| 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. |
-
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:1
-
Forskningsöversikt (refereegranskat)
|
|
| 3. |
|
|
| 4. |
|
|
| 5. |
|
|
| 6. |
|
|
| 7. |
|
|
| 8. |
|
|
| 9. |
|
|
| 10. |
|
|
| 11. |
|
|
| 12. |
|
|
| 13. |
|
|
| 14. |
|
|
| 15. |
|
|
| 16. |
|
|
| 17. |
|
|
| 18. |
|
|
| 19. |
|
|
| 20. |
|
|
| 21. |
|
|
| 22. |
|
|
| 23. |
- 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.
|
|
| 24. |
- 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.
|
|
| 25. |
- 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.
|
|
| 26. |
- Meyer, H.F., et al.
(författare)
-
Overview of physics studies on ASDEX Upgrade
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 59:11
-
Forskningsöversikt (refereegranskat)abstract
- The ASDEX Upgrade (AUG) programme, jointly run with the EUROfusion MST1 task force, continues to significantly enhance the physics base of ITER and DEMO. Here, the full tungsten wall is a key asset for extrapolating to future devices. The high overall heating power, flexible heating mix and comprehensive diagnostic set allows studies ranging from mimicking the scrape-off-layer and divertor conditions of ITER and DEMO at high density to fully non-inductive operation (q 95 = 5.5, ) at low density. Higher installed electron cyclotron resonance heating power 6 MW, new diagnostics and improved analysis techniques have further enhanced the capabilities of AUG. Stable high-density H-modes with MW m-1 with fully detached strike-points have been demonstrated. The ballooning instability close to the separatrix has been identified as a potential cause leading to the H-mode density limit and is also found to play an important role for the access to small edge-localized modes (ELMs). Density limit disruptions have been successfully avoided using a path-oriented approach to disruption handling and progress has been made in understanding the dissipation and avoidance of runaway electron beams. ELM suppression with resonant magnetic perturbations is now routinely achieved reaching transiently . This gives new insight into the field penetration physics, in particular with respect to plasma flows. Modelling agrees well with plasma response measurements and a helically localised ballooning structure observed prior to the ELM is evidence for the changed edge stability due to the magnetic perturbations. The impact of 3D perturbations on heat load patterns and fast-ion losses have been further elaborated. Progress has also been made in understanding the ELM cycle itself. Here, new fast measurements of and E r allow for inter ELM transport analysis confirming that E r is dominated by the diamagnetic term even for fast timescales. New analysis techniques allow detailed comparison of the ELM crash and are in good agreement with nonlinear MHD modelling. The observation of accelerated ions during the ELM crash can be seen as evidence for the reconnection during the ELM. As type-I ELMs (even mitigated) are likely not a viable operational regime in DEMO studies of 'natural' no ELM regimes have been extended. Stable I-modes up to have been characterised using -feedback. Core physics has been advanced by more detailed characterisation of the turbulence with new measurements such as the eddy tilt angle - measured for the first time - or the cross-phase angle of and fluctuations. These new data put strong constraints on gyro-kinetic turbulence modelling. In addition, carefully executed studies in different main species (H, D and He) and with different heating mixes highlight the importance of the collisional energy exchange for interpreting energy confinement. A new regime with a hollow profile now gives access to regimes mimicking aspects of burning plasma conditions and lead to nonlinear interactions of energetic particle modes despite the sub-Alfvénic beam energy. This will help to validate the fast-ion codes for predicting ITER and DEMO.
|
|
| 27. |
- Litaudon, X., et al.
(författare)
-
14 MeV calibration of JET neutron detectors-phase 1: Calibration and characterization of the neutron source
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:2
-
Tidskriftsartikel (refereegranskat)abstract
- In view of the planned DT operations at JET, a calibration of the JET neutron monitors at 14 MeV neutron energy is needed using a 14 MeV neutron generator deployed inside the vacuum vessel by the JET remote handling system. The target accuracy of this calibration is 10% as also required by ITER, where a precise neutron yield measurement is important, e.g. for tritium accountancy. To achieve this accuracy, the 14 MeV neutron generator selected as the calibration source has been fully characterised and calibrated prior to the in-vessel calibration of the JET monitors. This paper describes the measurements performed using different types of neutron detectors, spectrometers, calibrated long counters and activation foils which allowed us to obtain the neutron emission rate and the anisotropy of the neutron generator, i.e.The neutron flux and energy spectrum dependence on emission angle, and to derive the absolute emission rate in 4π sr. The use of high resolution diamond spectrometers made it possible to resolve the complex features of the neutron energy spectra resulting from the mixed D/T beam ions reacting with the D/T nuclei present in the neutron generator target. As the neutron generator is not a stable neutron source, several monitoring detectors were attached to it by means of an ad hoc mechanical structure to continuously monitor the neutron emission rate during the in-vessel calibration. These monitoring detectors, two diamond diodes and activation foils, have been calibrated in terms of neutrons/counts within ± 5% total uncertainty. A neutron source routine has been developed, able to produce the neutron spectra resulting from all possible reactions occurring with the D/T ions in the beam impinging on the Ti D/T target. The neutron energy spectra calculated by combining the source routine with a MCNP model of the neutron generator have been validated by the measurements. These numerical tools will be key in analysing the results from the in-vessel calibration and to derive the response of the JET neutron detectors to DT plasma neutrons starting from the response to the generator neutrons, and taking into account all the calibration circumstances.
|
|
| 28. |
- 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.
|
|
| 29. |
- 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.
|
|
| 30. |
- 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.
|
|
| 31. |
- 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.
|
|
| 32. |
- 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
|
|
| 33. |
- 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.
|
|
| 34. |
- 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
|
|
| 35. |
- 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).
|
|
| 36. |
|
|
| 37. |
- 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.
|
|
| 38. |
- 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.
|
|
| 39. |
- 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.
|
|
| 40. |
- Widdowson, A., et al.
(författare)
-
Testing of beryllium marker coatings in PISCES-B for the JET ITER-like wall
- 2009
-
Ingår i: Journal of Nuclear Materials. - : Elsevier BV. - 0022-3115 .- 1873-4820. ; 390-91, s. 988-991
-
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.
|
|
| 41. |
- 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.
|
|
| 42. |
- Badziak, J., et al.
(författare)
-
Application of ion diagnostics to control the laser-induced removal of surface layer of a carbon substrate
- 2006
-
Ingår i: Plasma 2005. - : AIP. - 073540304X ; , s. 295-298
-
Konferensbidrag (refereegranskat)abstract
- Among other methods of detritiation of in-vessel tokamak components the application of lasers for removal of fuel trapped in co-deposited layers is under investigation. The paper presents preparation and tests of ion diagnostic methods for on-line measurement of the amount and characteristics of ablated carbon, hydrogen/deuterium and contaminant species from the graphite target (plate) of the main toroidal limiter of the TEXTOR tokamak. For removal of the surface layer from the graphite limiter plate Nd:YAG laser was used. Determination of the characteristics of laser-produced ions has been performed by means of ion collectors and an electrostatic ion-energy analyser. The main ion stream parameters were measured depending on the number of laser shots and the laser power density on the target surface. The properties of modified carbon sample surface were determined with the use of optical methods and compared with the results of the ion measurements.
|
|
| 43. |
- 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.
|
|
| 44. |
- Coad, J. P., et al.
(författare)
-
Deposition results from rotating collector diagnostics in JET
- 2009
-
Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847. ; T138, s. 014023-
-
Tidskriftsartikel (refereegranskat)abstract
- Rotating collectors (RC) were installed in JET during the period 2005-2007, each providing a time-resolved deposition pattern on the surface of a rotating silicon disc, which could be analysed once retrieved from the vessel. This paper reports results from the silicon disc removed from the RC located under the load-bearing septum replacement plate in JET in 2007. Nuclear reaction analysis results of the deposits on the disc have been correlated directly with the pattern of erosion and deposition observed by the quartz microbalance (QMB) located in an equivalent position. The thickest film in the time-resolved region (i.e. deposited in similar to 60 pulses) was similar to 250 nm, and the Be/C ratio was generally found to be 0.1 or lower, with two regions where the ratio rose to 0.2. The deposition observed with the QMB appears to be about a factor of four less.
|
|
| 45. |
- 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.
|
|
| 46. |
- Coad, J. P., et al.
(författare)
-
Erosion/deposition in JET during the period 1999-2001
- 2003
-
Ingår i: Journal of Nuclear Materials. - 0022-3115 .- 1873-4820. ; 313, s. 419-423
-
Tidskriftsartikel (refereegranskat)abstract
- Coated divertor and wall tiles exposed in JET for the 1999-2001 operations have been used to assess erosion/deposition. Deposited films of up to 90 mum thickness at the inner wall of the divertor tiles are, for the most part, enriched in beryllium and other metals, whilst carbon is probably chemically sputtered from these tiles and transported to shadowed regions of the inner divertor. However, from the composition at the surface of the tiles, it appears that the chemical erosion was 'switched off' by reducing the JET vessel wall temperature for the last part of the operations to 200 degreesC. Thick powdery deposits localised at the ion transport limit at each corner of the divertor may be due to physical sputtering. Erosion of the coatings is seen at the outer divertor wall, and on all the inner wall and outer limiter tiles.
|
|
| 47. |
- Coad, J. P., et al.
(författare)
-
Overview of material re-deposition and fuel retention studies at JET with the Gas Box divertor
- 2006
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 46:2, s. 350-366
-
Tidskriftsartikel (refereegranskat)abstract
- in the period 1998-2001 the JET tokamak was operated with the MkII Gas Box divertor. On two occasions during that period a number of limiter and divertor tiles were retrieved from the torus and then examined ex situ with surface sensitive techniques. Erosion and deposition patterns were determined in order to assess the material erosion, material migration and fuel inventory on plasma facing components. Tracer techniques, e.g. injection of C-13 labelled methane and tiles coated with a low-Z and high-Z marker layer, were used to enhance the volume of information on the material transport. The results show significant asymmetry in the distribution of fuel and plasma impurity species between the inner (net deposition area) and the outer (net erosion) divertor channels. No significant formation of highly hydrogenated carbon films has been found in the Gas Box structure. The important processes for material migration, and the influence of operation scenarios on the morphology of the deposits are discussed. Comparison is also made with results obtained following previous divertor campaigns.
|
|
| 48. |
- Counsell, G., et al.
(författare)
-
Tritium retention in next step devices and the requirements for mitigation and removal techniques
- 2006
-
Ingår i: Plasma Physics and Controlled Fusion. - 0741-3335 .- 1361-6587. ; 48:12B, s. B189-B199
-
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.
|
|
| 49. |
- De Temmerman, G., et al.
(författare)
-
Interactions of diamond surfaces with fusion relevant plasmas
- 2009
-
Ingår i: Physica scripta. T. - : Institute of Physics Publishing (IOPP). - 0281-1847 .- 0031-8949 .- 1402-4896. ; T138, s. 014013-
-
Tidskriftsartikel (refereegranskat)abstract
- The outstanding thermal properties of diamond and its low reactivity towards hydrogen may make it an attractive plasma-facing material for fusion and calls for a proper evaluation of its behaviour under exposure to fusion-relevant plasma conditions. Micro and nanocrystalline diamond layers, deposited on Mo and Si substrates by hot filament chemical vapour deposition (CVD), have been exposed both in tokamaks and in linear plasma devices to measure the erosion rate of diamond and study the modification of the surface properties induced by particle bombardment. Experiments in Pilot-PSI and PISCES-B have shown that the sputtering yield of diamond (both physical and chemical) was a factor of 2 lower than that of graphite. Exposure to detached plasma conditions in the DIII-D tokamak have evidenced a strong resistance of diamond against erosion under those conditions.
|
|
| 50. |
- Garzotti, L., et al.
(författare)
-
Scenario development for D-T operation at JET
- 2019
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 59:7
-
Tidskriftsartikel (refereegranskat)abstract
- The JET exploitation plan foresees D-T operations in 2020 (DTE2). With respect to the first D-T campaign in 1997 (DTE1), when JET was equipped with a carbon wall, the experiments will be conducted in presence of a beryllium-tungsten ITER-like wall and will benefit from an extended and improved set of diagnostics and higher additional heating power (32 MW neutral beam injection + 8 MW ion cyclotron resonance heating). There are several challenges presented by operations with the new wall: a general deterioration of the pedestal confinement; the risk of heavy impurity accumulation in the core, which, if not controlled, can cause the radiative collapse of the discharge; the requirement to protect the divertor from excessive heat loads, which may damage it permanently. Therefore, an intense activity of scenario development has been undertaken at JET during the last three years to overcome these difficulties and prepare the plasmas needed to demonstrate stationary high fusion performance and clear alpha particle effects. The paper describes the status and main achievements of this scenario development activity, both from an operational and plasma physics point of view.
|
|
