| 1. |
- Bombarda, F., et al.
(författare)
-
Runaway electron beam control
- 2019
-
Ingår i: Plasma Physics and Controlled Fusion. - : IOP Publishing. - 1361-6587 .- 0741-3335. ; 61:1
-
Tidskriftsartikel (refereegranskat)
|
|
| 2. |
- Krasilnikov, A., et al.
(författare)
-
Evidence of 9 Be + p nuclear reactions during 2ω CH and hydrogen minority ICRH in JET-ILW hydrogen and deuterium plasmas
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:2
-
Tidskriftsartikel (refereegranskat)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.
|
|
| 3. |
-
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:1
-
Forskningsöversikt (refereegranskat)
|
|
| 4. |
|
|
| 5. |
|
|
| 6. |
|
|
| 7. |
|
|
| 8. |
|
|
| 9. |
|
|
| 10. |
|
|
| 11. |
|
|
| 12. |
|
|
| 13. |
|
|
| 14. |
|
|
| 15. |
- Murari, A., et al.
(författare)
-
A control oriented strategy of disruption prediction to avoid the configuration collapse of tokamak reactors
- 2024
-
Ingår i: Nature Communications. - 2041-1723 .- 2041-1723. ; 15:1
-
Tidskriftsartikel (refereegranskat)abstract
- The objective of thermonuclear fusion consists of producing electricity from the coalescence of light nuclei in high temperature plasmas. The most promising route to fusion envisages the confinement of such plasmas with magnetic fields, whose most studied configuration is the tokamak. Disruptions are catastrophic collapses affecting all tokamak devices and one of the main potential showstoppers on the route to a commercial reactor. In this work we report how, deploying innovative analysis methods on thousands of JET experiments covering the isotopic compositions from hydrogen to full tritium and including the major D-T campaign, the nature of the various forms of collapse is investigated in all phases of the discharges. An original approach to proximity detection has been developed, which allows determining both the probability of and the time interval remaining before an incoming disruption, with adaptive, from scratch, real time compatible techniques. The results indicate that physics based prediction and control tools can be developed, to deploy realistic strategies of disruption avoidance and prevention, meeting the requirements of the next generation of devices.
|
|
| 16. |
|
|
| 17. |
|
|
| 18. |
|
|
| 19. |
|
|
| 20. |
|
|
| 21. |
|
|
| 22. |
|
|
| 23. |
|
|
| 24. |
|
|
| 25. |
|
|
| 26. |
|
|
| 27. |
|
|
| 28. |
|
|
| 29. |
- Joffrin, E., et al.
(författare)
-
Overview of the JET preparation for deuterium-tritium operation with the ITER like-wall
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:11
-
Forskningsöversikt (refereegranskat)abstract
- For the past several years, the JET scientific programme (Pamela et al 2007 Fusion Eng. Des. 82 590) has been engaged in a multi-campaign effort, including experiments in D, H and T, leading up to 2020 and the first experiments with 50%/50% D-T mixtures since 1997 and the first ever D-T plasmas with the ITER mix of plasma-facing component materials. For this purpose, a concerted physics and technology programme was launched with a view to prepare the D-T campaign (DTE2). This paper addresses the key elements developed by the JET programme directly contributing to the D-T preparation. This intense preparation includes the review of the physics basis for the D-T operational scenarios, including the fusion power predictions through first principle and integrated modelling, and the impact of isotopes in the operation and physics of D-T plasmas (thermal and particle transport, high confinement mode (H-mode) access, Be and W erosion, fuel recovery, etc). This effort also requires improving several aspects of plasma operation for DTE2, such as real time control schemes, heat load control, disruption avoidance and a mitigation system (including the installation of a new shattered pellet injector), novel ion cyclotron resonance heating schemes (such as the three-ions scheme), new diagnostics (neutron camera and spectrometer, active Alfven eigenmode antennas, neutral gauges, radiation hard imaging systems...) and the calibration of the JET neutron diagnostics at 14 MeV for accurate fusion power measurement. The active preparation of JET for the 2020 D-T campaign provides an incomparable source of information and a basis for the future D-T operation of ITER, and it is also foreseen that a large number of key physics issues will be addressed in support of burning plasmas.
|
|
| 30. |
-
Overview of the JET results
- 2015
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:10
-
Tidskriftsartikel (refereegranskat)
|
|
| 31. |
-
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:9
-
Tidskriftsartikel (refereegranskat)
|
|
| 32. |
- Abel, I, et al.
(författare)
-
Overview of the JET results with the ITER-like wall
- 2013
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 53:10, s. 104002-
-
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.
|
|
| 33. |
- Romanelli, F, et al.
(författare)
-
Overview of the JET results
- 2011
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 51:9
-
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.
|
|
| 34. |
- Labit, B., et al.
(författare)
-
Dependence on plasma shape and plasma fueling for small edge-localized mode regimes in TCV and ASDEX Upgrade
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:8
-
Tidskriftsartikel (refereegranskat)abstract
- © 2019 Institute of Physics Publishing. All rights reserved. Within the EUROfusion MST1 work package, a series of experiments has been conducted on AUG and TCV devices to disentangle the role of plasma fueling and plasma shape for the onset of small ELM regimes. On both devices, small ELM regimes with high confinement are achieved if and only if two conditions are fulfilled at the same time. Firstly, the plasma density at the separatrix must be large enough (ne,sep/nG ∼ 0.3), leading to a pressure profile flattening at the separatrix, which stabilizes type-I ELMs. Secondly, the magnetic configuration has to be close to a double null (DN), leading to a reduction of the magnetic shear in the extreme vicinity of the separatrix. As a consequence, its stabilizing effect on ballooning modes is weakened.
|
|
| 35. |
- Stroth, U., et al.
(författare)
-
Progress from ASDEX Upgrade experiments in preparing the physics basis of ITER operation and DEMO scenario development
- 2022
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 62:4
-
Tidskriftsartikel (refereegranskat)abstract
- An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L-H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced D α H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts.
|
|
| 36. |
- Pegourie, B., et al.
(författare)
-
Deuterium inventory in Tore Supra : Coupled carbon-deuterium balance
- 2013
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 438:Suppl., s. S120-S125
-
Tidskriftsartikel (refereegranskat)abstract
- This paper presents an analysis of the carbon-deuterium circulation and the resulting balance in Tore Supra over the period 2002-2007. Carbon balance combines the estimation of carbon gross erosion from spectroscopy, net erosion and deposition using confocal microscopy, lock-in thermography and SEM, and a measure of the amount of deposits collected in the vacuum chamber. Fuel retention is determined from post-mortem (PM) analyses and gas balance (GB) measurements. Special attention was paid to the deuterium outgassed during the nights and weekends of the experimental campaign (vessel under vacuum, Plasma Facing Components at 120 degrees C) and during vents (vessel at atmospheric pressure, PFCs at room temperature). It is shown that this outgassing is the main process reconciling the PM and GB estimations of fuel retention, closing the coupled carbon-deuterium balance. In particular, it explains why the deuterium concentration in deposits decreases with increasing depth.
|
|
| 37. |
- Loarer, T., et al.
(författare)
-
Gas balance and fuel retention in fusion devices
- 2007
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 47:9, s. 1112-1120
-
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.
|
|
| 38. |
- Pitts, R. A., et al.
(författare)
-
Material erosion and migration in tokamaks
- 2005
-
Ingår i: Plasma Physics and Controlled Fusion. - 0741-3335 .- 1361-6587. ; 47, s. B303-B322
-
Tidskriftsartikel (refereegranskat)abstract
- The issue of first wall and divertor target lifetime represents one of the greatest challenges facing the successful demonstration of integrated tokamak burning plasma operation, even in the case of the planned next step device, ITER, which will run at a relatively low duty cycle in comparison to future fusion power plants. Material erosion by continuous or transient plasma ion and neutral impact, the susbsequent transport of the released impurities through and by the plasma and their deposition and/or eventual re-erosion constitute the process of migration. Its importance is now recognized by a concerted research effort throughout the international tokamak community, comprising a wide variety of devices with differing plasma configurations, sizes and plasmafacing component material. No single device, however, operates with the first wall material mix currently envisaged for ITER, and all are far from the ITER energy throughput and divertor particle fluxes and fluences. This paper aims to review the basic components of material erosion and migration in tokamaks, illustrating each by way of examples from current research and attempting to place them in the context of the next step device. Plans for testing an ITER-like first wall material mix on the JET tokamak will also be briefly outlined.
|
|
| 39. |
- Rosanvallon, S., et al.
(författare)
-
Tritium related studies within the JET Fusion Technology work programme
- 2005
-
Ingår i: Fusion science and technology. - 1536-1055 .- 1943-7641. ; 48:1, s. 268-273
-
Tidskriftsartikel (refereegranskat)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.
|
|
| 40. |
- Widdowson, A., et al.
(författare)
-
Efficacy of photon cleaning of JET divertor tiles
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 341-345
-
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
|
|
| 41. |
- Coad, J. P., et al.
(författare)
-
Diagnostics for studying deposition and erosion processes in JET
- 2005
-
Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 74:1-4, s. 745-749
-
Tidskriftsartikel (refereegranskat)abstract
- Estimates of erosion, deposition and H-isotope retention in JET from previous divertor campaigns have relied on analysis of in-vessel components removed at shutdowns. The components analysed have also provided an incomplete coverage of the vessel. In 2004, new diagnostics are being installed to give a more complete picture (such as smart tiles) and to provide some time resolution. The latter includes further quartz microbalances (QMB), following the successful operation of a prototype in 2002-2004 [H.-G. Esser, G. Neill, P. Coad, G.F. Matthews, D. Jolovic, D. Wilson, M. Freisinger, V. Philipps, Quartz microbalance: a time-resolved diagnostic to measure material deposition in JET, Fusion Eng. Des. 66-68 (2003) 855-860; H.-G. Esser, V. Philipps, M. Freisinger, G.F. Matthews, J.P. Coad, G.F. Neill, JET EFDA Contributors, Effect of plasma configuration on carbon migration measured in the inner divertor of JET using quartz microbalance, J. Nucl. Mater. 337-339 (2005) 84-87], which will also have temperature control. Other diagnostics include rotating collectors and deposition monitors [M. Mayer, V. Rohde, P. Coad, P. Wienhold, ASDEX Upgrade Team, JET EFDA Contributors, Carbon erosion and migration in fusion devices, Phys. Scr. T111 (2004) 55-59]. Units are also being installed to provide information on mirrors for ITER.
|
|
| 42. |
- Coad, J. P., et al.
(författare)
-
Erosion and deposition in the JET MkII-SRP divertor
- 2007
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 363, s. 287-293
-
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
|
|
| 43. |
- Hirai, T., et al.
(författare)
-
Thermal load testing of erosion-monitoring beryllium marker tile for the ITER-Like Wall Project at JET
- 2008
-
Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 83:7-9, s. 1072-1076
-
Tidskriftsartikel (refereegranskat)abstract
- ITER-Like Wall Project has been launched at JET in order to perform a fully integrated test of plasma-facing materials. During the next major shutdown a full metal wall will be installed: tungsten in the divertor and beryllium in the main chamber. Beryllium erosion is one of key issues to be addressed. Special marker tiles have been designed for this purpose. Test coupons of such markers have been manufactured and examined. The performance test under high power deposition was carried in the electron beam facility JUDITH. The results of material characterization before and after high heat flux loads are presented. The samples survived, without macroscopic damage, power loads of up to 4.5 MW/m(2) for 10s (surface temperature similar to 650 degrees C) and 50 cyclic loads at 3.5 MW/m(2) lasting 10s each (surface temperature similar to 600 degrees C).
|
|
| 44. |
- Rubel, Marek J., et al.
(författare)
-
Beryllium plasma-facing components for the ITER-Like Wall Project at JET
- 2008
-
Ingår i: PROCEEDINGS OF THE 17TH INTERNATIONAL VACUUM CONGRESS/13TH INTERNATIONAL CONFERENCE ON SURFACE SCIENCE/INTERNATIONAL CONFERENCE ON NANOSCIENCE AND TECHNOLOGY. - : IOP Publishing.
-
Konferensbidrag (refereegranskat)abstract
- ITER-Like Wall Project has been launched at the JET tokamak in order to study a tokamak operation with beryllium components on the main chamber wall and tungsten in the divertor. To perform this first comprehensive test of both materials in a thermonuclear fusion environment, a broad program has been undertaken to develop plasma-facing components and assess their performance under high power loads. The paper provides a concise report on scientific and technical issues in the development of a beryllium first wall at JET.
|
|
| 45. |
- Strachan, J. D., et al.
(författare)
-
Modelling of carbon migration during JET C-13 injection experiments
- 2008
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 48:10
-
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.
|
|
| 46. |
- Tsitrone, E., et al.
(författare)
-
Multi machine scaling of fuel retention in 4 carbon dominated tokamaks
- 2011
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 415:1, s. S735-S739
-
Tidskriftsartikel (refereegranskat)abstract
- In order to benchmark predictions for the in vessel tritium inventory in ITER, a survey of fuel retention measured in 4 carbon dominated tokamaks (TEXTOR, ASDEX Upgrade in the 2002-2003 carbon configuration, Tore Supra and JET) was performed, showing retention rates from similar to 1 g D/h in TEXTOR (L mode, limiter machine) up to similar to 6-12 g D/h in AUG (H mode, divertor machine). A simple scaling used for ITER predictions is applied for comparison with experimental values: (1) estimate of wall fluxes, (2) estimate of the gross carbon erosion, (3) estimate of the net erosion/redeposition assuming a redeposition fraction and (4) estimate of the retention rate using D/C ratio scalings. The validity of each step is discussed, showing that this approach yields the right order of magnitude, but tends to underestimate the experimental values unless a high wall flux, a low local redeposition fraction and/or a high D/C ratio are used.
|
|
| 47. |
- Airila, M. I., et al.
(författare)
-
ERO modelling of local deposition of injected C-13 tracer at the outer divertor of JET
- 2009
-
Ingår i: Physica Scripta. - 0031-8949 .- 1402-4896. ; T138, s. 014021-
-
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.
|
|
| 48. |
- Bergsåker, Henric, et al.
(författare)
-
First results from the Be-10 marker experiment in JET with ITER-like wall
- 2014
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 54:8, s. 082004-
-
Tidskriftsartikel (refereegranskat)abstract
- When the ITER-like wall was installed in JET, one of the 218 Be inner wall guard limiter tiles had been enriched with Be-10 as a bulk isotopic marker. During the shutdown in 2012-2013, a set of tiles were sampled nondestructively to collect material for accelerator mass spectroscopy measurements of Be-10 concentration. The letter shows how the marker experiment was set up, presents first results and compares them to preliminary predictions of marker redistribution, made with the ASCOT numerical code. Finally an outline is shown of what experimental data are likely to become available later and the possibilities for comparison with modelling using the WallDYN, ERO and ASCOT codes are discussed.
|
|
| 49. |
- Brezinsek, S., et al.
(författare)
-
Beryllium migration in JET ITER-like wall plasmas
- 2015
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 55:6
-
Tidskriftsartikel (refereegranskat)abstract
- JET is used as a test bed for ITER, to investigate beryllium migration which connects the lifetime of first-wall components under erosion with tokamak safety, in relation to long-term fuel retention. The (i) limiter and the (ii) divertor configurations have been studied in JET-ILW (JET with a Be first wall and W divertor), and compared with those for the former JET-C (JET with carbon-based plasma-facing components (PFCs)). (i) For the limiter configuration, the Be gross erosion at the contact point was determined in situ by spectroscopy as between 4% (E-in = 35 eV) and more than 100%, caused by Be self-sputtering (E-in = 200 eV). Chemically assisted physical sputtering via BeD release has been identified to contribute to the effective Be sputtering yield, i.e. at E-in = 75 eV, erosion was enhanced by about 1/3 with respect to the bare physical sputtering case. An effective gross yield of 10% is on average representative for limiter plasma conditions, whereas a factor of 2 difference between the gross erosion and net erosion, determined by post-mortem analysis, was found. The primary impurity source in the limiter configuration in JET-ILW is only 25% higher (in weight) than that for the JET-C case. The main fraction of eroded Be stays within the main chamber. (ii) For the divertor configuration, neutral Be and BeD from physically and chemically assisted physical sputtering by charge exchange neutrals and residual ion flux at the recessed wall enter the plasma, ionize and are transported by scrape-off layer flows towards the inner divertor where significant net deposition takes place. The amount of Be eroded at the first wall (21 g) and the Be amount deposited in the inner divertor (28 g) are in fair agreement, though the balancing is as yet incomplete due to the limited analysis of PFCs. The primary impurity source in the JET-ILW is a factor of 5.3 less in comparison with that for JET-C, resulting in lower divertor material deposition, by more than one order of magnitude. Within the divertor, Be performs far fewer re-erosion and transport steps than C due to an energetic threshold for Be sputtering, and inhibits as a result of this the transport to the divertor floor and the pump duct entrance. The target plates in the JET-ILW inner divertor represent at the strike line a permanent net erosion zone, in contrast to the net deposition zone in JET-C with thick carbon deposits on the CFC (carbon-fibre composite) plates. The Be migration identified is consistent with the observed low long-term fuel retention and dust production with the JET-ILW.
|
|
| 50. |
- Coad, J. P., et al.
(författare)
-
Distribution of hydrogen isotopes, carbon and beryllium on in-vessel surfaces in the various jet divertors
- 2005
-
Ingår i: Fusion science and technology. - 1536-1055 .- 1943-7641. ; 48:1, s. 551-556
-
Tidskriftsartikel (refereegranskat)abstract
- JET has operated with divertors of differing geometries since 1994. Impurities accumulated in the inner leg of all the divertors, and operation of the first (Mk I) divertor with beryllium tiles demonstrated that most are eroded from the main chamber walls and swept along the scrape-off layer to the inner divertor. Carbon deposited at the inner divertor is then locally transported to shadowed regions such as the inner louvres, where, for example, most of the tritium was trapped during the deuterium-tritium experiment (DTE1). Factors affecting these transport processes (e.g. temperature) are important for ITER, but are not well understood.
|
|
