to 1200°C with gradients reaching 200°C/mm perpendicularly and 40°C/mm in parallel to the fiber. FBGs are inserted in actively cooled ITER-like PFCs through a 2.5mm deep lateral groove localized at 5mm beneath the top of bulk tungsten mono-blocks. A PFC mock-up machined with a groove has been tested under HHF facility to assess the effect of the groove on mono-blocks thermal behavior. The test demonstrates that machined mono-blocks behave as expected from simulation and can withstand 20 MW/m 2 heat flux (i.e. 1200°C in the fiber) with 20% overheating as compared to intact mono-blocks

Langmuir probe measurements in the WEST tokamak

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CEA contribution to the ITER ICRH antenna design

The ITER ion-cyclotron range of frequency (ICRH) heating system is required to couple 20 MW of power in the frequency range 40-55 MHz for a large range of scenarios with Edge Localized Modes. To mitigate the associated risks, it is foreseen to design and install on ITER two port-plug antennas for a total of 20 MW coupled power on long pulse operation [1]. The CEA activity to this antenna design within the CYCLE consortium was focused on the Faraday screen design (Fig. 1) and associated radio frequency (RF) sheath modelling, the reflectometers design for edge density measurement in the antenna vicinity and a contribution to the remote handling/tooling of the antenna. This paper is an overview of each item and proposes some R&D activities on key component

WEST regular in-vessel inspections with the Articulated Inspection Arm robot

International audienceCurrently on fusion devices, diagnostics are mainly aiming at plasma analysis and control. However, operational and programmatic needs have appeared for regular in-vessel components monitoring during plasma campaign. Light robotics systems could meet this requirement and may be a way as well to replace human interventions to fix damaged in vessel components. To minimize the impact on machine operation, the robotic system has to be mini-invasive and compatible with operating conditions (vacuum, temperature, radiation…). To fulfill this goal, CEA has developed a multipurpose carrier able to be operated inside WEST vessel between plasma pulses. A prototype of this robot, called Articulated Inspection Arm (AIA), was tested in 2008 in Tore Supra vacuum vessel. A major upgrade was performed in 2014-2015 with the aim of converting this prototype into a reliable tool in support to WEST operation. During the WEST components manufacturing and installation (2014-2016), the robot was integrated and tested in the EAST Tokamak. Since 2017, the AIA has been regularly used during the WEST plasma campaigns. Movies provided by the embedded camera allow to assess the evolution of Plasma Facing Components surface state and the effects of plasma loads, runaways and disruptions. The robot operation was also very helpful to assess the needs for maintenance, to assist mechanical assembly without man entry and to perform diagnostics calibration under relevant conditions. The paper will detail lessons learned from the robot integration and use on the WEST Tokamak. Future developments for innovative embedded diagnostics will also be presented

Status of the ITER Ion Cyclotron H&CD system

The ongoing design of the ITER Ion Cyclotron Heating and Current Drive system (20 MW, 40-55 MHz) is rendered challenging by the wide spectrum of requirements and interface constraints to which it is subject, several of which are conflicting and/or still in a high state of flux. These requirements include operation over a broad range of plasma scenarios and magnetic fields (which prompts usage of wide-band phased antenna arrays), high radio-frequency (RF) power density at the first wall (and associated operation close to voltage and current limits), resilience to ELM-induced load variations, intense thermal and mechanical loads, long pulse operation, high system availability, efficient nuclear shielding, high density of antenna services, remote-handling ability, tight installation tolerances, and nuclear safety function as tritium confinement barrier. R&D activities are ongoing or in preparation to validate critical antenna components (plasma-facing Faraday screen, RF sliding contacts, RF vacuum windows), as well as to qualify the RF power sources and the transmission and matching components. Intensive numerical modeling and experimental studies on antenna mock-ups have been conducted to validate and optimize the RF design. The paper highlights progress and outstanding issues for the various system component

Science and technology research and development in support to ITER and the Broader Approach at CEA

In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented