| 1. |
|
|
| 2. |
|
|
| 3. |
|
|
| 4. |
- Philipps, V., et al.
(författare)
-
Development of laser-based techniques for in situ characterization of the first wall in ITER and future fusion devices
- 2013
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 53:9, s. 093002-
-
Tidskriftsartikel (refereegranskat)abstract
- Analysis and understanding of wall erosion, material transport and fuel retention are among the most important tasks for ITER and future devices, since these questions determine largely the lifetime and availability of the fusion reactor. These data are also of extreme value to improve the understanding and validate the models of the in vessel build-up of the T inventory in ITER and future D-T devices. So far, research in these areas is largely supported by post-mortem analysis of wall tiles. However, access to samples will be very much restricted in the next-generation devices (such as ITER, JT-60SA, W7-X, etc) with actively cooled plasma-facing components (PFC) and increasing duty cycle. This has motivated the development of methods to measure the deposition of material and retention of plasma fuel on the walls of fusion devices in situ, without removal of PFC samples. For this purpose, laser-based methods are the most promising candidates. Their feasibility has been assessed in a cooperative undertaking in various European associations under EFDA coordination. Different laser techniques have been explored both under laboratory and tokamak conditions with the emphasis to develop a conceptual design for a laser-based wall diagnostic which is integrated into an ITER port plug, aiming to characterize in situ relevant parts of the inner wall, the upper region of the inner divertor, part of the dome and the upper X-point region.
|
|
| 5. |
- Rubel, Marek, et al.
(författare)
-
Efficiency of fuel removal techniques tested on plasma-facing components from the TEXTOR tokamak
- 2012
-
Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 87:5-6, s. 935-940
-
Tidskriftsartikel (refereegranskat)abstract
- An overview of several techniques considered for fuel and co-deposits removal is given. The methods were tested both on plasma-facing components from the TEXTOR tokamak and on laboratory-prepared layers: (a) chemical approach based on oxidative or nitrogen-assisted plasma; (b) photonic methods with laser-induced fuel desorption or ablation of co-deposits; (c) thermal desorption in vacuum or under oxidative conditions at a broad range of temperatures. The emphasis is on outstanding issues associated with every technique aiming at the reduction of fuel content: the efficiency of fuel and co-deposit removal, the surface state of PFC following the treatment and dust generation.
|
|
| 6. |
|
|
| 7. |
- Wauters, T., et al.
(författare)
-
Isotope removal experiment in JET-ILW in view of T-removal after the 2nd DT campaign at JET
- 2022
-
Ingår i: Physica Scripta. - : IOP Publishing. - 0031-8949 .- 1402-4896. ; 97:4
-
Tidskriftsartikel (refereegranskat)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 the plasma deuterium concentration to less than 1% in preparation for operation with tritium. This was also a key activity with regard to refining the clean-up strategy to be implemented at the end of the 2nd DT campaign in JET (DTE2) and to assess the tools that are envisaged to mitigate the tritium inventory build-up in ITER. The sequence began with 4 days of main chamber baking at 320 degrees C, followed by a further 4 days in which Ion Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) were applied with hydrogen fuelling, still at 320 degrees C, followed by more ICWC while the vessel cooled gradually from 320 degrees C to 225 degrees C on the 4th day. While baking alone is very efficient at recovering fuel from the main chamber, the ICWC and GDC sessions at 320 degrees C still removed slightly higher amounts of fuel than found previously in isotopic changeover experiments at 200 degrees C in JET. Finally, GDC and ICWC are found to have similar removal efficiency per unit of discharge energy. The baking week with ICWC and GDC was followed by plasma discharges to remove deposited fuel from the divertor. Raising the inner divertor strike point up to the uppermost accessible point allowed local heating of the surfaces to at least 800 degrees C for the duration of this discharge configuration (typically 18 s), according to infra-red thermography measurements. In laboratory thermal desorption measurements, maintaining this temperature level for several minutes depletes thick co-deposit samples of fuel. The fuel removal by 14 diverted plasma discharges is analysed, of which 9, for 160 s in total, with raised inner strike point. The initial D content in these discharges started at the low value of 3%-5%, due to the preceding baking and conditioning sequence, and reduced further to 1%, depending on the applied configuration, thus meeting the experimental target.
|
|
| 8. |
- Zayachuk, Y., et al.
(författare)
-
Fuel desorption from JET-ILW materials : assessment of analytical approach and identification of sources of uncertainty and discrepancy
- 2023
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 63:9, s. 096010-
-
Tidskriftsartikel (refereegranskat)abstract
- This work was carried out to identify sources of errors, uncertainties and discrepancies in studies of fuel retention in wall components from the JET tokamak using methods based on thermal desorption. Parallel aims were to establish good practices in measurements and to unify procedures in data handling. A comprehensive program designed for deuterium quantification comprised the definition and preparation of two types of materials (samples of JET limiter Be tiles and deuterium-containing targets produced in the laboratory by magnetron-assisted deposition), their pre-characterization, quantitative analyses of the desorption products in three different thermal desorption spectroscopy systems and a detailed critical comparison of the results. Tritium levels were also determined by several techniques in samples from JET and in tritiated targets manufactured specifically for this research program. Facilities available for studies of Be- and tritium-contaminated materials from JET are presented. Apparatus development, future research options and challenges are discussed.
|
|
