DIII-D RESEARCH OPPORTUNITIES FORUM FOR THE 2008 EXPERIMENTAL CAMPAIGN
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| Title | 356: Measuring the contribution of the main wall to tritium retention | ||
| Name: | Peter C. Stangeby ( |
Affiliation: | GA ,LLNL and U of Toronto |
| Research Area: | Hydrogenic Retention | Presentation time: | Not requested | Co-Author(s): | -- |
| Description: | The ability of DIII-D to affect a 2D wall system is highly valuable, not only because it makes possible quantifiable and interpretable measurements of the plasma-wall interaction, but also because it directly replicates ITER reference condition. When operating in LSN the toroidally symmetrical structure at the top of DIII-D protecting the upper, outer cryo pump, can be used as the wall since the gap to the separatrix can be made smaller than the other gaps, including to the non-toroidal limiters. When operating in USN, the lower divertor surfaces can be used similarly to affect a toroidally symmetrical wall at the bottom. The latter arrangement is especially useful since this region of DIII-D is especially well diagnosed, including uniquely DTS. This region also contains DiMES which permits material samples to be placed in the wall surface for one or more shots, with easy removal for surface analysis. The lower surfaces are extensively fitted with Langmuir probes for measuring the spatial distribution and intensity of plasma fluxes to the wall. Comprehensive spectroscopic diagnostics are concentrated in the lower region of DIII-D, enabling measurements of the plasma-surface interaction by deuterium and carbon emissions.
Specific questions that will be answered by these experiments and their interpretation: 1 What is the spatial distribution and magnitude of ion fluxes to the first-wall? and how do they depend on the size of the gap between the separatrix and the wall, i.e. the 2nd separatrix? 2 What is the strength of the impurity source and its spatial distribution? i.e. what is the gross erosion rate? and how does it depend on the size of the wall gap? 3 What are the spatial distributions and magnitudes of the net erosion, net deposition and codeposition rates at the wall? 4 How much of the material released from the walls ends up in codeposits in the divertor? and how much ends up in codeposits on the walls? where are the latter located? What is the relative importance of codeposition retention at the walls and divertor? 5 What is the spatial distribution and intensity of the charge exchange fluxes at the walls? 6 It is well known that for true SN there is a fast parallel flow in the SOL that convects low-Z impurities released from the walls, to the inner divertor, concentrating the codeposits there. If the 2nd Xpt is inside the vessel, does that flow still occur in the 1st SOL, i.e. the one between the 1st and 2nd separatrix? Presumably it does not occur in the 2nd SOL, i.e. the one outside the 2nd separatrix. What SOL flows exist there? What role do they play in the codeposition process? 7 Since the local ionization at the 2nd Xpt could be substantial, how much of the C sputtered from the surfaces subtending the 2nd SOL reaches the 1st SOL? How much reaches the wall proper? Is the wall still an important source of materials ending up in divertor codeposits when a 2nd Xpt is present? | ||
| Experimental Approach/Plan: | 1 The primary focus will be on the reference plasma conditions for ITER, namely high density and high power H-mode, with and without ELM suppression.
2 The preferred configuration will be USN with the lower divertor surfaces being used to constitute the wall surface. Magnetic configurations with and without a 2nd X-point at the bottom will be used. The gaps to the edge structure on the low field side (which are not toroidally symmetrical) will be made large enough to prevent any significant plasma interaction from occurring there. The gap to the toroidally symmetrical inner wall, however, will be one of the variables explored in the experiment. The principal variable to be explored will be the distance between the 1st and 2nd separatrices (case of two X-points inside the vessel) or the distance from the separatrix to the last closed flux surface, LCFS, defined by the outermost magnetic flux surface that touches the wall at the bottom (case of one X-point in the vessel). 3 The source rate of carbon entering the plasma would be measured using filterscopes and MDS CI, CII, etc emissivities. The spatial distribution of the C-source (gross erosion) would also be measured with the tangential tv, TTV, system in CI, CII, CD, etc light. 4 The net erosion, net deposition and codeposition will be measured using DiMES. Such measurements have been made extensively in DIII-D over a number of years but with DiMES located in the divertor, while here it will be used to make such measurements at the wall. 5 A special PPI head will be made which injects methane over one half of the head only, using the rest of the head to register the local deposition resulting from the injection for ex situ surface analysis. 6 The PPI will also be used to create emission plumes in CII, CIII, etc light which will be recorded by the DiMES viewing camera system (which views DiMES almost directly from above). Such impurity plumes can be used to measure the local SOL flow field, both parallel and cross-field. 7 13CH4 will be injected toroidally symmetrically into both USN and LSN configurations, both for the two X-point and one X-point configurations, to measure the large scale impurity transport patterns. | ||
| Background: | For PFCs protected by C or Be tiles the most important T retention process is codeposition. If W is used the most important process is diffusion to traps created throughout the bulk by 14 MeV neutron damage [Haasz, et al, US-BPO: ITER Summary/WG-1/Task 5/Topic #5, Sept 13 2007]. T retention is due to plasma interaction at (a) the divertor and (b) the main wall. In order to reduce the peak power load on the divertor target, ITER will operate at high density to achieve detachment. The ion flux to the main wall increases as ne2 or ne3 and for high density the total fluxes to the divertor and walls are comparable, potentially making comparable the contribution of divertor and wall to T retention. Generally the wall is a major source of the low-Z C or Be that ends up in the divertor codeposits and may become the dominant impurity source at high density. Bombardment of W-walls by T ions/atoms drives T into the n-damage traps via permeation resulting in total T retention that may be comparable to that in the divertor at high density. Since the total sputtering yields of C and Be are comparable and since, for a given temperature, the T/C and T/Be ratios in codeposits are also comparable, the codeposition retention for a C-wall and a Be-wall system will be roughly similar. Therefore DIII-D is suitable for assessing the basic aspects of T codeposition retention in ITER for a wall tiled with either C or Be.
The plasma interaction with the wall is less well studied and understood than the interaction with the divertor, largely due to the fact that, while the divertor is toroidally symmetrical (2D), the wall structures are usually not, making for a 3D problem which is scarcely tractable. The present proposal is to make measurements in DIII-D of the plasma interaction with the wall by exploiting the unique feature of DIII-D to be able to affect a 2D wall. The magnetic configuration in ITER will result in a plasma-wall interaction which is in fact primarily 2D since there is a strong tendency for a 2nd X-point to occur at the top. A toroidal limiter/target will have to be installed there in order to handle the high power load. Therefore the plasma interaction outside the divertor will tend to occur primarily in toroidal bands at the top of the vessel and at the entrance to the divertor at the bottom. Most of the wall proper may be located further out, magnetically, and may experience relatively little plasma interaction. With regard to T retention, the wall in ITER will therefore in large part be these toroidally symmetrical bands, although cx fluxes will reach the wall proper and this has to be included. In any case, so long as the plasma interaction is 2D, the interpretive analysis is tractable, even if the neutrals explore 3D solid structures. The 3D-EIRENE Monte Carlo code was successfully applied to a 2D OEDGE plasma background in C-Mod by S Lisgo to interpretively model neutral pressure measurements made in the 3D C-Mod divertor-structure. | ||
| Resource Requirements: | -- | ||
| Diagnostic Requirements: | -- | ||
| Analysis Requirements: | -- | ||
| Other Requirements: | -- | ||
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