Functions of intrinsic magnetic field line structures on lithium ion transport in the LHD plasma periphery

Plasma discharge operation with lithium coating suggests that the lithium effectively control neutral particles in the plasma periphery, which can lead to improvement of plasma parameters. The effect of lithium coating on the large helical device (LHD) for a closed helical divertor configuration is discussed from viewpoints of neutral particle and impurity ion transport in the plasma periphery. It shows that the closed helical divertor configuration can enhance the neutral particle density in the divertor region, which is enough to achieve efficient particle control, and that it can effectively confine neutral lithium atoms near divertor plates. A one-dimensional impurity (lithium) ion transport analysis along magnetic field lines on divertor legs indicates that the friction force due to the plasma flow from the main plasma is dominant over the thermal force caused by the temperature gradient on the divertor legs, which prevents lithium ion contamination in the main plasma and excessive cooling of the plasma temperature in an ergodic layer. The analysis shows that the lithium coating is compatible with LHD plasma discharge operation for the closed helical divertor configuration.     

MHD Instabilities and Their Effects on Plasma Confinement in Large Helical Device Plasmas with Intense Neutral Beam Injection

MHD stability of the Large Helical Device (LHD) plasmas produced with intense neutral beam injection is experimentally studied. When the steep pressure gradient near the edge is produced through L-H transition or linear density ramp experiment, interchange-like MHD modes whose rational surface is located very close to the last closed flux surface are strongly excited in a certain discharge condition and affect the plasma transport appreciably. In NBI-heated plasmas produced at low toroidal field, various Alfven eigenmodes are often excited. Bursting toroidal Alfven egenmodes excited by the presence of energetic ions induce appreciable amount of energetic ion loss, but also trigger the formation of internal and edge transport barriers.     

Large Helical Device (LHD) program

The largest superconducting fusion machine, Large Helical Device (LHD), is now under construction in Japan and will begin operation in 1997. Design and construction of related R&D programs are now being carried out. The major radius of this machine is 3.9 m and the magnetic field on the plasma center is 3 T. The NbTi superconducting conductors are used in both helical coils and poloidal coils to produce this field. This will be upgraded in the second phase a using superfluid coil cooling technique. A negative ion source is being successfully developed for the NBI heating of LHD. This paper describes the present status and progress in its experimental planning and theoretical analysis on LHD, and the design and construction of LHD torus, heating, and diagnostics equipments.     

Divertor detachment operation in helium plasmas with ITER-like tungsten divertor in EAST

Detachment in helium (He) discharges has been achieved in the EAST superconducting tokamak equipped with an ITER-like tungsten divertor. This paper presents the experimental observations of divertor detachment achieved by increasing the plasma density in He discharges. During density ramp-up, the particle flux shows a clear rollover, while the electron temperature around the outer strike point is decreasing simultaneously. The divertor detachment also exhibits a significant difference from that observed in comparable deuterium (D) discharges. The density threshold of detachment in the He plasma is higher than that in the D plasma for the same heating power, and increases with the heating power. Moreover, detachment assisted with neon (Ne) seeding was also performed in L- and H-mode plasmas, pointing to the direction for reducing the density threshold of detachment in He operation. However, excessive Ne seeding causes confinement degradation during the divertor detachment phase. The precise feedback control of impurity seeding will be performed in EAST to improve the compatibility of core plasma performance with divertor detachment for future high heating power operations.     

Use of Ammonium Hydrogen Fluoride Solution to Clean Mixed Oxide Deposit on Inconel Surface

In order to carry out deuterium plasma experiments on the Large Helical Device (LHD), the National Institute for Fusion Science (NIFS) is planning to install a system for the recovery of tritium from exhaust gas and effluent liquid. As well as adopting proven conventional tritium recovery systems, NIFS is planning to apply the latest technologies such as proton-conducting ceramics and membrane-type dehumidifiers in an overall strategy to ensure minimal risk in the tritium recovery process. Application of these new technologies to the tritium recovery system for the LHD deuterium plasma experiment is evaluated quantitatively using recent experimental data.     

Development of lithium vapor injector for boundary control

A lithium (Li) vapor injector for boundary control has been developed. A diverter covered with lithium is expected to reduce particle recycling. Recycling reduction is considered to be one of the triggers for the L-H transition. In this paper, the method of lithium dispersion is investigated under the assumption that the experiment is carried out in the Large Helical Device in National Institute for fusion Science, Japan (LHD). A performance test is performed on a prototype of the vapor injector. The amount of injected lithium was approximately 1% of the value expected from the vapor pressure data, due to the generation of lithium oxide. It is also found that nozzle temperature is quite important to suppress the Li dispersion.     

中国环流器二号A(HL-2A)装置物理设计

文章是关于中国环流器二号A(HL-2A)装置物理设计的总结报告,包括以下几方面的内容:分析计算等离子体截面变形及由截面拉长引起的垂直不稳定性,提出对HL-2A极向磁场线圈电流和控制系统的要求;研究通过中性束注入加热(NBI)和低混杂波电流驱动(LHCD)实现等离子体剖面控制,模拟并设计HL-2A的高性能的运行模式;分析HL-2A先进约束位形(RS位形)下的磁流体力学不稳定性,为实现高性能模式稳态运行的等离子体控制指出方向;同时,利用数值模拟分析HL-2A偏滤器等离子体性能,为偏滤器的改进提供依据。     

Active radiative liquid lithium divertor concept

Developing a reactor compatible divertor has been identified as a particularly challenging technology problem for magnetic confinement fusion. Application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising Li results in NSTX and related modeling calculations motivated the radiative liquid lithium divertor (RLLD) concept [1]. In the RLLD, Li is evaporated from the liquid lithium (LL) coated divertor strike point surface due to the intense heat flux. The evaporated Li is readily ionized by the plasma due to its low ionization energy, and the poor Li particle confinement near the divertor plate enables ionized Li ions to radiate strongly, resulting in a significant reduction in the divertor heat flux. This radiative process has the desired effect of spreading the localized divertor heat load to the rest of the divertor chamber wall surfaces, facilitating divertor heat removal. The modeling results indicated that the Li radiation can be quite strong, so that only a small amount of Li (∼a few mol/s) is needed to significantly reduce the divertor peak heat flux for typical reactor parameters. In this paper, we examine an active version of the RLLD, which we term ARLLD, where LL is injected in the upstream region of divertor. We find that the ARLLD has similar effectiveness in reducing the divertor heat flux as the RLLD, again requiring only a few mol/s of LL to significantly reduce the divertor peak heat flux for a reactor. An advantage of the ARLLD is that one can inject LL proactively even in a feedback mode to insure the divertor peak heat flux remains below an acceptable level, providing the first line of defense against excessive divertor heat loads which could result in damage to divertor PFCs. Moreover, the low confinement property of the divertor (i.e., <1 ms for Li particle confinement time) makes the ARLLD response fast enough to mitigate the effects of possible transient events such as large ELMs.     

Integrated Operating Scenario to Achieve 100-Second,High Electron Temperature Discharge on EAST

Stationary long pulse plasma of high electron temperature was produced on EAST for the first time through an integrated control of plasma shape,divertor heat flux,particle exhaust,wall conditioning,impurity management,and the coupling of multiple heating and current drive power.A discharge with a lower single null divertor configuration was maintained for 103 s at a plasma current of 0.4 MA,q_(95)≈7.0,a peak electron temperature of 4.5 keV,and a central density n_e(0)~2.5×10~(19) m~(-3).The plasma current was nearly non-inductive(V_(loop) 0.05 V,poloidal beta ~0.9) driven by a combination of 0.6 MW lower hybrid wave at 2.45 GHz,1.4 MW lower hybrid wave at 4.6 GHz,0.5 MW electron cyclotron heating at 140 GHz,and 0.4 MW modulated neutral deuterium beam injected at 60 kV.This progress demonstrated strong synergy of electron cyclotron and lower hybrid electron heating,current drive,and energy confinement of stationary plasma on EAST.It further introduced an example of integrated "hybrid" operating scenario of interest to ITER and CFETR.     

Scale Effects on Magnet Systems of Heliotron-Type Reactors

For power plants heliotron-type reactors have attractive advantages, such as no current-disruptions, no current-drive, and wide space between helical coils for the maintenance of in-vessel components. However, one disadvantage is that a major radius has to be large enough to obtain large Q-value or to produce sufficient space for blankets. Although the larger radius is considered to increase the construction cost, the influence has not been understood clearly, yet. Scale effects on superconducting magnet systems have been estimated under the conditions of a constant energy confinement time and similar geometrical parameters. Since the necessary magnetic field with a larger radius becomes lower, the increase rate of the weight of the coil support to the major radius is less than the square root. The necessary major radius will be determined mainly by the blanket space. The appropriate major radius will be around 13 in for a reactor similar to the Large Helical Device (LHD).     

HL-2A Tokamak Edge Modeling with B2

The outer divertor plasma of HL-2A and its associated scrape-off plasma have been simulated using a two-dimensional multi-species fluid code of Braams with a simplified neutral gas model. HL-2A has a double-null closed divertor in separate divertor chambers above and below the nearly circular plasma tours. The computed numerical grid is developed according to an ideal magnetic surface. The calculation is involved only with pure hydrogen plasma. The emphasis has been focused on parametric studies involving variation of the assumptions made for the core plasma. The peak temperatures and the heat flux near the target are of the principal concern。     

The KSTAR plasma facing components for 2010 operation

The KSTAR plasma facing components (PFCs) consist of inboard limiter, poloidal limiter, divertor, passive stabilizer and neutral beam armor. The main function of the PFCs is to define boundary of operating plasma and to protect the vacuum vessel and in-vessel components such as diagnostic components, in vessel control coil and several kinds of launchers for heating and current drive systems. The divertor is designed to enhance effective particle control to keep high quality plasma with various flexibilities in the shaping control for wide range of operational regime. The passive stabilizer that is made of CuCrZr alloy is designed to passively control the vertical position and MHD instabilities during operation as well as outer boundary of the plasma. Since fabrication has been started for all of the plasma facing components from middle of 2009, the inboard limiter, the divertor, and the passive stabilizer were successfully installed in the vacuum vessel, in turn. Moreover, one set of neutral beam armor and three strings of poloidal limiters were also installed according to the heating system that newly comes in 2010. All the PFCs tiles were baked to 200 °C and the PFC system showed no vacuum leakage and other mechanical troubles. In this paper, key features, fabrication, results of assembly, and baking of the KSTAR PFCs are summarized in detail.     

Nuclear Characteristics of Gas-Suspended Boron Carbide Cooling Catalyzed D Fusion Reactor Blanket

Transient behaviors of plasma and in-vessel components have been investigated considering the divertor plasma state (detached/attached) transition. The SAFALY code consisting of a zero-dimensional plasma model and a one-dimensional heat transfer model of components has been modified to take account of the divertor plasma state transition on the basis of the updated divertor plasma physics. Several plasma events, i.e., over fueling, sudden auxiliary heating injection and Confinement improvement events which would be expected to result in overpower, were selected for the International Thermonuclear Experimental Reactor (ITER) and the transient behaviors were calculated on the assumption of a combined failure of plasma control and machine interlock in addition with a postulated plasma transient. The results show that plasma burning passively terminates due to sublimated impurity penetration from the carbon target surface, but there are possibilities of dry out of the coolant for the high heat flux in sudden attached state transition under the multifailure of plasma control. However, effects by the aggravating failure of the divertor are expected to be safely terminated by the confinement boundary, the vacuum vessel and its pressure suppression system.     

On the impurity problem in quasi steady-state toroidal plasma experiments and fusion reactors

A simple equation for estimating the impurity build-up in a plasma due to sputtering is discussed under various assumptions. It is shown that the D-T burning time in an experiment (or reactor) without a divertor or cold gas blanket is one particle confinement time at most. If the accumulation of impurities in the center of the plasma cannot be avoided, steady-state operation of a reactor even with a divertor will not be achievable. The effect of neutron sputtering is included in the discussion.     

HL-2M Divertor Geometry Exploration with SOLPS5.0

One of the critical issues to be solved for HL-2M is the power and particle exhaust. Divertor target plate geometry strongly influences the plasma profiles by controlling the neutral recycling pattern, which has in turn a strong effect on the symmetry and stability of the divertor plasma and finally on the whole edge region. The numerical simulation software SOLPS5.0 Pack- age is used to design and explore the divertor target plates for HL-2M. We choose two divertor geometries, and assess the heat flux on the target plates and first wall, then further discuss the di- vertor plasma parameters, and how private flux baffling affects both neutral recirculation pattern and pumping efficiency.     

Realization of minute-long steady-state H-mode discharges on EAST

In the 2016 EAST experimental campaign,a steady-state long-pulse H-mode discharge with an ITER-like tungsten divertor lasting longer than one minute has been obtained using only RF heating and current drive,through an integrated control of the wall conditioning,plasma configuration,divertor heat flux,particle exhaust,impurity management,and effective coupling of multiple RF heating and current drive sources at high injected power.The plasma current (Ip ～ 0.45 MA) was fully-noninductively driven (Vloop ＜ 0.0 V) by a combination of ～2.5 MW LHW,～0.4 MW ECH and ～0.8 MW ICRF.This result demonstrates the progress of physics and technology studies on EAST,and will benefit the physics basis for steady state operation of ITER and CFETR.     

Experimental study of density pump-out on EAST

In the experimental advanced superconducting tokamak,density pump-out phenomena were observed by using a multi-channel polarimeter-interferometer system under different heating schemes of ion cyclotron resonant heating,electron cyclotron resonance heating,and neutral beam injection.The density pump-out was also induced with application of resonant magnetic perturbation,accompanied with a degradation of particle confinement.For the comparison analysis in all heating schemes,the typical plasma parameters are plasma current 400 k A,toroidal field 2 T,and line average density 2?×?10~(19)m~(-3).The experimental results show that the degree of pump-out is concerned with electron density and heating power.Low density deuterium low confinement(L-mode) plasmas(3.5?×?10~(19)m~(-3)) show strong pump-out effects.The density pump-out correlated with a significant drop of particle confinement.     

Upgrade of an integrated Langmuir probe system on the closed divertor target plates in the HL-2A tokamak

A newly designed divertor Langmuir probe diagnostic system has been installed in a rare closed divertor of the HL-2A tokamak and steadily operated for the study of divertor physics involved edge-localized mode mitigation, detachment and redistribution of heat flux, etc. Two sets of probe arrays including 274 probe tips were placed at two ports (approximately 180° separated toroidally), and the spatial and temporal resolutions of this measurement system could reach 6 mm and 1 μs, respectively. A novel design of the ceramic isolation ring can ensure reliable electrical insulation property between the graphite tip and the copper substrate plate where plasma impurities and the dust are deposited into the gaps for a long experimental time. Meanwhile, the condition monitoring and mode conversion between single and triple probe of the probe system could be conveniently implemented via a remote-control station. The preliminary experimental result shows that the divertor Langmuir probe system is capable of measuring the high spatiotemporal parameters involved the plasma density, electron temperature, particle flux as well as heat flux during the ELMy H-mode discharges.     

Line identification of boron and nitrogen emissions in extreme-and vacuumultraviolet wavelength ranges in the impurity powder dropping experiments of the Large Helical Device and its application to spectroscopic diagnostics

An impurity powder dropper was installed in the 21 st campaign of the Large Helical Device experiment(Oct. 2019–Feb. 2020) under a collaboration between the National Institute for Fusion Science and the Princeton Plasma Physics Laboratory for the purposes of real-time wall conditioning and edge plasma control. In order to assess the effective injection of the impurity powders,spectroscopic diagnostics were applied to observe line emission from the injected impurity. Thus,extreme-ultraviolet(EUV) and vacuum-ultraviolet(VUV) emission spectra were analyzed to summarize observable impurity lines with B and BN powder injection. Emission lines released from B and N ions were identified in the EUV wavelength range of 5–300 ? measured using two grazing incidence flat-field EUV spectrometers and in the VUV wavelength range of 300–2400 ? measured using three normal incidence 20 cm VUV spectrometers. BI–BV and NIII–NVII emission lines were identified in the discharges with the B and BN powder injection, respectively. Useful B and N emission lines which have large intensities and are isolated from other lines were successfully identified as follows: BI(1825.89, 1826.40) ?(blended), BII 1362.46 ?, BIII(677.00, 677.14,677.16) ?(blended), BIV 60.31 ?, BV 48.59 ?, NIII(989.79, 991.51, 991.58) ?(blended), NIV765.15 ?, NV(209.27, 209.31) ?(blended), NVI 1896.80 ?, and NVII 24.78 ?. Applications of the line identifications to the advanced spectroscopic diagnostics were demonstrated, such as the vertical profile measurements for the BV and NVII lines using a space-resolved EUV spectrometer and the ion temperature measurement for the BII line using a normal incidence 3 m VUV spectrometer.     

Improvement of Plasma Performance Using Carbon Pellet Injection in Large Helical Device

A cylindrical carbon pellet with a size of 1.2L?1.2? to 1.8L?1.8?mm and a velocity of 100 to 300 m/s was injected into Large Helical Device (LHD) for an efficient fueling based on its deeper deposition instead of hydrogen gas puffing and ice pellet injection. Electron density increment of ?ne=1014cm-3 is successfully obtained by single carbon pellet injection without plasma collapse. Typical density and temperature of the ablation plasma of the carbon pellet, e.g., 6.5x1016cm-3 and 2.5eV for CII, are examined respectively by spectroscopic method. A confinement improvement up to 50% compared to ISS-95 stellarator scaling is clearly observed in a relatively low-density regime of ne=2 to 4?1013cm-3, and high ion temperature Ti(0) of about 6keV is also observed with an internal transport barrier at ne=1.2?1013cm-3. In particular, the improvement in the ion temperature largely exceeds that observed in hydrogen gas- puffed discharges, which typically ranges below 3 keV.