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[en] In this work, Wendelstein 7-X (W7-X) is the world’s largest superconducting nuclear fusion experiment of the optimized stellarator type. In the first Operation Phase (OP1.1) helium and hydrogen plasmas were studied in limiter configuration. The heating energy was limited to 4 MJ and the main purpose of that campaign was the integral commissioning of the machine and diagnostics, which was achieved very successfully. Already from the beginning a comprehensive set of diagnostics was available to study the plasma. On the path towards high-power, high-performance plasmas, W7-X will be stepwise upgraded from an inertially cooled (OP1.2, limited to 80 MJ) to an actively cooled island divertor (OP2, 10 MW steady-state plasma operation). The machine is prepared for OP1.2 with 10 inertially cooled divertor units, and the experimental campaign has started recently.The paper describes a subset of diagnostics which will be available for OP1.2 to study the plasma edge, divertor and scrape-off layer physics including those already available for OP1.1, plus modifications, upgrades and new systems. In conclusion, the focus of this summary will be on technical and engineering aspects, like feasibility and assembly but also on reliability, thermal loads and shielding against magnetic fields.
[en] At DIII-D, lithium granules were radially injected into the plasma at the outer midplane to trigger and pace edge localized modes (ELMs). Granules ranging in size from 300 to 1000 microns were horizontally launched into H-mode discharges with velocities near 100 m/s, and granule to granule injection frequencies less than 500 Hz. While the smaller granules were only successful in triggering ELMs approximately 20% of the time, the larger granules regularly demonstrated ELM triggering efficiencies of greater than 80%. A fast visible camera looking along the axis of injection observed the ablation of the lithium granules. We used the duration of ablation as a benchmark for a neutral gas shielding calculation, and approximated the ablation rate and mass deposition location for the various size granules, using measured edge plasma profiles as inputs. In conclusion, this calculation suggests that the low triggering efficiency of the smaller granules is due to the inability of these granules to traverse the steep edge pressure gradient region and reach the top of the pedestal prior to full ablation
[en] The behavior of 16Cr3Al ODS steel (oxide dispersion strengthened steel), widely employed structural fusion material, under high-intensity laser radiation with intensity up to 10(14)W/cm(2) was investigated in air, helium and vacuum surrounding. Employed system was 65 fs laser at 804 nm, with applied pulse energy up to 5.25 mJ. Morphological effects were studied - cracking, crater parameters (depth, cross-section), LIPSS (laser-induced periodic surface structures) formation at the crater periphery, hydrodynamic effects, as well as chemical variations on the surface. Ablation thresholds were also determined for all three ambiences (for 100 applied pulses), and they were 0.30 J/cm(2), 0.23 J/cm(2) and 0.39 J/cm(2) in air, helium and vacuum, respectively. Plasma occurred in all experiments and it was most prominent in vacuum due to strongest laser-material coupling.
[en] One of the key missions of the Ion Cyclotron Resonant Heating (IRCH) system for WEST is to provide sufficient RF heating power in order to obtain a heat flux on the divertor target of 10 MW/m2 during 1000 s and 20 MW/m2 during a few tens of seconds. Based on the experience acquired in Tore Supra, the ICRH system has been upgraded for long pulse operation and Edge Localized Modes (ELM) resilience. To achieve this performance, three antennas have been designed through a European collaboration and are now under fabrication at CAS/ASIPP, at the Keye Company, Hefei, in China, within the framework of the Associated Laboratory IRFM-ASIPP. This paper describes the electrical and mechanical design of the antenna, together with the main manufacturing steps and leak test procedure used for validating the water-cooled components. Accessibility and maintenance studies on WEST have been performed with the help of virtual reality. The first ICRH antenna was delivered at Cadarache in July 2016, and is foreseen to be installed on WEST in 2017. (authors)
[en] In the framework of the European 'HORIZON 2020' research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). CEA-Saclay, with the support of Wigner-CR and IPP-CR, is in charge of one of the four Breeding Blanket (BB) concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The BB directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion. In this article, the equatorial outboard module of the HCLL reference DEMO BB ('advanced-plus' BB) concept is studied regarding thermal and mechanical behavior during normal and accidental conditions that led to modify the design and enhance the performances. A numerical approach based on the Finite Element Method (FEM) is followed. The methodology established, the assumptions done as well as the results obtained in each case are reported and critically analysed, scrutinizing some open issues on this 'advanced-plus' reference concept. (authors)
[en] For the future ITER (International Thermonuclear Experimental Reactor) operation needs, D-T fusion experiments named DTE2 will be conducted in 2020 at JET (Joint European Torus). During the DTE2 operations, 14 MeV fusion neutrons will be generated. One of the aims of the DTE2 neutronics experiments is to investigate the D-T neutron streaming along the penetrations of biological shield of the JET Torus Hall. The TRIPOLI-4 Monte-Carlo radiation transport code has been extensively used on radiation shielding analyses. The purpose of this study is to evaluate the variance reduction (VR) techniques of TRIPOLI-4 on the 14 MeV neutron streaming calculations for a JET-like maze entrance. A JET-like torus hall building of 40 m x 40 m x 4.2 m and a maze entrance of 11.8 m length in 5-section configuration of concrete were first modeled. The standard INIPOND VR method and the new developed VR method AMS (Adaptive Multilevel Splitting) were then presented and discussed. Using the single-step approach combined with AMS techniques, the TRIPOLI-4 calculations results including neutron flux and ambient dose equivalent maps, the VR performance, and the user-friendly AMS input of TRIPOLI-4 code are reported. (authors)
[en] The methodological approach employed for the neutronics in the PPPT (Power Plant Physics and Technology) programme of EUROfusion is presented. It encompasses development works on advanced computational tools and activities related to the nuclear design and performance evaluation of the DEMO power plant including safety, maintenance, and waste management issues. Development work is conducted on Monte Carlo codes, on the CAD geometry conversion for Monte Carlo simulations, and on coupled radiation transport and activation computation systems. The role of nuclear data for reliable DEMO neutronics design analyses and uncertainty assessments is also addressed. Specific examples of nuclear analyses are presented including breeder blanket and shielding analyses for the different DEMO blanket concepts as well as related activation, decay heat and shut-down dose rate analyses. (authors)
[en] Reduction of hydrogen content in deuterium-fueled fusion plasmas is important not only to avoid diluting reacting core deuterons but also so as to allow the optimization of hydrogen-minority-heating efficiency in those fusion devices which employ ion-cyclotron-radio-frequency heating systems. In EAST, the amount of hydrogen released from plasma-facing components has been shown to depend strongly on both their composition and their temperature. In this paper, as measured by thermal desorption spectroscopy, the hydrogen inventory in graphite – used in EAST as lower divertor material – has been determined to be >25 times larger than that of tungsten which comprises the upper divertor. This difference in hydrogen inventory is attributed mostly to the intrinsically porous nature of bulk graphite. Thus the main source of hydrogen release into EAST discharges was identified as the graphite tiles used in the lower divertor. The hydrogen content in EAST plasmas were clearly reduced by first employing a high-temperature vacuum baking of all graphite tiles and then renewing a 100–200 μm thick SiC coating before an EAST experimental run campaign. Subsequent active surface conditioning of all wall components with elemental silicon and then with elemental lithium were seem to again reduce the plasma hydrogen content significantly – with lithium proving to be more effective than silicon. Combining these several techniques, H/(H + D) levels as low as ~3% have been achieved in EAST discharges. Additionally, the effects of lithium thickness on H surface implantation and retention has been re-examined semi-quantitatively using data from a previous run campaign. These data suggest that relatively thick Li films coated on the first wall can effectively isolate the rich source of hydrogen stored in the porous bulk of the underlying graphite from the deuterium-fueled plasma so as to minimize hydrogen release. In conclusion, an operational maneuver whereby a diverted plasma is repetitively switched from an upper single null configuration to a lower single null configuration is presented. This switching maneuver has been shown to suppress hydrogen influx into a 35s-long EAST discharge by alternately mitigating the rise in divertor temperature.
[en] The equatorial visible/infrared Wide Angle Viewing System (WAVS) is one of the ITER key diagnostics for machine protection. It has to monitor the Plasma Facing Components (PFCs) by infrared thermography and visible imaging. Foreseen to be installed in 4 equatorial port plugs to maximize the coverage of divertor, first wall, heating antennas and upper strike zone, the WAVS is composed of 15 lines of sight and 15 optical systems transferring the light along several meters from the PFCs through the port plug and interspace up to the detectors located in the port cell. After a conceptual design phase led by ITER Organization, the design is being further developed through a Framework Partnership Agreement signed between the European Domestic Agency, Fusion for Energy, and a consortium gathering CEA, CIEMAT (with INTA as third party) and Bertin Technologies company. First the WAVS measurement specifications are presented. Secondly the description of the current design is given both for the in-vessel system and for the ex-vessel one. The on-going neutronic studies are depicted as well as the cameras and data acquisition system foreseen for the back-end of the diagnostic. (authors)
[en] During ITER operation, it is expected that the large panel of plasma-wall interactions triggers the production of dust particles from the tungsten and beryllium first wall. These particles can be loaded with tritium (T) used as a fuel for the fusion reaction; T inventory is of prime importance for the safety assessment, particularly when considering dispersible matters that could be released in case of a Loss Of Vacuum Accident (LOVA),. The impact of accidental inhalation of such particles has therefore to be evaluated. Yet, particles properties in terms of tritium inventory and general behaviour are strongly dependent on their physical and chemical characteristics, and large uncertainties remain on those parameters; It is therefore important to obtain and characterize relevant W dust samples for toxicity studies before ITER operation starts. We produced model tungsten (W) Nano-Particles (NPs) by two different methods with characteristics closer to the plasma-wall interaction processes than grinding (usually used to produce commercial NPs): (i) magnetron plasma sputtering followed by gas condensation as well as (ii) laser ablation. The two types of tungsten NPs obtained exhibit very different properties investigated by a large characterization techniques panel; but both sets show similarities with samples collected in tokamak or observed in dusty plasmas setups, therefore could potentially be expected in ITER. Producing samples with strong differences is an asset to lead a first global study, highlighting which parameters are key, whether it affects the tritium inventory or the toxicity for lung cells during in vitro studies. In a second phase, it will help precise the impact of ITER particles when their definition will clarify. The aim of this paper is therefore to study and describe the characteristics of different types of W dusts, this knowledge being mandatory for the next steps of the project dealing with tritium inventory in the described W dusts and their suspension in various liquid media for cell exposure for toxicity studies. (authors)