Poloidal Limiter Research Articles

Conceptual design of poloidal horseshoe limiter layout for JA DEMO

In the development of the JA DEMO fusion reactor, the proposed poloidal horseshoe limiter aims to mitigate the surface heat load localized on the breeding blanket. The limiter protrudes from the first wall toward the plasma to shade the blanket from the charged particles, which account for the majority of the surface heat load. To concretize the limiter design, this study presents a limiter layout that implements a concept that allows for sufficient cooling and remote maintenance. The limiter shape was determined by evaluating the surface heat load using 3-D magnetic field line tracing. A fully poloidal limiter with a width of 300 mm, which corresponds to 2.4 % of the FW area, was selected. The limiter heating, consisting of surface heat load and nuclear heating, was evaluated to discuss the required cooling, and the concept of a limiter cassette and modules for cooling was proposed. For the limiter replacement method, the outboard limiter is designed to be extracted from the upper port with the mounting segment. The inboard limiter is then grasped and transported to reach the upper port after the outboard limiter extraction. Both the inboard and outboard limiter are expected to be replaced without affecting other in-vessel components in this study.

Development of high-performance long-pulse discharge in KSTAR

High-performance long-pulse plasma operation is essential for producing economically viable fusion energy in tokamak devices. To achieve such discharges in KSTAR, firstly, the rapid increase in the temperature of plasma-facing components was mitigated. The temperature increase of the poloidal limiter, especially, was associated with beam-driven fast ion orbit loss and the discrepancy of the equilibrium reconstructed with heated magnetic probes of signal drift. The fast ions lost to the poloidal limiter were reduced by optimizing the plasma shape and the composition of neutral beam injection (NBI). This nonlinear signal drift was successfully reduced by a new thermal shielding protector on the magnetic probes. Secondly, a lower loop voltage approach was implemented to reduce a poloidal flux consumption rate. A plasma current of 400 kA and a line-averaged electron density of ∼2.0 × 1019 m−3 were chosen by considering the L–H power threshold, fast ion orbit loss, and beam shine-through power loss for low loop voltage in KSTAR. In addition, the application of electron cyclotron heating also helped maintain the plasma with low loop voltage (∼25 mV) by enhancing the NBI-driven current and achieving a high poloidal beta (β P) state. KSTAR has achieved a long pulse (∼90 s) operation with the high performance of β P ⩽ 2.7, thermal energy confinement enhancement factor (H98y2) ∼ 1.1, and fraction of non-inductive current (f NI) ⩽ 0.96. Still, gradual degradation of the plasma performance has been observed over time in the discharges. In one of the long-pulse discharges, β P reduced by ∼18% over the time of ∼8τ R (current relaxation time, τ R ∼ 5 s) and ∼1067τ E,th (thermal energy confinement time, τ E,th ∼ 45 ms). The degradation may be closely associated with weak, yet growing, and persistent toroidal Alfvén eigenmodes and their effect on fast ion confinement.

Tailoring resonant magnetic perturbation to optimize fast-ion confinement during ELM control in KSTAR

3D resonant magnetic perturbation (RMP) is one promising way to control edge localized modes that can cause excessive material erosion of tokamak first walls. However, RMP can lead to undesired degradation of plasma confinement, including fast-particle losses, which can impact the performance and safety of the reactor. This work investigates the optimization of the poloidal spectrum of the 3D field to optimize fast ion confinement during edge localized mode (ELM) suppression. In the initial step, the validity of the modeling framework is tested against experimental data. Simulations successfully replicate an increase in poloidal limiter temperature with different poloidal spectra. Then, the simulation shows improvement of fast ion confinement with a reduction of core resonant response, while edge resonant magnetic fields are maintained above the threshold to sustain the ELM suppression. Reduction of the core resonant fields keeps the Kolmogorov–Arnold–Moser surface and reduces the fast particle losses due to the stochastic magnetic field lines. The results highlight the potential of edge localization of the resonant fields to enhance the performance of fusion reactors, but further investigation is needed to improve the validation of this approach.

LIBS characterization of the Frascati Tokamak Upgrade (FTU) limiter tiles after the experimental campaign with fusion plasma

In this work we present the results of the LIBS surface mapping and depth profiling of the Frascati Tokamak Upgrade (FTU) tiles of the poloidal limiter, made of a Molybdenum alloy (TZM) and removed from the reactor after its last experimental campaign with fusion plasma (late 2019). The measurements were performed by using a spectrometer array that covers the range 200–780 nm and a high-resolution spectrometer (0.1 Å at 500 nm) covering a spectral window of 10 nm. By applying a depth profiling procedure, the contamination of the tiles by B, Ca, Cr, Fe, K, Li, Mn, Mo, Ni, Ti was observed in their surface layers. Contamination with H, O, N was instead observed in deeper layers after some laser shots. B was attributed by the preliminary boronization procedure of the first wall (FW), carried out with deuterated diborane (D2B6), although D was not detected on the surface. Li was detected as deposited layer coming from previous experiments with liquid metals limiters in FTU. Ca and K are believed to come from the manipulation of the tiles after FTU decommissioning or as residual impurities of the deposited together with Li. H, O and N, from exposure of the tiles to the environmental gases, Mo and Ti were detected as main constituents of the tiles but also as eroded material during exposure to fusion plasma. Cr, Fe, Mn and Ni as eroded materials from the vacuum vessel (VV). In all cases, the surface concentration and thickness of the contaminated layer strongly varies across the TZM tile, showing a tendency of major impurity accumulations in areas more distant from the reactor's equatorial section.

Forward modelling of [formula omitted] camera view in ST40 informed by experimental data

Embedding diagnostics in future pilot plants will be a challenging task, because of space- and irradiation-related concerns. Relying on high-fidelity synthetic diagnostics would then be valuable.The 3D Monte-Carlo ray-tracing code CHERAB allows the development of numerous synthetic spectroscopic diagnostics. Focus of the present work is the introduction of new CHERAB models. The forward modelling of a synthetic Dα camera in ST40, the privately funded, high-field spherical tokamak, owned and operated by Tokamak Energy Ltd, and the comparison against experimental data is chosen as a testbed for quality assessment. Main output of the study then consists of estimates of the neutral particle densities throughout the chamber, of crucial relevance within edge plasma studies.Starting from simple analytical models, a 2D Dα source in the poloidal plane is generated. However, the centre column limited plasmas in ST40 display an intrinsically-3D Dα emission, mostly localised around the discrete poloidal limiters on the centre column, not captured by any axisymmetric source model. Hence, a novel methodology is introduced in CHERAB to approximate the 3D non-toroidally-symmetric pattern via a piece-wise emission distribution.Irrespective of the geometry of the emission and size of the tokamak, the pronounced non-homogeneity in the edge plasma emission requires sub-millimetric (∼ power fall-off length) spatial resolution to guarantee an accurate estimate of the peak emission. Minimising the associated burden via implementation of a non-uniform source sampling algorithm, which is a modification of the standard CHERAB uniform sampling, results in a >10-fold reduction of the computational cost.The significantly-shortened simulation time also makes the inclusion of more sophisticated models affordable. Of potential appeal in view of highly-detached divertors, the approximation of optically thin plasma is dropped, and photon-plasma interactions are accounted for. Brand-new CHERAB models able to take into account phenomena of photon absorption and scattering are so introduced.

Determination of limiter design and material composition of MT-II spherical tokamak

Abstract MT-II is a spherical tokamak with a major radius of 0.15 m and a minor radius of 0.09 m, currently under development at the Pakistan Tokamak Plasma Research Institute. It is designed with a higher elongation of 2.67. This paper presents the design and material analysis of the limiter configuration for the MT-II tokamak, which is being carried out in two phases. In the first phase, theoretical studies and calculations are performed to estimate the plasma edge temperature, density, particle velocity, input power, heat flux, heat load and surface temperature on the limiter tile. In the second phase, computational techniques are applied to analyses the material properties, the maximum/minimum surface temperature rise (∆T °C) at stable heat load and power deposition based on theoretical calculations that will help optimize the design parameters of the limiter. The type of material and the surface temperature of the limiter as well as the general design parameters of MT-II are included in the proposed poloidal limiter. The results suggest that crystalline vein graphite is a suitable candidate for the proposed poloidal limiter. A combination of mechanical and electrical feedthrough techniques are used to improve the performance of the limiter. The proposed limiter is able to meet the requirements of MT-II.

Characterizing W sources in the all-W wall, all-RF WEST tokamak environment * , **

Experimental data, together with interpretive modeling tools, are examined to study trends in the tungsten (W) source in the all-W environment of the WEST tokamak, both from the divertor and from the main chamber. In particular, a poloidal limiter protecting an ion cyclotron resonance heating (ICRH) antenna is used as proxy for main chamber sourcing. The key study is carried out by stepping up lower hybrid current drive (LHCD) power, as the only auxiliary power source. Limiter and divertor W sources exhibit a qualitatively similar proportionality to the total power crossing the separatrix, P SEP, although the main chamber source remains substantially lower than the divertor source, for the range of P SEP accessible in the experiments. Intepretive modeling of the limiter source is carried out with a particle-in-cell (PIC) sheath model coupled to a surface sputtering model. Oxygen is used as a proxy for all light impurity species allowing for characterization of the critical W erosion regions. To get a good quantitative match to the data, it is necessary to assume that the oxygen arrives at the surface mostly at high ionization stages (4+ and above). A separate simulation with SOLEDGE-EIRENE, constrained to measured upstream scrape-off-layer plasma profiles, gives oxygen fractional abundances that are compatible with the PIC simulation result. This is understood to arise from transport processes that dominate over recombination. Substituting the LHCD by ICRH, in an equivalent experiment, the local W source exhibits a 3× enhancement. This can be matched by the simulation, by assuming local RF electric field rectification, based on ∼100 eV peak-to-peak, near-antennna electric field. This work has highlighted the particular importance of understanding the ion charge state balance of light impurities as these are most likely the dominant sputtering species in fusion devices with high-Z walls.

First mirror erosion–deposition studies in JET using an ITER-like mirror test assembly

Mirror tests for ITER have been carried out in JET for over 15 years. During the third JET campaign with the ITER-like wall (2015–2016), comprising a total tokamak plasma exposure duration of 23.4 h and 1027 h of glow discharge cleaning, a new experiment was performed with a specially designed ITER-like test assembly housing six polycrystalline molybdenum mirror samples and featuring trapezoidal entrance apertures simulating the geometry of cut-outs in the diagnostic first wall of the ITER shielding blanket. The assembly was installed on the vacuum vessel wall at the outer midplane, set back radially behind the JET poloidal outer limiters such that the contact with thermal plasma should be largely avoided. The total and diffuse reflectivity of all mirrors was measured in the range 300–2500 nm before and after exposure. Post-exposure studies of mirror surface composition and of surfaces outside and inside the assembly were performed using microscopy, x-ray spectroscopy and ion beam analysis methods. The main results are: (i) no measured degradation of total reflectivity; (ii) diffuse reflectivity increased especially at short wavelengths (below 500 nm) from 1.1 to 2.7% and from 0.8%–1.3% above 1000 nm; (iii) mirrors were coated with a thin co-deposited layer (∼20–30 nm) containing carbon, oxygen and traces of nitrogen, beryllium and metals (Ni, Cr, Fe); (iv) no deuterium was detected; (v) surface composition of the mirror box inner walls was similar to that of the mirrors; (vi) ≲100 nm thick beryllium was the main component on external surfaces of the assembly. These results provide new input to ITER both for the modelling of FM erosion/deposition and for the consideration of requirements for mirror cleaning methods.

Tritium distribution analysis of Be limiter tiles from JET-ITER like wall campaigns using imaging plate technique and β-ray induced X-ray spectrometry

Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump plate (DP). The concentration of T on OPL was high at the central part. However, deuterium (D) and metallic impurities showed maximum concentration at the edges. This difference in distributions indicated different deposition and retention mechanisms between T and D. In contrast, deposition profiles of T concentrations on the castellated surfaces extended up to ∼ 5 mm into the gap, i.e. were similar to those of D and metallic impurities found by ion beam analysis.

Metallography and mechanical parameters of plasma-exposed plasma-facing materials and components

The performance of materials in a fusion environment is strongly dependent on the loading history, potentially leading to material modification due to thermal and particle (charged and neutral) loads. One clear indication of material modification by thermal loads is the modification of the microstructure by recrystallization, which is assessed by metallographic means as well as hardness measurements. Thereby, the recrystallization behaviour of a material strongly depends on its manufacturing route, the impurity level and on plasma impact, i.e. processes related to plasma–wall interactions. In this study, surface roughness and morphology as well as metallographic examinations and investigation by indentation techniques of components coming from different joint European torus (JET) experimental campaigns were performed for the first time on tritium-containing and therefore radioactive materials and components. This comprises on the one hand standard, marker and dedicated melt tungsten lamellae obtained from the divertor. On the other hand, beryllium components covered by nickel/beryllium marker coatings from the inner wall of JET, i.e. the inner wall guard limiter, wide poloidal limiter and dump plate were investigated. These have undergone various loading conditions and temperature excursions leading to surface modifications like material erosion, deposition and melting, and have also been assessed by scanning electron microscopy and energy-dispersive x-ray spectroscopy in view of the influence of the marker layers.

Thermal desorption of hydrogen isotopes from the JET Be plasma facing components

Understanding of fuel retention and release processes from materials for ITER (International Thermonuclear Experimental Reactor) is important from fundamental and technological aspects. Detailed information regarding fuel retention and release characteristics will allow global fuel inventories to be estimated in fusion devices as well as indicate the requirements for development detritiation methods and the re-use of tritium. Selected beryllium (Be) tiles were extracted from the JET vacuum vessel after each experimental campaign period with the ITER-like wall (ILW); so called ILW1 (2011–2012), ILW2 (2013–2014) and ILW3 (2015–2016). Desorption of hydrogen isotopes of samples taken from the inner wall guard limiter (IWGL), outer poloidal limiter (OPL) and dump plate (DP) tiles were analysed by means of thermal desorption spectrometry (TDS). The results presented compare data across ILW1, ILW2 and ILW3 and show the long term trends of fuel retention in Be limiter tiles. For all three campaigns the level of retention correlates with erosion and deposition that takes place during plasma operations Deuterium retention varies from 0.01–1 × 1018 atoms cm–2. Deuterium release takes place in several stages, related to different types of traps which can be within the co-deposit layer and/or below the surface in the Be bulk, with the main release stages around 700–760, 850–900 and 1020 K. The level of tritium in Be was found to be 104 times lower than deuterium for ILW campaigns 1–3.

Numerical study of neutral beam injection prompt loss at KSTAR poloidal limiters

We use a newly developed simulation tool to numerically study fast-ion loading on plasma-facing components (PFCs) at the Korea Superconducting Tokamak Advanced Research facility in the high-poloidal-β plasma operation regime. The new code can calculate neutral beam ionization and follow the guiding center orbit of ionized particles. The results of the simulation indicate that fast ions ionized in the high-field side drift out and strike the PFCs as they rotate poloidally. Momentum projection onto a phase space defined by canonical toroidal angular momentum and magnetic moment leads to a simple criterion to avoid fast-ion loading on poloidal limiters (PLs), which are PFCs that are toroidally localized on the low-field side. Control of fast-ion loss is examined by varying the plasma current and plasma boundary. A larger plasma current and inner shifting of the outer plasma boundary is predicted to substantially reduce fast-ion loading on the PLs. The proposed phase-space criterion is qualitatively consistent with the results of the simulation of fast-ion control.

Droplet transport from main HT-6M poloidal limiter

The HT-6M is a tokamak installed only with a main metallic poloidal limiter, which may not tolerate energy deposition due to a limited plasma-wetted area. This leads to the production of fresh metallic impurities, including macro-particles, in its main vessel. The macro-particles are transported and then ablated far away from where they were created. Understanding transport and ablation is beneficial to characterize impurity generation in the HT-6M, which will be recommissioned in Thailand in the next few years. A computational study simulated the motion of macro-particles in the HT-6M using the specifically written Dust and Droplet Tracking (DDT) code with the steady state HT-6M plasma profiles produced from the Transport Analyzing System for tokamaK (TASK) code. The results showed that a substantial amount of vapor containing impurities in the scrape-off layer (SOL) plasma was produced by the ablation of the liquid macro-particles emitted from the HT-6M poloidal limiter when the macro-particles were located near the limiter and nearby plasma-facing components (PFCs). Heavy ablation on the macro-particles transported directly towards the core plasma led to complete vaporization locally. The updated DDT code is discussed.

Active conditioning of ASDEX Upgrade tungsten plasma-facing components and discharge enhancement through boron and boron nitride particulate injection

The injection of boron (B) and boron nitride (BN) powders into ASDEX Upgrade H-mode discharges have demonstrated effective control of tungsten influx in low density/collisionality operational regimes, similar to conventional boronization methods. Sub-mm powder particles are gravitationally accelerated into the upper edge of a lower single null H-mode plasma with a boundary shape roughly conforming to the shape of the poloidal midplane limiters. Visible spectroscopy measurements at one of the outer limiter showed increases in both B and N signal levels, as well as elevated B levels in the divertor, and an increase in total radiated power by greater than a factor of two during BN injection. Globally the BN injection improved energy confinement by 10%–20%, associated with improvements in pedestal performance similar to gaseous N injection. Following conditioning discharges with B powder injection, three low gas-fueling discharges with magnetic perturbations for ELM suppression were successfully conducted. These first results suggest that the application of B containing powders can be used to both improve plasma performance in real-time, and to improve overall wall conditions for subsequent discharges.

Analysis of metallic impurity content by means of VUV and SXR diagnostics in hybrid discharges with hot-spots on the JET-ITER-like wall poloidal limiter

In preparation for the upcoming JET D–T campaign, great effort has been devoted during the 2015–2016 JET campaigns with the ITER-like wall (ILW) to the extension of the high performance H-mode phase in baseline and hybrid scenarios. Hybrid discharges were the only ones that have been stopped by the real-time vessel protection system due hot-spot formation on the outboard poloidal limiter. Generation of hot-spots was linked to the application of high neutral beams injection and ion cyclotron resonance heating (ICRH) power. In tokamaks with high-Z plasma components, the use of ICRH heating is also accompanied by an increased metallic impurity content. Simultaneous control of hot-spot temperature and the core impurity content was crucial due to the fact that the same plasma-wall interaction mechanism is responsible for both phenomena. Impurity data collected by SXR, EUV and VUV diagnostics were able to provide for the first time comprehensive information concerning tungsten and mid-Z impurities such as nickel, iron, and cooper. To determine absolute mid-Z impurity concentrations a new relative calibration technique, compatible with JET-ILW, has been developed based on cross-calibration with a calibrated spectrometer via the quasicontinuum of W in the 200–400 Å wavelength range. In hybrid discharges, it was found that local D2 gas injection, plasma current, separatrix density, and fast ion losses appeared to impact hot-spot temperature and core impurity levels. Analysis showed a reduced maximum hot-spot temperature and impurity concentration at higher gas rate. Changes in the plasma current had a strong impact on the plasma-wall interaction, both via modifications in the edge density and in the fast ion losses. At constant gas injection rate, both the hot-spot temperature and the core impurity content decreased with the separatrix density. The main mechanism responsible for the formation of the hot-spots was found to be linked to the fast ion losses, but RF sheath effects may also be playing a role in the high limiter temperatures observed in these experiments.

Overview of operation and experiments in the ADITYA-U tokamak

The first Indian tokamak, ADITYA, operated for over two decades with a circular poloidal limiter, has been upgraded to a tokamak named ADITYA Upgrade (ADITYA-U) to attain shaped-plasma operations with an open divertor in single and double-null configurations. Experimental research using ADITYA-U has made significant progress since the last FEC in 2016. After installation of a plasma facing component and standard tokamak diagnostics in ADITYA-U, the Phase-I plasma operations were initiated in December 2016 with a graphite toroidal belt limiter. Ohmically heated circular plasmas supported by filament pre-ionization with plasma parameters Ip ~ 80–95 kA, duration ~80–180 ms, with a maximum toroidal field ~1 T and chord averaged electron density ~2.5 × 1019 m−3, have been obtained. The runaway electron (RE) generation, transport and mitigation experiments, along with magneto hydrodynamic (MHD) activities and density enhancement with H2 gas puffing experiments were carried out in Phase-I, which was completed in March 2017. Preparation for the Phase-II operation in ADITYA-U includes calibration of magnetic diagnostics followed by commissioning of major diagnostics and installation of a baking system. After repeated cycles of baking the vacuum vessel up to ~135 °C, the Phase-II operations resumed in February 2018 and are continuing to achieve plasma parameters close to the design parameters of circular limiter plasmas, using real-time plasma position control. The plasma current has been raised to ~135 kA in Phase-II, with a maximum chord averaged electron density of ~4 × 1019 m−3. Hydrogen gas breakdown has been observed in more than 2000 discharges, including Phase-I and Phase-II operations, without a single failure. Several experiments have been carried out, including the control of REs with the fuelling of supersonic molecular beam injection as well as sonic H2 gas puffing during current flat-top, MHD mode studies using multiple periodic gas puffs, and radiative improved modes using neon gas puffs. The experimental results from Phase-I and Phase-II operations of the ADITYA-U tokamak are discussed in this paper.

Magnetic configuration effects on the Reynolds stress in the plasma edge

Breaking the poloidal symmetry of the magnetic shear induced tilt of turbulent structures, by either divertor X-point resistivity or limiter positions, can lead to a finite (residual) contribution to the flux-surface averaged radial-binormal Reynolds stress. This residual stress supports or works against the radial electric field at the plasma edge of a tokamak. The impact of divertor geometry on the poloidal pattern of the Reynolds stress is studied by flux-coordinate-independent fluid simulations. Clear modifications of the Reynolds stress are found due to the magnetic shear in the confined region. The impact of different poloidal limiter positions on the radial electric field and the Reynolds stress is studied by means of magnetic field aligned gyrofluid simulations. Only if the limiter is close to the outer midplane can its position have a substantial effect on the radial electric field.

Technical Issues With Steady-State Operation in KSTAR

Achieving and sustaining long pulse and ultimately steady-state operation of a fusion reactor in terms of physics and engineering parameters beyond transient period is essentially one of key requirements in present non Deuterium Tritium (DT) discharge tokamaks. Since the successful long pulse operation for 25 s at the plasma current of 0.5 MA in 2013, the duration of the H-mode pulse length has been extended to over 70 s, which corresponds to a few multiples of the current diffusion time. In addition to long pulse operation, the plasma performance is enhanced further to high poloidal beta discharge. This characterizes a fully noninductive discharge longer than 12 s, and also achieves both ion and electron temperatures of over 5 keV simultaneously for a line-integrated electron density of $5 {\times} 10^{19}$ /m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">3</sup> and the plasma current of 0.8 MA. Recently, KSTAR reported the long pulse operation beyond 1 min at the injected power of about 5 MW and the plasma current of ~0.5 MA. The normalized beta was about 1.5 and the total injected energy to the plasma reached to about 300 MJ. It is also shown that the bootstrap current fraction is about 40% so that sufficient conditions are met for 100-s operation in terms of physics requirements of fully noninductive state. The plasma-facing components overheating issues due to prompt ion loss in the low plasma current operation was overcome by enlarging the outer gap between the plasma and outer poloidal limiter. An advanced tokamak scenario via profile control would be achieved with the new 6-MW neutral beam injection-2, and it is expected that 100-s operation can be possible in a fully noninductive manner at KSTAR. However, a reliable and innovative steady-state heating and current drive source is strongly recommended for achieving the final goal of the KSTAR.

Accumulated 2-D neutron flux distribution during KSTAR operation

Since the successful first H-mode in 2010, duration of H-mode is extended to 1 min and the highest stored plasma energy up to 0.6 MJ is obtained by developing hybrid and internal transport barrier (ITB) mode as well as typical H-mode. Major heating system is the neutral beam (NB) of 100 keV with 5.5 MW and it is planned that the new additional 6 MW NBI is installed in 2018. So there is a lot of production of fast neutrons coming from via D(d,n)3He reaction and it is found that most of neutrons are 2.45 MeV neutrons coming from deuterium beam plasma interaction and the total neutron fluence during a campaign could be estimated to be about 1019. In particular, several nickel activation samples are installed inside/outside the vacuum vessel poloidally and toroidally and the total accumulated neutron flux is estimated from 811 keV gamma emission measurement after the campaign. The results show that the neutron activation both inside and outside the vacuum vessel has toroidally non-axisymmetric distribution and it has the highest neutron flux around poloidal limiter area. Considering the prompt ion loss at nominal KSTAR operation including the high beta-p operation, prompt loss ions from 100 keV neutral ion beams hitting on three poloidal limiters was not negligible and the non-uniform distribution is discussed in term of prompt ion loss to the poloidal limiter. In addition, this work reports on the quantitative analysis of poloidal distribution of the neutron emission based on accumulated flux measurements.

Runaway electron studies with hard x-ray and microwave diagnostics in the FT-2 lower hybrid current drive discharges

Studies of the super-thermal and runaway electron behavior in ohmic and lower hybrid current drive FT-2 tokamak plasmas have been carried out using information obtained from measurements of hard x-ray spectra and non-thermal microwave radiation intensity at the frequency of 10 GHz and in the range of (53 ÷ 78) GHz. A gamma-ray spectrometer based on a scintillation detector with a LaBr3(Ce) crystal was used, which provides measurements at counting rates up to 107 s−1. Reconstruction of the energy distribution of RE interacting with the poloidal limiter of the tokamak chamber was made with application of the DeGaSum code. Super-thermal electrons accelerated up to 2 MeV by the LH waves at the high-frequency pumping of the plasma with low density ~ 2 × 1013 cm−3 and then up to 7 MeV by vortex electric field have been found. Experimental analysis of the runaway electron beam generation and evolution of their energy distribution in the FT-2 plasmas is presented in the article and compared with the numerical calculation of the maximum energy gained by runaway electrons for given plasma parameters. In addition, possible mechanisms for limiting the maximum energy gained by the runaway electrons are also calculated and described for a FT-2 plasma discharge.