Tritium distribution on plasma-facing tiles from ASDEX Upgrade

Sugiyama, K.; Tanabe, T.; Krieger, K.; Neu, R.; Bekris, N.

doi:10.1016/j.jnucmat.2004.10.158

Cited by 5 publications

(4 citation statements)

References 14 publications

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“…As mentioned, the fuel tritium is predominantly co-deposited with eroded first-wall material, but in D-D discharge machines the tritium distribution on the 65002-p4 plasma-facing components has been found to be similar to the distribution of high-energy triton implantation [29,30]. To confirm these findings, the D-D-triton flux onto the material surfaces was simulated.…”

Section: D-d Tritium Surface Distribution In Asdexmentioning

confidence: 65%

“…For comparison, the measured deuterium concentrations at the depth of 20 µm were found to be several 10 −4 D/C [28]. Qualitatively, the distribution of tritium surface deposition has been studied by Photo-Stimulate Luminescence (PSL) measurements carried out at JT-60U [29] and at ASDEX Upgrade [30]. Recent measurement results on ASDEX Upgrade are shown in figs.…”

Section: D-d Tritium Surface Distribution In Asdexmentioning

confidence: 99%

“…Since the reaction rates for the two branches are practically equal, the triton source rate can be experimentally inferred from the measured neutron rate. The tritium surface distribution in ASDEX Upgrade has been studied qualitatively by Photo-Stimulate Luminescence (PSL) measurements [2]. Previously, these and more recent measurements have been compared to simulation results of the Improved H-mode discharge # 17219, t = 2.5 s [3,4].…”

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confidence: 99%

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Fusion tritons and plasma-facing components in a fusion reactor

Kurki-Suonio¹,

Hynönen²,

Ahlgren

et al. 2007

Europhys. Lett.

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In future fusion reactors the long-term tritium retention will be a critical issue. In ITER, which will be the first device to demonstrate a burning plasma, the in-vessel tritium inventory limit will be 350 g. The primary retention mechanism of tritium is co-deposition with eroded first wall material. The measured tritium build-up rates at JET and TFTR tokamaks, both of which were operating with carbon walls at the time of their tritium campaigns, are too high for ITER. For these reasons there is today a growing interest in carbon-free fusion machines:ASDEX Upgrade in Garching, Germany, has replaced, step-by-step, all carbon fibre composite structures by tungsten-coated ones.The tritium inventory from the fuel tritium is typically found on top of the material surfaces, where it is deposited as amorphous hydrogen-rich carbon layers. This is because the fuel tritons are thermal, i.e., have low energy (E 100 eV) and, thus, a penetration depth of at most a few nanometers. However, tritium is also formed in deuterium-deuterium (D-D) reactions. The quantity of tritium formed inside the plasma will be insignificant compared to the amount of fuel tritium: the D-D fusion reaction rate is about 1/200 of the deuterium-tritium reaction rate. But due to its very high energy, contribution of the fusion-born tritium to the long-term tritium inventory and material damages can still be significant. This difference becomes evident by comparing the mean range and backscattered fraction of tritons in carbon, tungsten and beryllium, shown in tables 1 and 2. Energy 100 eV 1 MeV tungsten (2.8 ± 1.5) nm (4.3 ± 0.1) µm carbon (2.8 ± 1.6) nm (7.7 ± 0.1) µm beryllium (3.3 ± 1.6) nm (10.2 ± 0.1) µm 2) % (0.11 ± 0.01) % carbon (9.9 ± 0.1) % (0.000 ± 0.001) % beryllium (6.2 ± 0.1) % (0.000 ± 0.001) %

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Section: D-d Tritium Surface Distribution In Asdexmentioning

confidence: 65%

Section: D-d Tritium Surface Distribution In Asdexmentioning

confidence: 99%

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confidence: 99%

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Fusion tritons and plasma-facing components in a fusion reactor

Kurki-Suonio¹,

Hynönen²,

Ahlgren

et al. 2007

Europhys. Lett.

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“…The long-term tritium retention will be a critical issue in future fusion reactors. Tritium can be co-deposited with the eroded first wall material, but in D-D discharge machines the tritium distribution on the plasma-facing components has been found to be similar to the distribution of high-energy triton implantation [18,19]. To confirm these findings, the surface distribution of tritons born from beam-plasma and plasma-plasma D-D reactions was simulated.…”

Section: Tritium Surface Distributionmentioning

confidence: 93%

Erratum on surface loads and edge fast ion distribution for co- and counter-injection in ASDEX Upgrade

Hynönen¹,

Kurki-Suonio²

2007

Plasma Phys. Control. Fusion

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The fast particle flux onto the material surfaces and the fast ion edge distribution are compared between ASDEX Upgrade H-mode and quiescent H-mode (QH-mode) in the presence of toroidal ripple and radial electric field E r by using the orbit-following Monte Carlo code ASCOT. So far, the QH-mode has been obtained only with counter-injection of the neutral beams. The wall load caused by co-injected beams in H-mode is small and without ripple it is negligible. With counter-injection (QH-mode case), the wall load is substantial even without the ripple. The ripple always increases the wall load, but the divertor load is either decreased or is unchanged. The effect of E r alone is small, but it boosts the ripple-trapping of beam particles near the location of its maximum absolute value on the horizontal midplane. The fast ion density and its gradient in the pedestal region are higher for counterinjected than for co-injected particles. The ripple decreases the density gradient in both the cases. To make a connection with the experiment, the flux of high-energy tritons from beam-plasma interactions onto the surfaces is evaluated. The simulated flux of tritons impinging on the material surfaces is in qualitative agreement with the measured tritium distribution on the wall and divertor.

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