| 1. |
- Strachan, J. D., et al.
(författare)
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Modelling of carbon migration during JET C-13 injection experiments
- 2008
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 48:10
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Tidskriftsartikel (refereegranskat)abstract
- JET has performed two dedicated carbon migration experiments on the final run day of separate campaigns ( 2001 and 2004) using (CH4)-C-13 methane injected into repeated discharges. The EDGE2D/NIMBUS code modelled the carbon migration in both experiments. This paper describes this modelling and identifies a number of important migration pathways: ( 1) deposition and erosion near the injection location, ( 2) migration through the main chamber SOL, (3) migration through the private flux region (PFR) aided by E x B drifts and ( 4) neutral migration originating near the strike points. In H-Mode, type I ELMs are calculated to influence the migration by enhancing erosion during the ELM peak and increasing the long-range migration immediately following the ELM. The erosion/re-deposition cycle along the outer target leads to a multistep migration of C-13 towards the separatrix which is called 'walking'. This walking created carbon neutrals at the outer strike point and led to 13C deposition in the PFR. Although several migration pathways have been identified, quantitative analyses are hindered by experimental uncertainty in divertor leakage, and the lack of measurements at locations such as gaps and shadowed regions.
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| 2. |
- Corre, Y., et al.
(författare)
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Hybrid H-mode scenario with nitrogen seeding and type III ELMs in JET
- 2008
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Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 0741-3335 .- 1361-6587. ; 50:11, s. 115012-
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Tidskriftsartikel (refereegranskat)abstract
- The performance of the 'hybrid' H-mode regime (long pulse operation with high neutron fluency) has been extensively investigated in JET during the 2005-2007 experimental campaign up to normalized pressure beta(N) = 3, toroidal magnetic field B-t = 1.7T, with type I ELMs plasma edge conditions. The optimized external current drive sources, self-generated non-inductive bootstrap current and plasma core stability properties provide a good prospect of achieving a high fusion gain at reduced plasma current for long durations in ITER. One of the remaining issues is the erosion of the divertor target plates associated with the type I ELM regime. A possible solution could be to operate with a plasma edge in the type III ELM regime (reduced transient and stationary heat loads) obtained with impurity seeding. An integrated hybrid type III ELM regime with a normalized pressure beta(N) = 2.6 (PNBI similar to 20-22 MW) and a thermal confinement factor of H-98* 98(y, 2) similar to 0.83 has been recently successfully developed on JET with nitrogen seeding. This scenario shows good plasma edge condition (compatible with the future ITER-like wall on JET) and moderate MHD activity. In this paper, we report on the experimental development of the scenario (with plasma current I-p = 1.7MA and magnetic field B-t = 1.7T) and the trade-off between heat load reduction at the target plates and global confinement due to nitrogen seeding and type III ELM working conditions.
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| 3. |
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| 4. |
- Airila, M. I., et al.
(författare)
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ERO modelling of local deposition of injected C-13 tracer at the outer divertor of JET
- 2009
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Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T138, s. 014021-
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Tidskriftsartikel (refereegranskat)abstract
- The 2004 tracer experiment of JET with the injection of (CH4)-C-13 into H-mode plasma at the outer divertor has been modelled with the Monte Carlo impurity transport code ERO. EDGE2D solutions for inter-ELM and ELM-peak phases were used as plasma backgrounds. Local two-dimensional (2D) deposition patterns at the vertical outer divertor target plate were obtained for comparison with post-mortem surface analyses. ERO also provides emission profiles for comparison with radially resolved spectroscopic measurements. Modelling indicates that enhanced re-erosion of deposited carbon layers is essential in explaining the amount of local deposition. Assuming negligible effective sticking of hydrocarbons, the measured local deposition of 20-34% is reproduced if re-erosion of deposits is enhanced by a factor of 2.5-7 compared to graphite erosion. If deposits are treated like the substrate, the modelled deposition is 55%. Deposition measurements at the shadowed area around injectors can be well explained by assuming negligible re-erosion but similar sticking behaviour there as on plasma-wetted surfaces.
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| 5. |
- Coad, J. P., et al.
(författare)
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Erosion and deposition in the JET MkII-SRP divertor
- 2007
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 287-293
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Tidskriftsartikel (refereegranskat)abstract
- Carbon-13 labelled methane was injected into the outer divertor during a series of H-mode discharges on the last day of operations with the JET MkII-SRP divertor. Tiles from around the vessel were removed during the subsequent shutdown and surface deposits were analysed by IBA techniques and SIMS. First attempts to model the pattern of 13 C deposition using EDGE2D are reported. Erosion of W markers at the outer divertor was observed, with implications for the ITER-like wall experiment planned for JET, whilst thin film growth in the same region has been followed by the effect on infrared measurements. The composition of thick films deposited at the inner divertor during the MkII-SRP campaign, and the migration to the inner corner of the divertor observed by a quartz micro-balance, provide further information on divertor transport. Crown
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| 6. |
- Loarer, T., et al.
(författare)
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Gas balance and fuel retention in fusion devices
- 2007
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 47:9, s. 1112-1120
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Tidskriftsartikel (refereegranskat)abstract
- The evaluation of hydrogenic retention in present tokamaks is of crucial importance to estimate the expected tritium (T) vessel inventory in ITER, limited from safety considerations to 350 g. In the framework of the European Task Force on Plasma Wall Interaction (EU TF on PWI) efforts are underway to investigate gas balance and fuel retention during discharges, and to compare the data obtained with those from post-mortem analysis of in-vessel components exposed over whole experimental campaigns. This paper summarizes the principal findings from coordinated studies on gas balance and fuel retention from a number of European tokamaks, namely, ASDEX-Upgrade (AUG), JET, TEXTOR and Tore Supra (TS). For most devices, the long-term retention fraction deduced from integrated particle balance is similar to 10-20%. This is larger than the similar to 3-4% deduced from post-mortem analysis of plasma facing components (PFCs). However, from the database available for tokamaks with their main PFCs made of carbon, the important conclusion is that the T inventory limit (set by the working guideline for operations) could be reached in ITER within fewer than 100 discharges. This, therefore, would seriously impact on operation of the device unless efficient T removal processes are developed.
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| 7. |
- Rubel, Marek J., et al.
(författare)
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An overview of a comprehensive First Mirror Test for ITER at JET
- 2009
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Ingår i: Journal of Nuclear Materials. - Amsterdam : ELSEVIER SCIENCE BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 1066-1069
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Tidskriftsartikel (refereegranskat)abstract
- The test was performed with 32 stainless steel and molybdenum mirrors placed in pan-pipe shaped cassettes and exposed in JET in the divertor and on the main chamber wall for 127000 s including 97000 s of X-point operation. Surface composition and total reflectivity were determined afterwards All mirrors. from the divertor were coated with deuterated carbon deposits causing the reflectivity loss by a factor of 6-10 in the visible range. Flaking and exfoliation of deposits were observed in some cases On the main. chamber wall the deposition occurred mainly on mirrors located deep in cassette channels whereas mirrors close to the channels entrances were free from deposits and retained fair reflectivity (similar to 90% of initial value) especially in the infra-red range. No significant differences in behaviour of steel and molybdenum were noted. The need for development of methods for mirror cleaning and/or protection in a reactor-class device is addressed.
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| 8. |
- Widdowson, A., et al.
(författare)
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An overview of erosion-deposition studies for the JET Mk II high delta divertor
- 2009
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Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847. ; T138, s. 014005-
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Tidskriftsartikel (refereegranskat)abstract
- Post-mortem analyses of tiles removed from the JET MkII HD divertor in 2007 are presented. The results indicate an increase in deposition at the outer plasma-shadowed region of the divertor, not seen prior to 2004 and indicate a shift away from the asymmetric picture of net deposition at the inner divertor compared with no overall deposition or erosion at the outer divertor. Surface analysis of the inner and outer vertical divertor tiles is largely the same as observed previously; however, a notable increase in Be composition on the inner and outer floor tiles is observed. An attempt has been made to correlate these data with campaign-averaged plasma configurations and spectroscopy results. While some changes in deposition/erosion characteristics can be explained, further detailed analysis of diagnostics and especially of time-resolved data, such as from rotating collector and quartz microbalance diagnostics, is required.
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| 9. |
- Widdowson, A., et al.
(författare)
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Efficacy of photon cleaning of JET divertor tiles
- 2007
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 341-345
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Tidskriftsartikel (refereegranskat)abstract
- Photon cleaning by means of a flash-lamp was used for in-situ detritiation of the inner wall tiles of the JET divertor in May 2004. Additional trials were also performed ex-situ in October 2004 on divertor base tiles. Early work confirmed that for pulse energies between 150 J and 300 J some deposited material was removed. To increase the amount of material removed during photon cleaning, further experiments with higher pulse energies (500 J) were performed and are reported here. Analysis of cross sections confirmed a removal rate of 0.04 mu m/pulse, removing similar to 80 mu m from 200 mu m thick deposits over a treatment area of 15 x 10(-4) m(2). During the photon cleaning tests at least 12% of the tritium inventory for the tile was removed. It was also shown that deuterium was desorbed from a depth similar to 7 mu m beyond the depth of material removed. Crown
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| 10. |
- Widdowson, A., et al.
(författare)
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Testing of beryllium marker coatings in PISCES-B for the JET ITER-like wall
- 2009
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 988-991
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Tidskriftsartikel (refereegranskat)abstract
- Beryllium has been chosen as the first wall material for ITER. In order to understand the issues of material migration and tritium retention associated with the use of beryllium, a largely beryllium first wall will be installed in JET. As part of the JET ITER-like wall, beryllium tiles with marker coatings are proposed as a diagnostic tool for studying the erosion and deposition of beryllium around the vessel. The nominal structure for these coatings is a similar to 10 mu m beryllium surface layer separated from the beryllium tile by a 2-3 mu m metallic inter-layer. Two types of coatings are tested here; one with a nickel inter-layer anti one with a copper/beryllium mixed inter-layer. The coating samples were deposited by DC magnetron Sputtering at General Atomics and were exposed to deuterium plasma in PISCES-B. The results of this testing show that the beryllium/nickel marker coating would be suitable for installation in JET.
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