Internal Transport Barrier in Edge Plasma of Small Size Divertor Tokamak Using Neutral Beam Injection

We model the internal transport barrier “ITB” in edge plasma of small size divertor tokamak with B2SOLPS0.5.2D fluid transport code. The simulation results demonstrated the following: (1) we control the internal transport barrier by altering the edge particle transport through changes the edge toroidal rotation which agree with the result of Burrell et al. (Edge Pedestal control in quiescent H-mode discharges in DIII-D using co-plus counter-neutral beam injection, Nucl Fusion, 49, 085024 (9pp) in 2009). (2) The radial electric field has neoclassical nature near separatrix with discharge by co-injection NBI. (3) The toroidal plasma viscosity has strong influence on the toroidal velocity.     

The first results of deuterium retention on EAST with a full graphite wall via particle balance

This paper introduces the first results of deuterium retention on the Experimental Advanced Superconducting Tokamak (EAST) using particle balance.In the fall 2010 EAST experiments with a full graphite wall,the average deuterium retention fraction was about 19％ (including disruptive shots) and 38％ (not including disruptive shots).Fuel retention for the short-and long-pulse discharge was different.The H-mode discharges had a slightly lower fuel retention than the L-mode discharges.However,it was observed that disruptions introduced outgassing from the wall.Wall conditioning,such as lithium coating,increases retention.     

Heat Transport During H-Mode in the HL-2A Tokamak

Perturbative experiments on electron heat transport have been successfully con- ducted on the HL-2A tokamak. The pulse propagation of the electron temperature is induced by the supersonic molecular beam injection (SMBI), which has characteristics of good localization and deep deposition. A model based on the electron heat transport in cylindrical geometry has been applied to reconstruct the measured amplitude and phase profi les of the electron temperature perturbation. The results show that the heat transport is significantly reduced near the pedestal region of the H-mode plasma. In the \profi ness/resilience" region, similar heat diffusivities have been observed in L-mode and H-mode plasmas, which verifiesthe gradient-driven transport physics in tokamaks.     

Modelling of Small Size Divertor Tokamak Plasma Edge in the Higher Confinement Regime

A version of the B2SOLPS0.5.2D fluid transport code is the new version of B2SOLPS fluid transport code, which is suited technique to simulate the edge plasma of small size divertor tokamak in the H- regime. The results of simulation provide the following: (1) the radial electric field inside the transport barrier is consistent with the neoclassical nature of the radial electric field. (2) The absolute value of the radial electric field shear at inner side of internal transport barrier is small and consistent with the value of shear before the L–H transition, while the value of shear at barrier is significantly large. (3) As a result of strong radial electric field shear and strong barrier formation the diffusion coefficient reduced by factor ~3 with respect to L-mode while ion heat conductivity reduced by factor ~22 with respect to L-mode inside the barrier. (4) The toroidal (Parallel) flux is directed along co-current direction as L-mode but at inner side of barrier is significantly large in absolute value. (5) The radial profile of toroidal rotation in vicinity of transition layer is determined by the parameter δ (width of the transition layer) depending on the collisionality and anomalous diffusion coefficient.     

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak(EAST)L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer(SOL)widthλ_q and heat spreading 5,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I_p.Strong inverse scaling of the SOL width with I_p has been achieved for both L-mode and H-mode plasmas in the forms ofλ_(q,L-mode)=4.98×I_p~(-0.68)andλ_(q,H-mode)=1.86×I_p~(-1.08).Similar trends have also been demonstrated in the study of heat spreading with S_(L-mode)=1.95×I_p~(-0.542)and S_(H-mode)=0.756×I_p~(-0.872).In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor(CFETR).     

A new model of the L–H transition and H-mode power threshold

In order to understand the mechanism of the confinement bifurcation and H-mode power threshold in magnetically confined plasma,a new dynamical model of the L–H transition based on edge instability phase transition(EIPT) has been developed.With the typical plasma parameters of the EAST tokamak,the self-consistent turbulence growth rate is analyzed using the simplest case of pressure-driven ballooning-type instability,which indicates that the L–H transition can be caused by the stabilization of the edge instability through EIPT.The weak E?×?B flow shear in L-mode is able to increase the ion inertia of the electrostatic motion by increasing the radial wave number of the tilted turbulence structures,which play an important role for accelerating the trigger process of EIPT rather than directly to suppress the turbulent transport.With the acceleration mechanism of E?×?B flow shear,fast L–H and H–L transitions are demonstrated under the control of the input heating power.Due to the simplified scrape-offlayer boundary condition applied,the ratio between the heating powers at the H–L and L–H transition respectively differs from the ratio by Nusselt number.The results of the modeling reveal a scaling of the power threshold of the L–H transition,P_(L-H)?∝?n~(0.76) B~(0.8) for deuterium plasma.It is found finite Larmor radius induces an isotope effect of the H-mode power threshold.     

Observation of Central Toroidal Rotation Induced by ICRF on EAST

Core plasma rotation of both L-mode and H-mode discharges with ion cyclotron range of frequency(ICRF) minority heating(MH) scheme was measured with a tangential X-ray imaging crystal spectrometer on EAST(Experimental Advanced Superconducting Tokamak).Cocurrent central impurity toroidal rotation change was observed in ICRF-heated L-and H-mode plasmas.Rotation increment as high as 30 km/s was generated at ~1.7 MW ICRF power.Scaling results showed similar trend as the Rice scaling but with significant scattering,especially in L-mode plasmas.We varied the plasma current,toroidal field and magnetic configuration individually to study their effect on L-mode plasma rotation,while keeping the other major plasma parameters and heating unchanged during the scanning.It was found that larger plasma current could induce plasma rotation more efficiently.A scan of the toroidal magnetic field indicated that the largest rotation was obtained for on-axis ICRF heating.A comparison between lower-single-null(LSN)and double-null(DN) configurations showed that LSN discharges rendered a larger rotation change for the same power input and plasma parameters.     

Simulation of Small Size Divertor Tokamak Plasma Edge in the L-regime with ion Transport Barrier

Simulations of L-regimes of small size divertor tokamak plasma edge have been performed with the B2SOLPS5.0 2D fluid transport code for wide range parameters. A conclusion has been made that, radial electric field in the vicinity and inside separatrix is near to neoclassical electric field value. The poloidal E × B drifts and compensating parallel fluxes in the scrape off layer are large in the L-regime with ITB due to steeper gradients while the qualitative pattern of the flows is similar to that of the L-mode.     

Real-time fast ferrite ICRF tuning system on the Alcator C-Mod tokamak

A real-time ion cyclotron range of frequencies (ICRF) antenna matching system has been successfully implemented on Alcator C-Mod. This is a triple-stub tuning system working at 80 MHz, where one stub acts as a pre-matching stub and the other two stubs use fast ferrite tuners (FFTs) to accomplish fast tuning. It utilizes a digital controller for feedback control (200 μs per iteration) using real-time antenna loading measurements as inputs and the coil currents to the FFT as outputs. The system has achieved and maintained matching for a large range of plasma parameters, including L-mode, H-mode, and plasmas with edge localized modes. It has succeeded in delivering up to 1.85 MW net rf power into H-mode plasmas at maximum voltage of 37 kV on the unmatched side of the matching system.     

Effect of edge turbulent transport on scrape-off layer width on HL-2A tokamak

Effect of edge turbulent transport on scrape-off layer(SOL) width has been investigated in Ohmically heated L-mode plasma under limiter configurations on HL-2 A tokamak. It has been found that SOL width is doubled when plasma current decreases about 20%. With larger plasma current, E?×?B shear is stronger and has greater suppression effect on edge turbulent transport.SOL width is larger when power of relative density ?uctuation level in the edge region is larger.It is concluded that edge turbulent transport plays a significant role on SOL width. These experimental findings may provide a better understanding and controlling of power exhaust for present and future fusion devices.     

Study on Radial Position of Impurity Ions in Core and Edge Plasma of LHD Using Space-Resolved EUV Spectrometer

Radial profiles of impurity ions of carbon, neon and iron were measured for high-temperature plasmas in large helical device (LHD) using a space-resolved extreme ultraviolet (EUV) spectrometer in the wavelength range of 60 to 400?. The radial positions of the impurity ions obtained are compared with the local ionization energies, Ei of these impurity ions and the electron temperatures TeZ there. The impurity ions with 0.3?Ei?1.0 keV are always located in outer region of plasma, i.e., 0.7?ρ?1.0, and those with Ei?0.3keV are located in the ergodic layer, i.e., 1.0?ρ?1.1, with a sharp peak edge., where ρ is the normalized radial position. It is newly found that TeZ is approximately equal to Ei for the impurity ions with Ei?0.3keV, whereas roughly half the value of Ei for the impurity ions with 0.3?Ei?1.0keV. It is known that TeZ is considerably lower than Ei in the plasma edge and approaches to Ei in the plasma core. Therefore, this result seems to originate from the difference in the transverse transport between the plasma edge at ρ?1.0 and the ergodic layer at ρ?1.0. The transverse transport is studied with an impurity transport simulation code. The result revealed that the difference appearing in the impurity radial positions can be qualitatively explained by the different values of diffusion coefficient, e.g., D=0.2 and 1.0m2/s, which can be taken as a typical index of the transverse transport.     

Multi-scale interaction between tearing modes and micro-turbulence in the HL-2A plasmas

The influence of m/n=2/1(m and n are poloidal and toroidal mode numbers) tearing modes on plasma perpendicular flows and micro-fluctuations has been investigated in HL-2 A neutral beam injection heated L-mode plasmas. It is found that the local perpendicular rotation velocity and turbulence energy are modulated by the alternation between the island X-point and O-point of the naturally rotating tearing modes. Cross-correlation analysis indicates that the modulation of density fluctuations by the tearing mode is not only limited to the island region, but also occurs in the edge region near the last closed flux surface. The turbulence exhibits distinct spectral characteristics inside and outside the island region. In addition, it is observed that the particle flux near the strike point is also significantly impacted by the tearing modes. The experimental evidence reveals that there are strong core-edge interactions between the core tearing modes and the edge transport.     

Modelling of hydrogen isotope retention in the tungsten divertor of EAST during ELMy H-mode

In this work, we study hydrogen isotopes (HI) inventory inside tungsten plasma-facing materials during high confinement mode discharges with repetitive edge localized modes (ELMy H-mode) based on the operating parameters of the EAST device, since tungsten is considered as the primary plasma-facing material and the ELMy H-mode is an important operation regime for EAST and future devices. The upgraded Hydrogen Isotope Inventory Processes Code (HIIPC) is applied with the incident depth profile provided by SRIM-2013 to make the study. The code is first verified by comparison with experimental measurements. The effects of the incident ion energy and ion flux on the retention are then studied. Finally, using the parameters obtained from EAST diagnostics, the HI retention inside the W divertor during ELMy H-mode is studied, which indicates the retained HI can be increased dramatically mainly due to ion-induced trap sites by ELMs.     

Recent progress of NSTX lithium program and opportunities for magnetic fusion research

Lithium wall coating techniques have been experimentally explored on National Spherical Torus Experiment (NSTX) for the last six years. The lithium experimentation on NSTX started with a few milligrams of lithium injected into the plasma as pellets and it has evolved to a dual lithium evaporation system which can evaporate up to ～160 g of lithium onto the lower divertor plates between re-loadings. The unique feature of the NSTX lithium research program is that it can investigate the effects of lithium coated plasma-facing components in H-mode divertor plasmas. This lithium evaporation system has produced many intriguing and potentially important results. In 2010, the NSTX lithium program has focused on the effects of liquid lithium divertor (LLD) surfaces including the divertor heat load, deuterium pumping, impurity control, electron thermal confinement, H-mode pedestal physics, and enhanced plasma performance. To fill the LLD with lithium, 1300 g of lithium was evaporated into the NSTX vacuum vessel during the 2010 operations. The routine use of lithium in 2010 has significantly improved the plasma shot availability resulting in a record number of plasma shots in any given year. In this paper, as a follow-on paper from the 1st lithium symposium [1], we review the recent progress toward developing fundamental understanding of the NSTX lithium experimental observations as well as the opportunities and associated R&D required for use of lithium in future magnetic fusion facilities including ITER.     

Long Pulse H-Mode Scenarios Sustained by RF Heating on EAST

EAST has demonstrated its capability of long pulse operation using RF heating(LHCD and ICRF)in past experiments.One key issue to realize the long pulse H-mode experiments is to sustain the plasma current for steady state operation.Based on the calculations of the transport code ONETWO and its coupled RF code GENRAY,two scenarios have been proposed to achieve the 10 s H-mode plasma with Ip=400 kA,     

Design and test of voltage and current probes for EAST ICRF antenna impedance measurement

On the experimental advanced superconducting tokamak(EAST), a pair of voltage and current probes(V/I probes) is installed on the ion cyclotron radio frequency transmission lines to measure the antenna input impedance, and supplement the conventional measurement technique based on voltage probe arrays. The coupling coefficients of V/I probes are sensitive to their sizes and installing locations, thus they should be determined properly to match the measurement range of data acquisition card. The V/I probes are tested in a testing platform at low power with various artificial loads. The testing results show that the deviation of coupling resistance is small for loads RL??2.5 Ω, while the resistance deviations appear large for loads RL??1.5 Ω, which implies that the power loss cannot be neglected at high VSWR. As the factors that give rise to the deviation of coupling resistance calculation, the phase measurement error is the more significant factor leads to deleterious results rather than the amplitude measurement error. To exclude the possible ingredients that may lead to phase measurement error, the phase detector can be calibrated in steady L-mode scenario and then use the calibrated data for calculation under H-mode cases in EAST experiments.     

Effect of resonant magnetic perturbation on boundary plasma turbulence and transport on J-TEXT tokamak

The effect of resonant magnetic perturbation(RMP) on boundary turbulence and transport in J-TEXT plasma is experimentally investigated.Edge plasma fluctuations in discharges with and without the(m/n=3/1) RMP currents are diagnosed by using Langmuir probe arrays.It was found that fluctuations in the edge and scrape-off layer(SOL) regions decrease with the application of a 6 kA RMP.The broadband turbulence at the radial location of ρ～0.9 which has a characteristic frequency of 40-150 kHz was strongly suppressed when applying RMP,as was the radial turbulent particle flux and blob transport in the near-SOL region.These experimental findings make RMP a promising method of suppressing and controlling turbulence and particle transport in a plasma boundary.     

Improvement of Three-dimensional Monte Carlo Code PHITS for Heavy Ion Therapy

The general purpose particle and heavy ion transport code, PHITS, was modified for improved analysis of dose distribution in carbon therapy systems. We added two new functions into PHITS, one for an energy dispersion calculation and the other for transport in an AC magnetic field, which enabled 3-dimensional modelling of a carbon therapy system for the first time. With this code we calculated the dose distribution in a carbon therapy system, and these results showed good agreement with experimental data. This improved version of PHITS is a valuable tool for the design of carbon therapy aperture or for the estimation of the dose distribution in treatment planning.     

Calculation of radial burnup and nuclides atom density distributions in a VVER-1000 fuel rod using Monte Carlo method

A fundamental knowledge of fuel behavior in different situations is required for safe and economic nuclear power generation. Due to the importance of a fuel rod behavior modelling in high burnup, in this paper, the radial distribution of burnup, fission products, and actinides atom density and their variations by increasing burnup and other factors such as temperature, enrichment and power density are studied in a fuel pellet of a VVER-1000 reactor in an operational cycle using the MCNPX 2.7 Monte Carlo code. A benchmark including a Uranium-Gadolinium (UGD) fuel assembly is used for verification of the developed model in the MCNPX code for radial burnup calculation. A sensitivity study is carried out to investigate the effect of different parameters such as the number of particles per cycle, the number of geometrical radial nodes in the fuel pellet, the number of burnup steps and the selection of different fission-product contents (i.e. those isotopes that are used for particle transport) on the MCNPX model for speed and accuracy compromising. To calculate the radial temperature profiles and to analyze the effect of temperature on the radial burnup distribution and vice versa, the HEATING 7.2 code, which is a general-purpose conduction heat transfer program, and the MCNPX code are applied together. The results show the accuracy and capability of the proposed model in the MCNPX and HEATING codes for radial burnup calculation.     

Preliminary 2D code simulation of the quasi-snowflake divertor configuration in the FAST tokamak

The behaviour of the SOL in FAST for the new quasi snowflake configuration of the divertor is compared with that of the conventional single null case for three main scenarios: reference, advanced and extreme H-mode. The flexible, quick and versatile 2D code TECXY is used. The main physics processes occurring at the edge are carefully taken into account but the neutral dynamics is simplified. Even though the status close to detachment conditions cannot be detailed, new phenomena generated by the configuration are clearly highlighted. The heat load is strongly mitigated to a level that is much easier to tackle with the present technology. Mitigation is always stronger than expected from the magnetic topology only, especially in the high density regimes. This is attributed to the much longer time that the SOL particles spend in the quite cold region of the X point where mostly the connection length of the magnetic lines is strongly increased and hence to the much more intense interaction of the plasma particle with the background neutrals.