DIII-D RESEARCH OPPORTUNITIES FORUM FOR THE 2008 EXPERIMENTAL CAMPAIGN
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| Title | 414: Studies of ITER-like castellation in DIII-D: impact of castellation shape on fuel retention in gaps | ||
| Name: | Andrey Litnovsky ( |
Affiliation: | Forschungszentrum Juelich |
| Research Area: | Hydrogenic Retention | Presentation time: | Requested | Co-Author(s): | D. Rudakov (UCSD), V. Philipps (FZJ), C. Wong (GA), W. West (GA), R. Boivin(GA), N. Brooks (GA), P. Wienhold (FZJ), G. Sergienko (FZJ), R. Bastasz (SNL), J. Whaley (SNL), W. Wampler (SNL), J. Watkins (SNL), J. Brooks (ANL), T. Evans (GA), D. Whyte (UW), P. Stangeby (Univ. of Toronto), A.Mclean (Univ. of Toronto), J. Boedo (UCSD), R. Moyer (UCSD) |
| Description: | This experiment is aimed at investigating the carbon transport and fuel accumulation in the ITER-like tungsten castellated structures. The essential goal of the experiment is to test the optimized shaping of the castellation cells to minimize the fuel accumulation in gaps. | ||
| Experimental Approach/Plan: | The experiments performed in DIII-D in 2005 with gap samples exposed in the DIII-D divertor using the DiMES system have demonstrated the positive effect of elevated temperatures on the carbon deposition and fuel accumulation in gaps. The next step in minimization of the impurity and fuel transport into gaps is the optimization of the shape of castellation.
It is planned to expose two sets of castellation samples in the private flux region for series of identical ELMy H-Mode discharges. The exposure is to be made using the DiMES transport system. Tungsten castellation will be used for the experiment, since the tungsten is planned to be used as a plasma-facing material in ITER divertor. Two shapes of castellation will be used: conventional (rectangular) shape and the optimized (dome-like) shape. Exposure time of >40 seconds (>8 plasma discharges) is requested per each castellation set. | ||
| Background: | The armor of the first wall and divertor of ITER will be castellated by splitting it into small-size cells to maintain the durability under the thermal excursions during plasma operation. However, there are concerns about the impurity deposition and fuel accumulation in the gaps of castellated structures. Past research demonstrated that the fuel inventory may become an issue for ITER, given the difficulties on fuel removal. To address this problem, dedicated investigations are ongoing on several tokamaks.
The investigations on fuel retention in the gaps of castellated structures are presently recognized as an important activity of the ITPA Topical Group on Divertor and SOL and are the subject of IEA-ITPA Joint Experiments Program (Task DSOL 13). The optimization of the shape of castellation is the relatively natural way towards the minimization of the fuel retention in gaps. The understanding of the physical processes behind the transport into gaps is of significant importance. The flexible design of the castellation allows the studies of deposition patterns in both toroidal and poloidal gaps, to reveal the effect of the gap orientation. Another advantage of this experiment is that the exposure of the castellated samples in private flux region of the DIII-D divertor will be performed at the shallow angles of castellation with respect to magnetic field, similarly as expected in ITER. | ||
| Resource Requirements: | Machine Time: 2x1/2 day experiment
Number of neutral beam sources: 3 | ||
| Diagnostic Requirements: | all SOL and lower divertor diagnostics, DiMES TV, core Thomson scattering, CER. | ||
| Analysis Requirements: | SIMS, XPS, Ellipsometry, NRA, DEKTAK profiling (the most of analyses will be provided by FZJ) | ||
| Other Requirements: | Plasma shape and discharge parameters similar to those of experiment with molybdenum mirrors performed on September 8, 2006 | ||
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