| 1. |
- Bécoulet, A., et al.
(författare)
-
Science and technology research and development in support to ITER and the Broader Approach at CEA
- 2013
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 53:10
-
Tidskriftsartikel (refereegranskat)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.
|
|
| 2. |
-
- 2018
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 1741-4326 .- 0029-5515. ; 58:1
-
Forskningsöversikt (refereegranskat)
|
|
| 3. |
- Corre, Y., et al.
(författare)
-
Testing of ITER-grade plasma facing units in the WEST tokamak: Progress in understanding heat loading and damage mechanisms
- 2023
-
Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Tidskriftsartikel (refereegranskat)abstract
- Assessing the performance of the ITER design for the tungsten (W) divertor Plasma Facing Units (PFUs) in a tokamak environment is a high priority issue to ensure efficient plasma operation. This paper reviews the most recent results derived from experiments and post-mortem analysis of the ITER-grade PFUs exposed in the WEST tokamak and the associated modelling, with a focus on understanding heat loading and damage evolution. Several shaping options, sharp or chamfered leading edge (LE), unshaped or shaped blocks with a toroidal bevel as foreseen in ITER, were investigated, under steady state heat fluxes of up to 120 MW⋅m−2 and 6 MW⋅m−2 on the sharp LE and top surface of the block, respectively. A very high spatial resolution (VHR) infrared (IR) camera (0.1 mm/pixel) was used to derive the temporal and surface distribution of the temperature and heat load on the castellated tungsten blocks for different geometric alignment and plasma conditions. Photonic modelling was required to reproduce the IR measurements in particular in the toroidal and poloidal gaps of the mono-block (MB) stacks where high apparent temperatures are observed. Specular reflection is found to be the dominant emitter in these parts of the blocks. W-cracking was observed on the leading edge of the blocks already within the first phase of plasma operation, during which the divertor was equipped with unshaped PFUs, including some intentionally misaligned blocks. Numerical analysis taking into account softening processes and mechanical stresses, revealed brittle failure due to transients as the dominant failure mechanisms. Ductile failure was observed in one particular block used for the melting experiment, therefore under extremely high steady state heat load conditions. W-melting achieved on actively cooled PFU exhibits specific features: shallow melting and slow melt displacement. Plasma exposure of pre-damaged PFUs at various damage levels (crack network and melted droplets) was carried out under high heat load conditions with several hours of cumulated plasma duration. IR data and preliminary surface analyses show no evidence of significant degradation damage progression under these conditions.
|
|
| 4. |
- Dejarnac, R., et al.
(författare)
-
Numerical evaluation of heat flux and surface temperature on a misaligned JET divertor W lamella during ELMs
- 2014
-
Ingår i: Nuclear Fusion. - : IOP Publishing. - 0029-5515 .- 1741-4326. ; 54:12, s. 123011-
-
Tidskriftsartikel (refereegranskat)abstract
- A series of experiments has been performed on JET to investigate the dynamics of transient melting due to edge localized modes (ELMs). The experiment employs a deliberately misaligned lamella in one module of the JET bulk tungsten outer divertor, allowing the combination of stationary power flux and ELMs to transiently melt the misaligned edge. During the design of the experiment a number of calculations were performed using 2D particle-in-cell simulations and a heat transfer code to investigate the influence on the deposited power flux of finite Larmor radius effects associated with the energetic ELM ions. This has been performed using parameter scans inside a range of pedestal temperatures and densities to scope different experimentally expected ELM energies. On the one hand, we observe optimistic results, with smoothing of the heat flux due to the Larmor gyration on the protruding side of the lamella which sees the direct parallel flux-the deposited power tends to be lower than the nominal value expected from geometric magnetic field line impact over a distance smaller than 2 Larmor radii, a finding which is always valid during ELMs for such a geometry. On the other hand, the fraction of the flux not reaching the directly wetted side is transferred and spread to the top surface of the lamella. The hottest point of the lamella (corner side/top) does not always benefit from the gain from the Larmor smoothing effect because of an enhanced power deposition from the second contribution.
|
|
| 5. |
- Komm, M., et al.
(författare)
-
On thermionic emission from plasma-facing components in tokamak-relevant conditions
- 2017
-
Ingår i: Plasma Physics and Controlled Fusion. - : Institute of Physics Publishing (IOPP). - 0741-3335 .- 1361-6587. ; 59:9
-
Tidskriftsartikel (refereegranskat)abstract
- The first results of particle-in-cell simulations of the electrostatic sheath and magnetic pre-sheath of thermionically emitting planar tungsten surfaces in fusion plasmas are presented. Plasma conditions during edge localized modes (ELMs) and during inter-ELM periods have been considered for various inclinations of the magnetic field and for selected surface temperatures. All runs have been performed under two assumptions for the sheath potential drop; fixed or floating. The primary focus lies on the evaluation of the escaping thermionic current and the quantification of the suppression due to the combined effects of space-charge and Larmor gyration. When applicable, the results are compared with the predictions of analytical models. The heat balance in the presence of thermionic emission as well as the contribution of the escaping thermionic current to surface cooling are also investigated. Regimes are identified where emission needs to be considered in the energy budget.
|
|
| 6. |
- Komm, M., et al.
(författare)
-
Simulations of thermionic suppression during tungsten transient melting experiments
- 2017
-
Ingår i: Physica Scripta. - : Institute of Physics (IOP). - 0031-8949 .- 1402-4896. ; T170
-
Tidskriftsartikel (refereegranskat)abstract
- Plasma-facing components receive enormous heat fluxes under steady state and especially during transient conditions that can even lead to tungsten (W) melting. Under these conditions, the unimpeded thermionic current density emitted from the W surfaces can exceed the incident plasma current densities by several orders of magnitude triggering a replacement current which drives melt layer motion via the J x B force. However, in tokamaks, the thermionic current is suppressed by space-charge effects and prompt re-deposition due to gyro-rotation. We present comprehensive results of particle-in-cell modelling using the 2D3V code SPICE2 for the thermionic emissive sheath of tungsten. Simulations have been performed for various surface temperatures and selected inclinations of the magnetic field corresponding to the leading edge and sloped exposures. The surface temperature dependence of the escaping thermionic current and its limiting value are determined for various plasma parameters; for the leading edge geometry, the results agree remarkably well with the Takamura analytical model. For the sloped geometry, the limiting value is observed to be proportional to the thermal electron current and a simple analytical expression is proposed that accurately reproduces the numerical results.
|
|
| 7. |
- Meyer, H., et al.
(författare)
-
Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution
- 2017
-
Ingår i: Nuclear Fusion. - : Institute of Physics Publishing (IOPP). - 0029-5515 .- 1741-4326. ; 57:10
-
Tidskriftsartikel (refereegranskat)abstract
- Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n = 2 RMP maintaining good confinement H-H(98,H-y2) approximate to 0.95. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes.
|
|
| 8. |
- Meyer, H., et al.
(författare)
-
Overview of progress in European medium sized tokamaks towards an integrated plasma-edge/wall solution
- 2017
-
Ingår i: Nuclear Fusion. - : Institute of Physics Publishing (IOPP). - 0029-5515 .- 1741-4326. ; 57:10
-
Tidskriftsartikel (refereegranskat)abstract
- Integrating the plasma core performance with an edge and scrape-off layer (SOL) that leads to tolerable heat and particle loads on the wall is a major challenge. The new European medium size tokamak task force (EU-MST) coordinates research on ASDEX Upgrade (AUG), MAST and TCV. This multi-machine approach within EU-MST, covering a wide parameter range, is instrumental to progress in the field, as ITER and DEMO core/pedestal and SOL parameters are not achievable simultaneously in present day devices. A two prong approach is adopted. On the one hand, scenarios with tolerable transient heat and particle loads, including active edge localised mode (ELM) control are developed. On the other hand, divertor solutions including advanced magnetic configurations are studied. Considerable progress has been made on both approaches, in particular in the fields of: ELM control with resonant magnetic perturbations (RMP), small ELM regimes, detachment onset and control, as well as filamentary scrape-off-layer transport. For example full ELM suppression has now been achieved on AUG at low collisionality with n = 2 RMP maintaining good confinement H-H(98,H-y2) approximate to 0.95. Advances have been made with respect to detachment onset and control. Studies in advanced divertor configurations (Snowflake, Super-X and X-point target divertor) shed new light on SOL physics. Cross field filamentary transport has been characterised in a wide parameter regime on AUG, MAST and TCV progressing the theoretical and experimental understanding crucial for predicting first wall loads in ITER and DEMO. Conditions in the SOL also play a crucial role for ELM stability and access to small ELM regimes.
|
|
| 9. |
- Tichit, Q., et al.
(författare)
-
Infrared detection of tungsten cracking on actively cooled ITER-like component during high power experiment in WEST
- 2023
-
Ingår i: Nuclear Materials and Energy. - : Elsevier BV. - 2352-1791. ; 37
-
Tidskriftsartikel (refereegranskat)abstract
- The consequences of tungsten (W) damaging processes, such as cracking and melting, on divertor lifetime and plasma operation are high priority issues for ITER. A sustained melting experiment was conducted in WEST using a 2 mm deep groove geometry on the upstream mono-block (MB) to overexpose the sharp leading edge (LE) of the downstream MB. W-cracking has been evidenced for the first time with a very high spatial resolution infrared camera before tungsten melting was reached. These cracks develop when the monoblock temperature is about 2600 degrees C, thus higher than both ductile to brittle transition and softening threshold of tungsten, suggesting that these cracks are different from the ones observed in previous campaigns where brittle failure was involved, because of transient events on cold monoblock. Post-exposure analyses have been performed on the damaged monoblock, highlighting 12 main cracks on the LE, with a width varying from 33 mu m to 77 mu m, and an average spacing of 0.45 mm. Parallel heat flux about 90 MW/m2 has been derived from infrared temperature measurements, with a heat flux decay length on the target of 4 mm. The T-REX modelling code suggest here that with these thermal inputs, a crack can initiates due to thermal cycling without disruption, with a ductile failure, under 1 to 5 cycles for a tungsten DBTT varying from 400 degrees C to 500 degrees C.
|
|
| 10. |
- Van Oost, G., et al.
(författare)
-
The role of radial electric fields in the tokamaks TEXTOR-94, CASTOR, and T-10
- 2001
-
Ingår i: Czechoslovak Journal of Physics. - 0011-4626 .- 1572-9486. ; 51:10, s. 957-975
-
Tidskriftsartikel (refereegranskat)abstract
- Radial electric fields (E-r) and their role in the establishment of edge transport barriers and improved confinement have been studied in the tokamaks TEXTOR-94 and CASTOR, where E-r is externally applied to the plasma in a controlled way using a biased electrode, as well as in the tokamak T-10 where an edge transport barrier (H-mode) is obtained during electron-cyclotron resonance heating (ECRH) of the plasma. The physics of radial currents was studied and the radial conductivity in the edge of TEXTOR-94 (R = 1.75 m, a = 0.46 m) was found to be dominated by recycling (ion-neutral collisions) at the last closed flux surface (LCFS) and by parallel viscosity inside the LCFS. From a performance point of view (edge engineering), such electrode biasing was shown to induce a particle transport barrier, a reduction of particle transport, and a concomitant increase in energy confinement. An H-mode-like behaviour can be induced both with positive and negative electric fields. Positive as well as negative electric fields were shown to strongly affect the exhaust of hydrogen, helium, and impurities, not only in the H-mode-like regime. The impact of sheared radial electric fields on turbulent structures and flows at the plasma edge is investigated on the CASTOR tokamak (R = 0.4 m, a = 0.085 m). A non-intrusive biasing scheme that we call separatrix biasing is applied whereby the electrode is located in the scrape-off layer (SOL) with its tip just touching the LCFS. There is evidence of strongly sheared radial electric field and E x B flow, resulting in the formation of a transport barrier at the separatrix. Advanced probe diagnosis of the edge region has shown that the E x B shear rate that arises during separatrix biasing is larger than for standard edge plasma biasing. The plasma flows, especially the poloidal E x B drift velocity, are strongly modified in the sheared region, reaching Mach numbers as high as half the sound speed. The corresponding shear rates (approximate to5 x 10(6) s(-1)) derived from both the flow and electric field profiles are in excellent agreement and are at least an order of magnitude higher than the growth rate of unstable turbulent modes as estimated from fluctuation measurements. During ECRH in the tokamak T-10 (R = 1.5 m, a = 0.3 m), a regime of improved confinement is obtained with features resembling those in the H-mode in other tokamaks. Using a heavy ion beam probe, a narrow potential well is observed near the limiter together with the typical features of the L-H transition. The time evolution of the plasma profiles during L-H and H-L transitions is clearly correlated with that of the density profile and the formation of a transport barrier near the limiter, The edge electric field is initially positive after the onset of ECRH. It changes its sign during the L-H transition and grows till a steady condition is reached. Similar to the biasing experiments in TEXTOR-94 and CASTOR, the experimentally observed transport barrier is a barrier for particles.
|
|
