The WEST project recently launched at Cadarache consists in transforming Tore Supra in an X-point divertor configuration while extending its long pulse capability, in order to test the ITER divertor ...technology. The implementation of a full tungsten actively cooled divertor with plasma facing unit representative of ITER divertor targets will allow addressing risks both in terms of industrial-scale manufacturing and operation of such components. Relevant plasma scenarios are foreseen for extensive testing under high heat load in the 10–20MW/m2 range and ITER-like fluences (1000s pulses). Plasma facing unit monitoring and development of protection strategies will be key elements of the WEST program.
WEST is scheduled to enter into operation in 2016, and will provide a key facility to prepare and be prepared for ITER.
Abstract
The consequences of tungsten (W) melting on divertor lifetime and plasma operation are high priority issues for ITER. Sustained and controlled W-melting experiment has been achieved for the ...first time in WEST on a poloidal sharp leading edge of an actively cooled ITER-like plasma facing unit (PFU). A series of dedicated high power steady state plasma discharges were performed to reach the melting point of tungsten. The leading edge was exposed to a parallel heat flux of about 100 MW.m
−2
for up to 5 s providing a melt phase of about 2 s without noticeable impact of melting on plasma operation (radiated power and tungsten impurity content remained stable at constant input power) and no melt ejection were observed. The surface temperature of the MB was monitored by a high spatial resolution (0.1 mm/pixel) infrared camera viewing the melt zone from the top of the machine. The melting discharge was repeated three times resulting in about 6 s accumulated melting duration leading to material displacement from three similar pools. Cumulated on the overall sustained melting periods, this leads to excavation depth of about 230
μ
m followed by a re-solidified tungsten bump of 200
μ
m in the JxB direction.
A 20 MW/5 GHz lower hybrid current drive (LHCD) system was initially due to be commissioned and used for the second mission of ITER, i.e. the
Q
= 5 steady state target. Though not part of the ...currently planned procurement phase, it is now under consideration for an earlier delivery. In this paper, both physics and technology conceptual designs are reviewed. Furthermore, an appropriate work plan is also developed. This work plan for design, R&D, procurement and installation of a 20 MW LHCD system on ITER follows the ITER Scientific and Technical Advisory Committee (STAC) T13-05 task instructions. It gives more details on the various scientific and technical implications of the system, without presuming on any work or procurement sharing amongst the possible ITER partners
b
The LHCD system of ITER is not part of the initial cost sharing.. This document does not commit the Institutions or Domestic Agencies of the various authors in that respect.
Physics related to fast electrons in lower hybrid (LH) current drive (LHCD) plasma is a very important issue, since these particles will play an important role in runaway electron (RE) generation and ...lower hybrid wave (LHW)-related physics. Utilizing a new hard X-ray (HXR) pinhole camera, recent HL-2A tokamak experiments have devoted to enhancing the understanding of the physics on fast electrons and LHW. The fast electron bremsstrahlung (FEB) emission in the HXR energy range between 20 and 200 keV was measured by the HXR camera. To study the conversion of LHW-produced fast electrons into REs, a very short pulse of LHW, so-called “blip”, with duration of 5 ms was injected into the plasma during the current flattop phase. A strong enhancement of REs was induced by the blip injection. Measurements from the HXR camera show that the fast electrons generated by LHWs is mainly concentrated in 40-60 keV, which is well consistent with the calculated value based on Landau damping theory. The energy of these seed electrons is higher than the critical runaway energy. This phenomenon may be come from the synergetic effects of Dreicer and avalanche RE generation. Moreover, the measurements indicate that the spatial distribution of the fast electrons during LHCD has a peaked profile, implying that the fast electrons are mainly produced in the plasma core. It also suggests that the energy of the LHW mainly deposited in the plasma core region.
Nitrogen (N2) will be used in ITER to enhance the radiative fraction to ∼90%, thereby cooling the edge plasma and preventing damage to the plasma-facing components. However, the reactivity of N2 with ...hydrogen isotopes can lead to the formation of tritiated ammonia (NT3). This should be considered in terms of the in-vessel tritium inventory, the regeneration of the cryo pumps, and the processes in the ITER de-tritiation plant. In the 'W' Environment in Steady-state Tokamak (WEST), a series of long L-mode discharges (∼50 s), with a constant N2 seeding from the outer strike point region has been performed on the upper actively cooled divertor. In the absence of active pumping, the N2 balance shows steady-state retention during plasma discharge, and is partially (∼35%) released in between discharges. Although a significant amount of N2( 18.65 Pa m3) has been injected, the wall still exhibited N2 pumping capabilities. Under these conditions, as long as this N2 reservoir is not saturated, there is not enough N available for the detectable threshold of ND3 formation to be reached. In these WEST experiments, no ammonia is detected during the pulse or after the pulse in the outgassing phase. These results are consistent with and complementary to the N2 seeded experiments performed in the Joint European Torus (JET) with its ITER-like wall and in the Axially Symmetric Divertor Experiment (ASDEX) upgrade.
A new ITER-relevant lower hybrid current drive (LHCD) launcher, based on the passive-active-multijunction (PAM) concept, was brought into operation on the Tore Supra tokamak in autumn 2009. The PAM ...launcher concept was designed in view of ITER to allow efficient cooling of the waveguides, as required for long pulse operation. In addition, it offers low power reflection close to the cut-off density, which is very attractive for ITER, where the large distance between the plasma and the wall may bring the density in front of the launcher to low values. The first experimental campaign on Tore Supra has shown extremely encouraging results in terms of reflected power level and power handling. Power reflection coefficient <2% is obtained at low density in front of the launcher, i.e. close to the cut-off density, and very good agreement between the experimental results and the coupling code predictions is obtained. Long pulse operation at ITER-relevant power density has been demonstrated. The maximum power and energy reached so far is 2.7 MW during 78 s, corresponding to a power density of 25 MW m
−2
, i.e. its design value at
f
= 3.7 GHz. In addition, 2.7 MW has been coupled at a plasma–launcher distance of 10 cm, with a power reflection coefficient <2%. Finally, full non-inductive discharges have been sustained for 50 s with the PAM.
Aiming at high-power and long-pulse operation up to 1000 s, some improvements have been made for both 2.45 GHz and 4.6 GHz lower hybrid (LH) systems during the recent 5 years. At first, the guard ...limiters of the LH antennas with graphite tiles were upgraded to tungsten, the most promising material for plasma facing components in nuclear fusion devices. These new guard limiters can operate at a peak power density of 12.9 MW/m2. Strong hot spots were usually observed on the old graphite limiters when 4.6 GHz system operated with power >2.0 MW B. N. Wan et al., Nucl. Fusion57(2017) 102019, leading to a reduction of the maximum power capability. With the new limiters, 4.6 GHz LH system, the main current drive (CD) and electron heating tool for EAST, can be operated with power >2.5 MW routinely. Long-pulse operation up to 100 s with 4.6 GHz LH power of 2.4 MW was achieved in 2021 and the maximal temperature on the guard limiters measured by an infrared (IR) camera was about 540 °C, much below the permissible value of tungsten material (∼1200 °C). A discharge with a duration of 1056 s was achieved and the 4.6 GHz LH energy injected into the plasma was up to 1.05 GJ. Secondly, the fully-active-multijunction (FAM) launcher of 2.45 GHz system was upgraded to a passive-active-multijunction (PAM), for which the density of optimum coupling was relatively low (below the cut-off value). Good coupling with reflection coefficient ∼3% has been achieved with plasma-antenna distance up to 11 cm for the new PAM. Finally, in order to eliminate the effect of ion cyclotron range of frequencies (ICRF) wave on 4.6 GHz LH wave coupling, the location of the ICRF launcher was changed to a port that is located 157.5° toroidally from the 4.6 GHz LH system and is not magnetically connected.
The new WEST configuration of Tore Supra facility leads to control challenges and the need to handle events of a modern diverted and metallic machine (vertical stabilization, impurity control, …). To ...address them, a new Plasma Control System (PCS) has been built based on the ASDEX upgrade (AUG) real-time framework called DCS (Discharge Control System). This contribution summarizes the work done during the 5 years of the project from the definition of the PCS concepts to its operation during the WEST campaigns. The integration into the Tore Supra control infrastructure is detailed as well as the different real-time control processes. The efficiency and the versatility of the PCS are illustrated by several examples of plasma operation.
The ECRH system formerly used in Tore Supra is being upgraded to start on WEST in 2023, at a power level of 1MW and frequency of 105 GHz. Its ultimate 3MW/1000s capability is expected to enlarge the ...WEST operational domain by increasing margins with respect to H-mode access, and by providing additional flexibility in terms of achievable scenarios using impurity and/or MHD control. This flexibility is made possible using an antenna based on three steerable mirrors for controlled power injection. In order to determine an appropriate range of EC wave injection angles for WEST scenarios, the fast and reliable ray-tracing code REMA has been interfaced with the WEST IMAS database. This allows the EC power damping rate to be quickly assessed, as well as deposition profiles to be predicted in realistic plasma conditions. Based on a typical WEST discharge at central magnetic field B
0
~3.6 T, central line-averaged electron density n
l
~4 × 10
19
m
−3
and central electron temperature T
e0
~3keV, ray-tracing calculations have been performed. Comprehensive poloidal and toroidal angle scans, as well as variations of B
t
, nl and T
e0
with respect to the reference parameters have allowed an adequate range of injection angles to be determined for efficient use of ECRH and/or ECCD in typical WEST scenarios, and compared with the mechanical limits set by the antenna mechanical characteristics. In order to further characterize the effect of this new power source in WEST scenarios, EC wave deposition and current profiles from ray-tracing calculations have been included in integrated simulation codes. It has been shown that this additional power source could allow central electron heating to be achieved, potentially alleviating the issue of radiative collapse caused by impurities observed in some situations.
A multi-machine study has been carried out to investigate the impact of a strongly bounded wave propagation domain on the Lower Hybrid current drive, a condition which occurs principally in high ...aspect ratio tokamaks. In this regime, the condition of kinetic resonance can be far above the upper boundary of the propagation domain, and may not be achieved by the usual toroidal upshift. Therefore no tail of fast electrons can be pulled out from the thermal bulk. Nevertheless, while tokamak plasmas are in principle almost transparent to the wave in this regime so-called “unbridgeable spectral gap”, full current drive is well achieved for the two tokamaks considered in this study, TRIAM-1M (Zushi et al. Nucl Fusion 43:1600, 2003) and WEST (Bourdelle et al. Nucl Fusion 55:063017, 2015), both characterized by a very large aspect ratio
R
/
a
>
5.5
. The case of the high aspect ratio tokamak HL-2A (Liu et al. Nucl Fusion 45:S239, 2005) for which the wave propagation domain has also an upper boundary, but close to the resonance condition, is considered by comparison. First principles modeling of the rf-driven current and the fast electron bremsstrahlung using the ALOHA/C3PO/LUKE/R5-X2 chain of codes shows unambiguously that the spectral gap must be already filled at the separatrix in order to reproduce quantitatively observations and some important parametric dependencies. This result is an important milestone in the physics understanding of the Lower Hybrid current drive, highlighting the existence of a powerful and likely universal alternative mechanism to bridge the spectral gap, that is not related to toroidal magnetic refraction. With an initially broad power spectrum, lobes with low parallel refractive indexes that carry most of the plasma current can be absorbed in almost single pass, restoring the full validity of the ray-tracing approximation for describing the propagation of the Lower Hybrid wave in cold plasmas.
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