The amount and distribution of deuterium retained in the jet divertor after the C and Be phases in 1994–1995

Coad, J.P.; Rubel, M.; Wu, C.H.

doi:10.1016/s0022-3115(97)80073-6

Cited by 24 publications

(48 citation statements)

References 4 publications

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“…The corresponding decay length, k, is approximately 1.5 mm. These concentrations are 70-100 times lower than those on plasma-facing surfaces [12] and in the gaps separating tiles.…”

Section: Mk-i Divertor With Beryllium Tilesmentioning

confidence: 74%

“…As a result of roof-shaping and arrangement of the tiles with respect to the magnetic field lines, the lower part of a given tile was shadowed by the upper edge of the adjacent neighbour. Views of the JET vessel with the Mk-I divertor and the arrangement of tiles with respect to the magnetic field lines can be found elsewhere [9,12]. The divertor was first operated for $60 000 s with CFC tiles and then for $20 000 s with castellated (6 · 6 mm with 6 mm deep and 0.6 mm wide groove) beryllium blocks.…”

Section: Methodsmentioning

confidence: 99%

“…on surfaces located in gaps between the tiles. The appearance of the plasma exposed CFC tiles with flaking co-deposits can be found in [9], whereas details regarding the elemental composition of plasma-facing surfaces of the divertor floor have been reported in [12]. Only a brief summary of the most important facts is given here.…”

Section: Mk-i Divertor With Cfc Tilesmentioning

confidence: 99%

“…From the analysis results compiled in [12], the integrated fuel content on the plasma-facing side of the inner corner Be tile could be estimated at the level of 9.32 · 10 20 D atoms (C surface ). The recently determined integrated content in the castellation grooves of this tile is $0.22 · 10 19 D atoms (C groove ), whereas 3.05 · 10 20 D atoms (C gap ) are found on the surface located in the gap.…”

Section: Mk-i Divertor With Beryllium Tilesmentioning

confidence: 99%

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Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Rubel

Coad

Pitts

2007

Journal of Nuclear Materials

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The main focus of this work is fuel retention in plasma components of the JET water-cooled Mk-I divertors operated with small tiles, first with carbon fibre composite (CFC) and then with castellated beryllium. Until recently these have been the only large-scale structures of this type used in fusion experiments. Three issues regarding fuel retention and material migration are addressed: (i) accumulation in gaps separating tiles and in the grooves of castellation; (ii) comparison of deposition on carbon and beryllium; (iii) in-depth migration of deuterium into the bulk of CFC. The essential results are summarised as follows: (i) co-deposition occurs up to a few cm deep in the gaps between the Mk-I tiles; (ii) fuel inventory in the CFC tile gaps exceeds that on plasma-facing surfaces by up to a factor of 2; (iii) in gaps between the beryllium tiles from the inner divertor corner the fuel content reaches 30% of that on plasma-facing surfaces, whereas in the grooves of castellation in Be the fuel content is less than 3.0% of that found on the top surface; (iv) fuel inventory on the Be tiles is strongly associated with the carbon co-deposition; (v) the D content measured in the bulk (1.5 mm below the surface) on cleaved CFC tiles exceeds 1 · 10 15 cm À2 . Implications of these results for a next-step device are addressed and the transport mechanism into the gaps is briefly discussed. The results presented here suggest that in a machine with non-carbon walls in the main chamber (as foreseen for ITER) the material transport and subsequent fuel inventory in the castellation would be reduced.

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“…The corresponding decay length, k, is approximately 1.5 mm. These concentrations are 70-100 times lower than those on plasma-facing surfaces [12] and in the gaps separating tiles.…”

Section: Mk-i Divertor With Beryllium Tilesmentioning

confidence: 74%

Section: Methodsmentioning

confidence: 99%

Section: Mk-i Divertor With Cfc Tilesmentioning

confidence: 99%

Section: Mk-i Divertor With Beryllium Tilesmentioning

confidence: 99%

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Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Rubel

Coad

Pitts

2007

Journal of Nuclear Materials

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“…The term "fuel inventory" denotes accumulation and long-term retention of fuel in all in-vessel components, either those facing the plasma directly or located in remote (shadowed) areas such as water cooled louvers in the divertor structure [1,16,17] or other parts of pumping ducts. This applies to all hydrogen isotopes and has several detrimental effects on the reactor operation.…”

Section: Impact Of Erosion and Re-deposition On Fuel Inventory And mentioning

confidence: 99%

Structure Materials in Fusion Reactors: Issues Related to Tritium, Radioactivity and Radiation-Induced Effects

Rubel

2012

Fusion Science and Technology

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A concise overview is given on materials applied in fusion technology. The influence of plasma operation on the behaviour of reactor components and diagnostic systems is discussed with emphasis on effects caused by fast particles reaching the reactor wall. Issues related to primary and induced radioactivity are reviewed: tritium inventory and transmutation. Tritium breeding in the reactor blanket, separation of hydrogen isotopes and safety aspects in handling radioactively contaminated components are also included.

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Effective laser-induced removal of co-deposited layers from plasma-facing components in a tokamak

et al. 2006

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An experimental set-up and spectroscopy diagnostic method for laser-induced fuel removal and decomposition of co-deposited layers on plasma-facing components from tokamaks are described. For irradiation of a graphite limiter tile from the TEXTOR tokamak Nd:YAG 3.5-ns pulse laser with a repetition rate of 10 Hz and single pulse energy of up to 0,8 J at 1,06 µm has been used. The spectroscopy system allowed recording of spectra in the visible wavelength range including CII and Dα spectral lines . The evolution of CII and Dα spectral lines was observed pulse-by-pulse during the co-deposit removal. The efficient ablation of the 45 µm thick co-deposit occured after approximately 50 laser pulses.

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The amount and distribution of deuterium retained in the jet divertor after the C and Be phases in 1994–1995

Cited by 24 publications

References 4 publications

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Overview of co-deposition and fuel inventory in castellated divertor structures at JET

Structure Materials in Fusion Reactors: Issues Related to Tritium, Radioactivity and Radiation-Induced Effects

Effective laser-induced removal of co-deposited layers from plasma-facing components in a tokamak

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