| 1. |
- Litaudon, X., et al.
(författare)
-
Development of steady-state scenarios compatible with ITER-like wall conditions
- 2007
-
Ingår i: Plasma Physics and Controlled Fusion. - 0741-3335 .- 1361-6587. ; 49:12B, s. B529-B550
-
Tidskriftsartikel (refereegranskat)abstract
- A key issue for steady-state tokamak operation is to determine the edge conditions that are compatible both with good core confinement and with the power handling and plasma exhaust capabilities of the plasma facing components (PFCs) and divertor systems. A quantitative response to this open question will provide a robust scientific basis for reliable extrapolation of present regimes to an ITER compatible steady-state scenario. In this context, the JET programme addressing steady-state operation is focused on the development of non-inductive, high confinement plasmas with the constraints imposed by the PFCs. A new beryllium main chamber wall and tungsten divertor together with an upgrade of the heating/fuelling capability are currently in preparation at JET. Operation at higher power with this ITER-like wall will impose new constraints on non-inductive scenarios. Recent experiments have focused on the preparation for this new phase of JET operation. In this paper, progress in the development of advanced tokamak (AT) scenarios at JET is reviewed keeping this long-term objective in mind. The approach has consisted of addressing various critical issues separately during the 2006-2007 campaigns with a view to full scenario integration when the JET upgrades are complete. Regimes with internal transport barriers (ITBs) have been developed at q(95) similar to 5 and high triangularity, 3 (relevant to the ITER steady-state demonstration) by applying more than 30 MW of additional heating power reaching beta(N) similar to 2 at B(o) similar to 3.1 T. Operating at higher 6 has allowed the edge pedestal and core densities to be increased pushing the ion temperature closer to that of the electrons. Although not yet fully integrated into a performance enhancing ITB scenario, Neon seeding has been successfully explored to increase the radiated power fraction (up to 60%), providing significant reduction of target tile power fluxes (and hence temperatures) and mitigation of edge localized mode (ELM) activity. At reduced toroidal magnetic field strength, high beta(N) regimes have been achieved and q-profile optimization investigated for use in steady-state scenarios. Values of beta(N) above the 'no-wall magnetohydrodynamic limit' (beta(N) similar to 3.0) have been sustained for a resistive current diffusion time in high-delta configurations (at 1.2 MA/1.8 T). In this scenario, ELM activity has been mitigated by applying magnetic perturbations using error field correction coils to provide ergodization of the magnetic field at the plasma edge. In a highly shaped, quasi-double null X-point configuration, ITBs have been generated on the ion heat transport channel and combined with 'grassy' ELMs with similar to 30 MW of applied heating power (at 1.2 MA/2.7 T, q(95) similar to 7). Advanced algorithms and system identification procedures have been developed with a view to developing simultaneously temperature and q-profile control in real-time. These techniques have so far been applied to the control of the q-profile evolution in JET AT scenarios.
|
|
| 2. |
- Litnovsky, A., et al.
(författare)
-
Diagnostic mirrors for ITER : A material choice and the impact of erosion and deposition on their performance
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 1395-1402
-
Tidskriftsartikel (refereegranskat)abstract
- Metal mirrors will be implemented in about half of the ITER diagnostics. Mirrors in ITER will have to withstand radiation loads, erosion by charge-exchange neutrals, deposition of impurities, particle implantation and neutron irradiation. It is believed that the optical properties of diagnostic mirrors will be primarily influenced by erosion and deposition. A solution is needed for optimal performance of mirrors in ITER throughout the entire lifetime of the machine. A multi-machine research on diagnostic mirrors is currently underway in fusion facilities at several institutions and laboratories worldwide. Among others, dedicated investigations of ITER-candidate mirror materials are ongoing in Tore-Supra, TEXTOR, DIII-D, TCV, T-10 and JET. Laboratory studies are underway at IPP Kharkov (Ukraine), Kurchatov Institute (Russia) and the University of Basel (Switzerland). An overview of current research on diagnostic mirrors along with an outlook on future investigations is the subject of this paper.
|
|
| 3. |
- Loarer, T., et al.
(författare)
-
Gas balance and fuel retention in fusion devices
- 2007
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 47:9, s. 1112-1120
-
Tidskriftsartikel (refereegranskat)abstract
- The evaluation of hydrogenic retention in present tokamaks is of crucial importance to estimate the expected tritium (T) vessel inventory in ITER, limited from safety considerations to 350 g. In the framework of the European Task Force on Plasma Wall Interaction (EU TF on PWI) efforts are underway to investigate gas balance and fuel retention during discharges, and to compare the data obtained with those from post-mortem analysis of in-vessel components exposed over whole experimental campaigns. This paper summarizes the principal findings from coordinated studies on gas balance and fuel retention from a number of European tokamaks, namely, ASDEX-Upgrade (AUG), JET, TEXTOR and Tore Supra (TS). For most devices, the long-term retention fraction deduced from integrated particle balance is similar to 10-20%. This is larger than the similar to 3-4% deduced from post-mortem analysis of plasma facing components (PFCs). However, from the database available for tokamaks with their main PFCs made of carbon, the important conclusion is that the T inventory limit (set by the working guideline for operations) could be reached in ITER within fewer than 100 discharges. This, therefore, would seriously impact on operation of the device unless efficient T removal processes are developed.
|
|
| 4. |
- Matthews, G. F., et al.
(författare)
-
Current status of the JET ITER-like Wall Project
- 2009
-
Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847. ; T138, s. 014030-
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents an overview of the status and relevant technical issues for the ITER-like Wall Project with emphasis on progress since the 11th International Workshop on Plasma-Facing Materials and Components for Fusion Applications.
|
|
| 5. |
- Matthews, G. F., et al.
(författare)
-
Overview of the ITER-like wall project
- 2007
-
Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T128, s. 137-143
-
Tidskriftsartikel (refereegranskat)abstract
- Work is in progress to completely replace, in 2008/9, the existing JET CFC tiles with a configuration of plasma facing materials consistent with the ITER design. The ITER-like wall (ILW) will be created with a combination of beryllium ( Be), tungsten ( W), W-coated CFC and Be-coated inconel tiles, with the material depending on the local anticipated heat flux and geometry. It is part of an integrated package of JET enhancements whose aim is to develop an understanding of the ITER materials issues and develop the techniques required to operate with inductive and advanced scenarios as close as possible to ITER parameters. Over 4000 tiles will be replaced and the ILW will accommodate additional heating up to at least 50 MW for 10 s. This paper describes the scientific background to the project, the technical objectives, the material configuration selected, the R&D behind the practical realization of the objectives and the generic problems associated with the Be tiles ( power handling capacity and disruption induced eddy currents). One of the objectives is to maintain or improve the existing CFC tile power handling performance which has been achieved in most cases by hiding bolt holes, optimizing tile size and profile and introducing castellations on plasma facing surfaces.
|
|
| 6. |
- Murari, A., et al.
(författare)
-
New diagnostic techniques and technologies at JET (invited)
- 2006
-
Ingår i: Review of Scientific Instruments. - : AIP Publishing. - 0034-6748 .- 1089-7623. ; 77:10
-
Tidskriftsartikel (refereegranskat)abstract
- The European Fusion Development Agreement's mission for JET is the development of ITER scenarios exploiting the specific properties of the device. This task requires significant improvements in the measuring techniques. The most innovative diagnostic upgrades are in the fields of edge measurements, detection of fast magnetohydrodynamics modes and burning plasma diagnostics. The importance of plasma-wall interactions, and, in particular, the issue of tritium inventory promoted the development of the quartz microbalance, a detector with improved time resolution to measure material redeposition in the remote areas of the inner divertor. Measurement of Alfven cascades with unprecedented spectral resolution, reaching a toroidal n number of up to 16, was obtained using an O-mode microwave reflectometer as an interferometer. For the diagnosis of the fusion products, a new approach is being developed to measure the He ash based on double charge exchange between thermalized particles and neutrals from JET beams. There have been several upgrades of the neutron diagnostic systems, one of which, the new time of flight neutron spectrometer designed for high counting rates is described.
|
|
| 7. |
|
|
| 8. |
- Piazza, G., et al.
(författare)
-
R&D on tungsten plasma facing components for the JET ITER-like wall project
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 367, s. 1438-1443
-
Tidskriftsartikel (refereegranskat)abstract
- Currently, the primary ITER materials choice is a full beryllium main wall with carbon fibre composite at the divertor strike points and tungsten on the upper vertical targets and dome. The full tungsten divertor option is a possibility for the subsequent D-T phase. Neither of the ITER material combinations of first wall and divertor materials has ever been tested in a tokamak. To collect operational experience at JET with ITER relevant material combination (Be, C and W) would reduce uncertainties and focus the preparation for ITER operations. Therefore, the ITER-like wall project has been launched to install in JET a tungsten divertor and a beryllium main wall. This paper describes the R&D activities carried out for the project to develop an inertially cooled bulk tungsten divertor tile, to fully characterise tungsten coating technologies for CFC divertor tiles and to develop erosion markers for use as diagnostics on beryllium tiles.
|
|
| 9. |
- Ratynskaia, Svetlana V., et al.
(författare)
-
Capture by aerogel-characterization of mobile dust in tokamak scrape-off layer plasmas
- 2009
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 49:12
-
Tidskriftsartikel (refereegranskat)abstract
- The aim of this letter is to demonstrate the feasibility and potential of the novel in situ dust diagnostic method-capture by aerogel targets. Aerogel, a highly porous material with a density of a few tens of kg m(-3), allows capturing of dust particles present during the discharge without destroying them. The first exposures in the TEXTOR scrape-off layer plasma showed that such targets are able to capture both slow and fast particles with sizes in the range from submicrometre to similar to 100 mu m. The technique provides information on dust velocity and size distribution as well as dust flux estimates. The composition and texture of the captured dust can also be studied in detail to shed light on dust formation processes.
|
|
| 10. |
- Badziak, J., et al.
(författare)
-
Application of ion diagnostics to control the laser-induced removal of surface layer of a carbon substrate
- 2006
-
Ingår i: Plasma 2005. - : AIP. - 073540304X ; , s. 295-298
-
Konferensbidrag (refereegranskat)abstract
- Among other methods of detritiation of in-vessel tokamak components the application of lasers for removal of fuel trapped in co-deposited layers is under investigation. The paper presents preparation and tests of ion diagnostic methods for on-line measurement of the amount and characteristics of ablated carbon, hydrogen/deuterium and contaminant species from the graphite target (plate) of the main toroidal limiter of the TEXTOR tokamak. For removal of the surface layer from the graphite limiter plate Nd:YAG laser was used. Determination of the characteristics of laser-produced ions has been performed by means of ion collectors and an electrostatic ion-energy analyser. The main ion stream parameters were measured depending on the number of laser shots and the laser power density on the target surface. The properties of modified carbon sample surface were determined with the use of optical methods and compared with the results of the ion measurements.
|
|
