| 1. |
- Abel, I, et al.
(författare)
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Overview of the JET results with the ITER-like wall
- 2013
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 53:10, s. 104002-
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Tidskriftsartikel (refereegranskat)abstract
- Following the completion in May 2011 of the shutdown for the installation of the beryllium wall and the tungsten divertor, the first set of JET campaigns have addressed the investigation of the retention properties and the development of operational scenarios with the new plasma-facing materials. The large reduction in the carbon content (more than a factor ten) led to a much lower Z(eff) (1.2-1.4) during L- and H-mode plasmas, and radiation during the burn-through phase of the plasma initiation with the consequence that breakdown failures are almost absent. Gas balance experiments have shown that the fuel retention rate with the new wall is substantially reduced with respect to the C wall. The re-establishment of the baseline H-mode and hybrid scenarios compatible with the new wall has required an optimization of the control of metallic impurity sources and heat loads. Stable type-I ELMy H-mode regimes with H-98,H-y2 close to 1 and beta(N) similar to 1.6 have been achieved using gas injection. ELM frequency is a key factor for the control of the metallic impurity accumulation. Pedestal temperatures tend to be lower with the new wall, leading to reduced confinement, but nitrogen seeding restores high pedestal temperatures and confinement. Compared with the carbon wall, major disruptions with the new wall show a lower radiated power and a slower current quench. The higher heat loads on Be wall plasma-facing components due to lower radiation made the routine use of massive gas injection for disruption mitigation essential.
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| 2. |
- Romanelli, F, et al.
(författare)
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Overview of the JET results
- 2011
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Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 51:9
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Tidskriftsartikel (refereegranskat)abstract
- Since the last IAEA Conference JET has been in operation for one year with a programmatic focus on the qualification of ITER operating scenarios, the consolidation of ITER design choices and preparation for plasma operation with the ITER-like wall presently being installed in JET. Good progress has been achieved, including stationary ELMy H-mode operation at 4.5 MA. The high confinement hybrid scenario has been extended to high triangularity, lower ρ*and to pulse lengths comparable to the resistive time. The steady-state scenario has also been extended to lower ρ*and ν*and optimized to simultaneously achieve, under stationary conditions, ITER-like values of all other relevant normalized parameters. A dedicated helium campaign has allowed key aspects of plasma control and H-mode operation for the ITER non-activated phase to be evaluated. Effective sawtooth control by fast ions has been demonstrated with3He minority ICRH, a scenario with negligible minority current drive. Edge localized mode (ELM) control studies using external n = 1 and n = 2 perturbation fields have found a resonance effect in ELM frequency for specific q95values. Complete ELM suppression has, however, not been observed, even with an edge Chirikov parameter larger than 1. Pellet ELM pacing has been demonstrated and the minimum pellet size needed to trigger an ELM has been estimated. For both natural and mitigated ELMs a broadening of the divertor ELM-wetted area with increasing ELM size has been found. In disruption studies with massive gas injection up to 50% of the thermal energy could be radiated before, and 20% during, the thermal quench. Halo currents could be reduced by 60% and, using argon/deuterium and neon/deuterium gas mixtures, runaway electron generation could be avoided. Most objectives of the ITER-like ICRH antenna have been demonstrated; matching with closely packed straps, ELM resilience, scattering matrix arc detection and operation at high power density (6.2 MW m-2) and antenna strap voltages (42 kV). Coupling measurements are in very good agreement with TOPICA modelling. © 2011 IAEA, Vienna.
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| 3. |
- Counsell, G., et al.
(författare)
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Tritium retention in next step devices and the requirements for mitigation and removal techniques
- 2006
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Ingår i: Plasma Physics and Controlled Fusion. - 0741-3335 .- 1361-6587. ; 48:12B, s. B189-B199
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Tidskriftsartikel (refereegranskat)abstract
- Mechanisms underlying the retention of fuel species in tokamaks with carbon plasma-facing components are presented, together with estimates for the corresponding retention of tritium in ITER. The consequential requirement for new and improved schemes to reduce the tritium inventory is highlighted and the results of ongoing studies into a range of techniques are presented, together with estimates of the tritium removal rate in ITER in each case. Finally, an approach involving the integration of many tritium removal techniques into the ITER operational schedule is proposed as a means to extend the period of operations before major intervention is required.
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| 4. |
- Widdowson, A., et al.
(författare)
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Removal of beryllium-containing films deposited in JET from mirror surfaces by laser cleaning
- 2011
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Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 415:1, s. S1199-S1202
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Tidskriftsartikel (refereegranskat)abstract
- A set of stainless steel (SS) and molybdenum mirror samples located in the divertor and at the outer mid-plane of the vessel were exposed in JET from 2005 to 2007. A selection of these mirror samples with well adhered deposits (i.e. not flaking) of up to a few hundred nanometers in thickness and with Be/C ratios ranging from 0 to similar to 1 have been cleaned using a laser system developed at CEA, Saclay. Following laser cleaning the recovered reflectivity was generally better in the infrared than the visible spectrum, with recovery of up to 90% of the initial reflectivity being obtained at 1600 nm for both Mo and SS mirrors falling as low as 20-30% of initial reflectivity at a wavelength of 400 nm for some SS mirrors, rising to similar to 80% for Mo mirrors. Some deposit remained on the mirrors after the cleaning trials.
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| 5. |
- Leontyev, A., et al.
(författare)
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Theoretical and experimental studies on molybdenum and stainless steel mirrors cleaning by high repetition rate laser beam
- 2011
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 86:9-11, s. 1728-1731
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Tidskriftsartikel (refereegranskat)abstract
- Our studies were aimed to determine the damage threshold of molybdenum (Mo) and stainless steel (SS) mirrors to provide the maximum fluence which the mirror surfaces could withstand without affecting their reflectivity properties. A high repetition rate ytterbium fiber laser (20 kHz, 1.06 mu m, 120 ns) was applied. The experimental single-pulse and multiple-pulse damage thresholds were obtained. To calculate damage thresholds, a 1D analytical model which takes into account the temperature dependent absorptance and multiple-pulse damage based on plastic deformations accumulation was applied. The experimental damage thresholds and the theoretical ones are in a good agreement. Cleaning tests with the contaminated mirrors exposed in JET have been performed.
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| 6. |
- Rubel, Marek, et al.
(författare)
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Comprehensive First Mirror Test for ITER at JET with Carbon Walls
- 2010
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Ingår i: Proceedings of the23rd IAEA Fusion Energy Conference.
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Konferensbidrag (refereegranskat)abstract
- Metallic mirrors will be essential components of all optical spectroscopy and imaging systems forplasma diagnosis that will be used on the next-step magnetic fusion experiment, ITER. Any change of the mirrorperformance, in particular reflectivity, will influence the quality and reliability of detected signals. On therequest of the ITER Design Team, a First Mirror Test (FMT) has been carried out at JET during campaigns in2005-2007 and 2008-2009. To date, it has been the most comprehensive test performed with a large number oftest mirrors exposed in an environment containing both carbon and beryllium; the total plasma time (in 2005-2007 period) over 35 h including 27 h of X-point operation. 32 stainless steel and polycrystalline molybdenumflat-front and 45oangled mirrors were installed in separate channels of cassettes on the outer wall and in the MkII HD divertor: inner leg, outer leg and base plate under the load bearing tile. Post exposure studies comprisedreflectivity measurements and surface analyses with microscopy, secondary ion mass spectrometry, ion beamanalysis and energy dispersive X-ray spectroscopy.. The essential results are: (i) on the outer wall highreflectivity (~90%) is maintained for mirrors close to the channel entrance but it is degraded by 30-40 % deeperin the channel (ii) reflectivity loss by 70-90% is measured for mirrors placed in the divertor: outer, inner andbase; (iii) deuterium and carbon are the main elements detected on all mirror surfaces and the presence ofberyllium is also found; (iv) thick deposits show rough columnar structure and thickness is 1-20 μm; (v) bubblelike structures are detected in deposits; (vi) the deposition in channels in the divertor cassettes is pronounced atthe very entrance; (vii) photonic cleaning with laser removes deposits but the surface is damaged by laser pulses.In summary, reflectivity of all tested mirrors is degraded either by erosion with CX neutrals or by the formationof thick deposits. The implications of results obtained for first mirrors in next-step device are discussed andcritical assessment of various methods for in-situ cleaning of mirrors is presented. The conclusion is thatengineering solutions should be developed in order to install shutters or to implement a cassette with mirrors toreplace periodically the degraded ones
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| 7. |
- Grisolia, C., et al.
(författare)
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Treatment of ITER plasma facing components : Current status and remaining open issues before ITER implementation
- 2007
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 2390-2398
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Tidskriftsartikel (refereegranskat)abstract
- The in-vessel tritium inventory control is one of the most ITER challenging issues which has to be resolved to fulfil safety requirements. This is due mainly to the presence of carbon as a constituent of plasma facing components (PFCs) which leads to a high fuel permanent retention. For several years now, physics studies and technological developments have been undertaken worldwide in order to develop reliable techniques which could be used in ITER severe environment (magnetic field, vacuum, high temperature) for in situ tritium recovery. The scope of this contribution is to review the present status of these achievements and define the remaining work to be done in order to propose a dedicated work program. Different treatment techniques (chemical treatments, photonic cleaning) will be reviewed. In the frame of ITER, they will be compared in terms of fuel removal rate as well as surface accessibility, type of production (gas or particulates), ability to clean mixed material. And lastly, consequences of bulk trapping observed in tokamak on the techniques currently under development will be addressed.
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| 8. |
- Grisolia, C., et al.
(författare)
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JET contributions to ITER technology issues
- 2006
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Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 81:07-jan, s. 149-154
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Tidskriftsartikel (refereegranskat)abstract
- The Joint European Torus (JET) fusion machine is the only device capable of operation with tritium and of handling Be and therefore is best suited to the study of tritium and fusion-related issues. A large variety of activities are performed within the JET fusion technology task force (FT-TF). In this paper, some topics such as erosion/deposition and material transport, characterisation of flakes and detritiation techniques are highlighted. Recent examples of results obtained on waste management studies are also given. Data on some ITER-relevant components that have been tested at JET, such as a pumping cryopanel and hardened optics fibers, are presented. In all fields, the work to be addressed in future JET work programmes is discussed.
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