Tritium removal from TFTR

Mueller, D.; Blanchard, W.; Collins, J.; Hosea, J.; Kamperschroer, J.; LaMarche, P. H.; Nagy, A.; Owens, D. K.; Skinner, C.H.

doi:10.1016/s0022-3115(97)80162-6

Cited by 31 publications

(19 citation statements)

References 11 publications

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“…Larger retention rates would mean more frequent stoppages to effect reduction of the in-vessel T through some method. A recent analysis of the database of retention measurements from current carbon PFC machines [2], including T retention experiments in TFTR and JET [3][4][5][6], led to the conclusion that ITER retention rate will be higher than the requirement introduced above but with quite large uncertainties. The prediction of T retention is thus a central focus of edge plasma research as it has important implications for the operation of ITER and reactors.…”

Section: Introductionmentioning

confidence: 99%

Hydrogenic retention with high-Z plasma facing surfaces in Alcator C-Mod

et al. 2009

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Abstract. The retention of deuterium (D) fuel in the Alcator C-Mod tokamak is studied using a new 'static' gas balance method. C-Mod solely employs high-Z molybdenum (Mo) and tungsten (W) for its plasma-facing materials, with intermittent application of thin boron (B) films. The primarily Mo surfaces are found to retain large fractions, ~20-50%, of the D 2 gas fuelled per quiescent discharge, regardless if the Mo surfaces are cleaned of, or partially covered by, B films. Several experiments and calculations show that it is improbable that B is retaining significant fractions of the fuel. Rather, the retention is occurring in Mo and W surfaces through ion bombardment, implantation and diffusion to trap sites. Roughly 1% D of incident ion fluence, D , to surfaces is retained, and with no indication of the retention rate decreasing after 25 s of integrated plasma exposure. The magnitude of retention is significantly larger than extrapolated from the results of laboratory studies for either Mo or W. The high levels of D/Mo in the near surface, measured directly post-campaign (~ .01) in tiles and inferred from gas balance, are consistent with trapping sites for fuel retention in the Mo being created, or expanded, by high D atom densities in the near surface which arise as a result of high incident ion fluxes. Differences between C-Mod and laboratory retention results may be due to such factors as the multiply-ionized B ions incident on the surface directly creating traps, the condition of the Mo (impurities, annealing), and the high flux densities in the C-Mod divertor which are similar to ITER, but 10-100x those used in laboratory studies. Disruptions produce rapid heating of the surfaces, releasing the trapped hydrogenic species into the vessel for recovery. The measurements of the large amount of gas released in disruptions are consistent with the analysis of tiles removed from the vessel post-campaign -the campaign-integrated retention is very low, of order 1000x less than observed in a single, non-disruptive discharge.

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Section: Introductionmentioning

confidence: 99%

Hydrogenic retention with high-Z plasma facing surfaces in Alcator C-Mod

et al. 2009

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“…As one among several options [1,4], oxygen glow discharges have been proposed [5,6] as a means of detritiation. The technique was tested in several tokamaks [7][8][9][10].…”

Section: Introductionmentioning

confidence: 99%

“…The major advantages of low-temperature oxygen plasma cleaning in a tokamak are that-as opposed to the thermal combustion of the redeposited layers in oxygen-no elevated temperatures are required, that simple means of plasma generation can be used such as DC [7,10,20] or ion cyclotron resonance [9,21,22] glow discharges and that high rates are achievable. The mentioned laboratory studies focussed on the erosion of laboratory-deposited a-C:H under controlled conditions.…”

Section: Introductionmentioning

confidence: 99%

Oxygen glow discharge cleaning in nuclear fusion devices

Hopf

Jacob

Rohde

2008

Journal of Nuclear Materials

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Laboratory experiments were carried out as part of the preparations of an oxygen glow discharge cleaning experiment in ASDEX Upgrade. They aimed at evaluating the effect of mixing the oxygen with helium, the collateral damage caused by the glow discharge, as well as the influence of impurities in the films being eroded. Oxygen concentrations below 20 % in helium are sufficient to achieve high erosion rates. The discharge can lead to the formation of oxide layers on surfaces which-as demonstrated for tungsten-can be rapidly reduced by post-treatment in a hydrogen discharge. For carbon, aluminum and iron the physical sputtering yield may become similar to the erosion yield of redeposited layers, but it is by more than one order of magnitude smaller for tungsten. Using a-B:C:H films with varying boron content, it was found that impurities can cause the erosion rate to drop by orders of magnitude.

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“…Among others [6], oxygen containing glow discharges have been proposed as detritiation technique [7,8] with the advantages of operation at room temperature and of requiring only simple-and mostly already installed-hardware. First tests of the technique in a tokamak environment have been carried out at TFTR [9], TEXTOR [10], and HT-7 [11].…”

Section: Introductionmentioning

confidence: 99%

Oxygen glow discharge cleaning in ASDEX Upgrade

Hopf

Rohde

Jacob

et al. 2007

Journal of Nuclear Materials

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Oxygen-containing glow discharges have been proposed for the removal of redeposited hydrocarbon films in fusion experiments. After the 2004/2005 experimental campaign a 49 hours He/O 2 glow discharge experiment was performed in ASDEX Upgrade. A total of 25 g of carbon was removed as CO and CO 2 . In order to probe erosion in various locations in the vessel 48 well defined laboratory-deposited a-C:H film samples were distributed poloidally and toroidally. In principle, the films were successfully removed on all types of surfaces directly facing the discharge. Additionally, two sets of tungsten-coated tiles were removed from the vessel, one prior to and one after the cleaning discharge. Analysis of films on the tiles revealed that they consisted mainly of boron with only little carbon and deuterium. Neither erosion of these films nor any further reduction of their carbon and deuterium content due to the glow discharge could be detected.

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Tritium removal from TFTR

Cited by 31 publications

References 11 publications

Hydrogenic retention with high-Z plasma facing surfaces in Alcator C-Mod

Hydrogenic retention with high-Z plasma facing surfaces in Alcator C-Mod

Oxygen glow discharge cleaning in nuclear fusion devices

Oxygen glow discharge cleaning in ASDEX Upgrade

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