| 1. |
- Krasilnikov, A., et al.
(author)
-
Evidence of 9 Be + p nuclear reactions during 2ω CH and hydrogen minority ICRH in JET-ILW hydrogen and deuterium plasmas
- 2018
-
In: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:2
-
Journal article (peer-reviewed)abstract
- The intensity of 9Be + p nuclear fusion reactions was experimentally studied during second harmonic (2ω CH) ion-cyclotron resonance heating (ICRH) and further analyzed during fundamental hydrogen minority ICRH of JET-ILW hydrogen and deuterium plasmas. In relatively low-density plasmas with a high ICRH power, a population of fast H+ ions was created and measured by neutral particle analyzers. Primary and secondary nuclear reaction products, due to 9Be + p interaction, were observed with fast ion loss detectors, γ-ray spectrometers and neutron flux monitors and spectrometers. The possibility of using 9Be(p, d)2α and 9Be(p, α)6Li nuclear reactions to create a population of fast alpha particles and study their behaviour in non-active stage of ITER operation is discussed in the paper.
|
|
| 2. |
|
|
| 3. |
|
|
| 4. |
|
|
| 5. |
|
|
| 6. |
|
|
| 7. |
|
|
| 8. |
|
|
| 9. |
|
|
| 10. |
|
|
| 11. |
|
|
| 12. |
|
|
| 13. |
|
|
| 14. |
|
|
| 15. |
|
|
| 16. |
|
|
| 17. |
|
|
| 18. |
|
|
| 19. |
|
|
| 20. |
|
|
| 21. |
|
|
| 22. |
|
|
| 23. |
-
Overview of the JET results
- 2015
-
In: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:10
-
Journal article (peer-reviewed)
|
|
| 24. |
- Abel, I, et al.
(author)
-
Overview of the JET results with the ITER-like wall
- 2013
-
In: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 53:10, s. 104002-
-
Journal article (peer-reviewed)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.
|
|
| 25. |
- Romanelli, F, et al.
(author)
-
Overview of the JET results
- 2011
-
In: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 51:9
-
Journal article (peer-reviewed)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.
|
|
| 26. |
- Bécoulet, A., et al.
(author)
-
Science and technology research and development in support to ITER and the Broader Approach at CEA
- 2013
-
In: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 53:10
-
Journal article (peer-reviewed)abstract
- In parallel to the direct contribution to the procurement phase of ITER and Broader Approach, CEA has initiated research & development programmes, accompanied by experiments together with a significant modelling effort, aimed at ensuring robust operation, plasma performance, as well as mitigating the risks of the procurement phase. This overview reports the latest progress in both fusion science and technology including many areas, namely the mitigation of superconducting magnet quenches, disruption-generated runaway electrons, edge-localized modes (ELMs), the development of imaging surveillance, and heating and current drive systems for steady-state operation. The WEST (W Environment for Steady-state Tokamaks) project, turning Tore Supra into an actively cooled W-divertor platform open to the ITER partners and industries, is presented.
|
|
| 27. |
- Grisolia, C., et al.
(author)
-
JET contributions to ITER technology issues
- 2006
-
In: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 81:07-jan, s. 149-154
-
Journal article (peer-reviewed)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.
|
|
| 28. |
- Grisolia, C., et al.
(author)
-
Treatment of ITER plasma facing components : Current status and remaining open issues before ITER implementation
- 2007
-
In: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 2390-2398
-
Journal article (peer-reviewed)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.
|
|
| 29. |
- Rosanvallon, S., et al.
(author)
-
Tritium related studies within the JET Fusion Technology work programme
- 2005
-
In: Fusion science and technology. - 1536-1055 .- 1943-7641. ; 48:1, s. 268-273
-
Journal article (peer-reviewed)abstract
- The JET Fusion Technology (FT) work programme was launched in 2000, in the frame of the European Fusion Development Agreement, to address issues related to JET and ITER. In particular, there are four topics related to tritium being investigated Based on the experience gained on the existing tokamaks, first calculations indicate that in-vessel tritium retention could represent a burden for ITER operation. Therefore erosion/deposition studies are being performed in order to better understand the layer co-deposition and tritium retention processes in tokamaks. Moreover, testing of in-situ detritiation processes, in particular laser and flash lamp treatments, should assess detritiation techniques for in-vessel components in the ITER-relevant JET configuration. To reduce the constraints on waste disposal, dedicated procedures are being developed for detritiation Of metals, graphite, carbon-fibre composites, process and housekeeping waste. During the operational and decommissioning phases of a fusion reactor, many processes will produce tritiated water. Key components for an ITER relevant water detritiation facility are being studied experimentally with the aim of producing a complete design that could be implemented and tested at JET. This paper describes these topics of the FT-programme, the strategy developed and the results obtained so far.
|
|
| 30. |
- Leontyev, A., et al.
(author)
-
Theoretical and experimental studies on molybdenum and stainless steel mirrors cleaning by high repetition rate laser beam
- 2011
-
In: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 86:9-11, s. 1728-1731
-
Journal article (peer-reviewed)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.
|
|
