Helical Divertor Research Articles

Ultrahigh neutral pressures in the sub-divertor of the Large Helical Device

In the Large Helical Device (LHD) a low temperature mode (LTM) of the helical divertor was discovered. It combines particle detachment and very large sub-divertor pressures up to 1.4 Pa. During the LTM, the electron temperature in the divertor was in the range from 0.25 to 0.42 eV so that volume recombination occurred. This result is remarkable because in the stellarators LHD and Wendelstein 7-X only low sub-divertor pressures (0.03–0.3 Pa) were expected and measured up to now due the loss of pressure conservation along flux tubes by an enhanced cross-field transport. It demonstrates that the more complex, three-dimensional divertors of stellarators can achieve a similar performance with respect to particle exhaust and detachment like the geometrically simpler poloidal divertors in tokamaks even if the favorable effect of flux amplification is absent. The LTM of the helical divertor depends, however, on the magnetic configuration, i.e. on geometry. It was only observed in the inward shifted configuration with Rax=3.55 m, but not in the more frequently studied configuration with Rax=3.6 m. A nuclear fusion reactor based on the heliotron concept (DEMO) would benefit from the LTM by the very compact divertor configuration and the excellent performance.

Coordinated design of the cartridge-type blanket and the ceramic pebble divertor for the helical reactor FFHR-b3

The cartridge-type blanket for the helical reactor has been updated and named the CARDISTRY-B3. Together with this, the ceramic pebble ergodic limiter/divertor system named the REVOLVER-D3 has been adopted in the design. There was a difficulty in fabricating the blanket cartridges due to the first wall shape with three-dimensional surface. This difficulty has been removed by introducing the slit first wall, where the first wall surface is formed by alternately stacking solid metal plates and porous plates made of metal or ceramic. The liquid metal flowing inside the blanket cartridge oozes out through the porous plates and forms the surface flow covering the first wall. Solid target plates for the helical divertor has been omitted, because the divertor region is already covered and protected by the liquid metal surface flow. The liquid metal freely falls inside the blanket cartridges by the gravity force and then exhausted to the pool set on the lower port. The ceramic pebbles, which are dropped to the inboard-side ergodic layer to form the flowing limiter divertor, also flow into the liquid metal pool. The pebbles and liquid metal are individually elevated again by screw conveyers. Since the free surface of the liquid metal is exposed to the plasma, the vapor pressure of the working liquid metal should be low enough. Ternary or quadruple alloys, which include Li, Sn, Pb (or Bi), and Er, have been selected as the candidates of the functional liquid metal for the CARDISTRY-B3 that satisfies the requirements of low vapor pressure, low density, low melting point, and high tritium breeding ratio. In this study, the coordinated design of CARDISTRY-B3 and REVOLVER-D3 for the FFHR-b3 is presented. The FFHR-b3 is one of the design options that aims at demonstration of 100 MW net electricity with two times larger devise size than LHD.

Simulation of Impurity Transport and Deposition in the Closed Helical Divertor in the Large Helical Device

Simulation of Impurity Transport and Deposition in the Closed Helical Divertor in the Large Helical Device

Divertor heat load distribution measurements with infrared thermography in the LHD helical divertor

We evaluated the two dimensional (2D) distribution of the divertor heat flux in LHD. The Infrared (IR) thermography was performed to measure the surface temperature at a divertor plate. The 3D heat conduction equation was solved using the finite element method (FEM) by taking into account the practical divertor geometry including the heat sink behind the graphite tile and the cooling system. The FEM analysis successfully reconstructs heat load distribution from the measured temperature pattern. Significant difference was found between the temperature and the heat load patterns. The FEM analysis shows that the highest heat deposition is observed at the strike line with 5–10 MWm−2 of ∼ 10 mm width in NBI heated discharge. In addition to the strike line there is also found the lower heat deposition region of ∼ 1 MW m−2 with wide channel width ∼ 30 mm. The detailed heat transport analysis inside the divertor components shows that the heat transport process is different between the strike line and the other region due to the heat channel width and to the divertor component structure. The comparison between the heat flux obtained by the thermography and that by the Langmuir probes shows reasonable agreement, except that the peak value of the heat load is higher in the IR thermography than in the Langmuir probe. By relaxing the assumption that electron temperature, Te, equal to be ion temperature, Ti, in the sheath model for the Langmuir probe analysis, the agreement becomes better for the case with Ti > Te.

Boron transport simulation using the ERO2.0 code for real-time wall conditioning in the large helical device

Abstract The three-dimensional Monte-Carlo impurity transport and plasma surface interaction code ERO2.0 is applied to a full-torus model for the Large Helical Device (LHD). In order to find an optimum experimental condition for effective real-time wall conditioning (boronization) using an Impurity Powder Dropper (IPD), the toroidal and poloidal distribution of the boron flux density on the divertor components and the vacuum vessel are surveyed in various experimental conditions. The source profile of the neutral boron atoms originated from boron powders supplied from the IPD is calculated using the DUSTT code in background plasmas provided by the EMC3-EIRENE code. The simulations using ERO2.0 predict that higher plasma density operation is inappropriate for the effective wall conditioning because of the toroidally localized boron flux density in a closed helical divertor region. The ERO2.0 simulations have successfully revealed an optimum experimental condition for the wall conditioning with the toroidally uniform boron flux density in the closed helical divertor region.

Demonstration of suppression of dust generation and partial reduction of the hydrogen retention by tungsten coated graphite divertor tiles in LHD

Three sets of tungsten coated graphite divertor tiles (VPS-W tiles) were installed in the closed helical divertor of the Large Helical Device (LHD) instead of the graphite divertor tiles in the 2012FY plasma campaign for examining the plasma wall interaction (PWI) with divertor plasma. The first wall panels and divertor tiles of the LHD consist of stainless steels (SUS316L) and graphite, respectively. The carbon based mixed-material deposition layer including a very small amount of Fe element has been formed on not only the divertor tiles but also the first walls near the divertor tiles through the PWI processes. Such a mixed layer often causes undesirable influences for maintaining a long pulse discharge in LHD, e.g., changing the particle recycling properties and dust generation. After the single plasma campaign of the 2012FY, we confirmed drastic suppression of the mixed-material deposition layer on the first wall panels just under the VPS-W tiles. On the other hand, carbon based mixed-material deposition layer was formed on the VPS-W tile surface, where the amount of the hydrogen retention was estimated to be over ∼4 × 1021 H/m2. If we would want further suppression of the hydrogen retention on the VPS-W tiles, the plasma facing components should be replaced to a full metal wall to avoid formation of the carbon co-deposition layer.

Impurity transport simulation in the peripheral plasma in the large helical device with tungsten closed helical divertor

Long pulse plasma discharges in the Large Helical Device have been often interrupted by iron dust emission induced by electric arcing on the surface of the vacuum vessel. The iron ions in the peripheral plasma induced by the dust emission enhance physical sputtering and self-sputtering on the divertor plates, which can interrupt the long pulse plasma discharges by radiation collapse. The impurity transport simulation for a tungsten divertor configuration is performed using a three-dimensional edge plasma simulation code (EMC3-EIRENE) under the condition where the iron ions produced by the dust emission cause the sputtering on the tungsten divertor plates. The simulation shows that the sputtered tungsten significantly increases the radiation power by a factor of about 20 compared to that for the carbon divertor configuration in a high plasma temperature condition. The simulation reveals that plasma discharge operation with a high plasma density is desirable for the tungsten divertor. In this operational regime, the radiation power by tungsten ions is significantly reduced by the combined effect of the suppression of the sputtered tungsten and the control of the tungsten ion accumulation by the reduced thermal force in the peripheral plasma.

Progress in the Conceptual Design of the Helical Fusion Reactor FFHR-d1

The LHD-type helical fusion reactor FFHR has been studied to realize steady-state fusion power generation without a need for current drive and free from disruption. The conceptual design studies of FFHR are steadfastly progressing based on the presently ongoing experiments in the Large Helical Device (LHD). In order to enhance the attractive features of the base option of FFHR-d1A, which is similar to LHD, configuration optimization is being considered for FFHR-d1C. Slight modification of the helical coil trajectory gives an improved condition both for the plasma confinement and the MHD stability. In order to overcome the difficulty for construction and maintenance associated with the three-dimensional structure, innovative ideas are being explored for the superconducting magnet, divertor, and blanket. For the superconducting helical coils, the joint-winding method confirms a fast manufacturing process. The helical divertor is reexamined and practical feasibility is discussed. The maintenance method of the helical divertor and the helically-segmented breeder blanket is a serious issue and a plausible solution is proposed.

Maintainability of the helical reactor FFHR-c1 equipped with the liquid metal divertor and cartridge-type blankets

The maintenance scheme in the compact helical fusion reactor, FFHR-c1, equipped with the liquid metal divertor, REVOLVER-D (Reactor-oriented Effectively VOLumetric VERtical Divertor) and the cartridge-type molten salt blankets, CARDISTRY-B (CARtridges Divided and InSerTed RadiallY − Blanket), is investigated. The magnetic configuration of the compact variant FFHR-c1 is similar to that of the Large Helical Device (LHD), while the device size is 2.8 times enlarged from LHD and a strong magnetic field strength of ∼8 T at the plasma center is adopted. The maintenance of the REVOLVER-D is simpler than that of the helical divertor with a complicated structure as seen in the LHD. In the REVOLVER-D, showers of molten tin are injected into the ergodic layer at 10 inner ports. To circulate the molten tin, 10 sets of the shower system including a liquid metal pump, ducts, a showerhead, a pool, and a heat exchanger, are installed. These can be replaced with simple up/down motions. The CARDISTRY-B consists of 320 tritium breeding blanket cartridges, which are toroidally segmented every two degrees. These cartridges are maintained by using heavy manipulators with a simple 4DOF motion at a constant toroidal angle.

Establishment of a low recycling state with full density control by active pumping of the closed helical divertor at LHD

Superior control of particle recycling and hence full governance of plasma density has been established in the Large Helical Device (LHD) using largely enhanced active pumping of the closed helical divertor (CHD). In-vessel cryo-sorption pumping systems inside the CHD in five out of ten inner toroidal divertor sections have been developed and installed step by step in the LHD. The total effective pumping speed obtained was 67 ± 5 m3 s−1 in hydrogen, which is approximately seven times larger than previously obtained. As a result, a low recycling state was observed with CHD pumping for the first time in LHD featuring excellent density control even under intense pellet fueling conditions. A global particle confinement time () is used for comparison of operation with and without the CHD pumping. The was evaluated from the density decay after the fueling of hydrogen pellet injection or gas puffing in NBI plasmas. A reliably low base density before the fueling and short after the fueling were obtained during the CHD pumping, demonstrating for the first time full control of the particle balance with active pumping in the CHD.

Enhancement of helium exhaust by resonant magnetic perturbation fields at LHD and TEXTOR

The ability to exhaust helium as the fusion born plasma impurity is a critical requirement for burning plasmas. We demonstrate in this paper that resonant magnetic perturbation (RMP) fields can be used to actively manipulate helium exhaust characteristics. We present results from puff/pump studies at TEXTOR as example for a tokamak with a pumped limiter and from the Large Helical Device (LHD) with the closed helical divertor as example for a heliotron/stellarator device. For LHD, the effective helium confinement time is a factor of 7–8 higher in the low and high density regimes explored when compared to TEXTOR discharges. This is attributed to ion root impurity transport which is one particular impurity transport regime assessed experimentally at LHD and which facilitates helium penetration to the plasma core. However, when an edge magnetic island is induced by externally applied RMP fields, is decreased by up to and hence values closer to those of TEXTOR can be established. The combination of TEXTOR and LHD results suggest that a magnetic island induced by the RMP field in the plasma source region is an important ingredient for improving helium exhaust. The reduction in seen is caused by a combination of improved helium exhaust due to an enhanced coupling to the pumping systems, increased outward transport and a reduced fueling efficiency for the helium injected and recycling from the wall elements.

Simulation Analysis of Carbon Deposition Profile in the Closed Helical Divertor Configuration in the Large Helical Device

AbstractRecent long pulse plasma discharges in the Large Helical Device have been interrupted by radiation collapse induced by large amounts of dusts released from a closed divertor region. Traces of exfoliation of carbon‐rich mixed material deposition layers were found in the divertor region after the experimental campaigns, indicating that the thickly accumulated deposited layers were exfoliated and broken into dusts. Simulation of the density profiles of carbon deposition in the divertor region by EMC3‐EIRENE reasonably explains the observation of the exfoliation of the deposition layers. It shows that the positions of the exfoliated layers correspond to the area where the density of the carbon deposition is high. It also proves that change of the configuration of target plates in the closed divertor is effective for controlling the accumulation of the carbon deposition, and carbon atoms released from the new target plates does not influence the main plasma confinement region. (© 2015 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Development of New Concept In-Vessel Cryo-Sorption Pump for LHD Closed Helical Divertor

Development of New Concept In-Vessel Cryo-Sorption Pump for LHD Closed Helical Divertor

Development of impurity seeding and radiation enhancement in the helical divertor of LHD

Impurity seeding to reduce the divertor heat load was conducted in the large helical device (LHD) using neon (Ne) and krypton (Kr) puffing. Radiation enhancement and reduction of the divertor heat load were observed. In the LHD, the ratio between the total radiated power and the heating power, f rad = Prad/Pheating, is limited up to around 30% in hydrogen plasmas even for high density plasma just below the radiative collapse (ne, bar > 1 × 1020 m−3), where ne, bar is the line averaged density. With Ne seeding, the ratio could be raised to 52% at ne, bar ~ 1.3 × 1019 m−3, albeit with a slight reduction in confinement. f rad ~ 30% could be sustained for 3.4 s using multi-pulse Ne seeding at ne, bar ~ 4 × 1019 m−3. The localized supplemental radiation was observed along the helical divertor X-points (HDXs) which is similar to the estimated structure by the EMC3-EIRENE code. Kr seeding was also conducted at ne, bar ~ 3.1 × 1019 m−3. f rad ~ 25% was obtained without a significant change in stored energy. The radiation enhancement had a slower time constant. The supplemental radiation area of the Kr seeded plasma moved from the HDXs to the core plasma. Highly charged states of Kr ions are considered to be the dominant radiators from the plasma core region.

Improved signal to noise ratio and sensitivity of an infrared imaging video bolometer on large helical device by using an infrared periscope.

An Infrared imaging Video Bolometer (IRVB) diagnostic is currently being used in the Large Helical Device (LHD) for studying the localization of radiation structures near the magnetic island and helical divertor X-points during plasma detachment and for 3D tomography. This research demands high signal to noise ratio (SNR) and sensitivity to improve the temporal resolution for studying the evolution of radiation structures during plasma detachment and a wide IRVB field of view (FoV) for tomography. Introduction of an infrared periscope allows achievement of a higher SNR and higher sensitivity, which in turn, permits a twofold improvement in the temporal resolution of the diagnostic. Higher SNR along with wide FoV is achieved simultaneously by reducing the separation of the IRVB detector (metal foil) from the bolometer's aperture and the LHD plasma. Altering the distances to meet the aforesaid requirements results in an increased separation between the foil and the IR camera. This leads to a degradation of the diagnostic performance in terms of its sensitivity by 1.5-fold. Using an infrared periscope to image the IRVB foil results in a 7.5-fold increase in the number of IR camera pixels imaging the foil. This improves the IRVB sensitivity which depends on the square root of the number of IR camera pixels being averaged per bolometer channel. Despite the slower f-number (f/# = 1.35) and reduced transmission (τ0 = 89%, due to an increased number of lens elements) for the periscope, the diagnostic with an infrared periscope operational on LHD has improved in terms of sensitivity and SNR by a factor of 1.4 and 4.5, respectively, as compared to the original diagnostic without a periscope (i.e., IRVB foil being directly imaged by the IR camera through conventional optics). The bolometer's field of view has also increased by two times. The paper discusses these improvements in apt details.

A New Deduction Method of Heat Flux Evolution From Thermal Probe Data

AbstractMeasurement and control of divertor heat flux are a crucial issue for realization of fusion burning plasma. However, time‐dependent heat flux is often difficult to be estimated experimentally, because thermal diffusion process in thermal probe tips takes time longer than the discharge duration of present experiments. A new reconstruction method for time dependent heat flux from the measurement of the thermal probe tip temperature is proposed in this paper. The heat flux of the Large Helical Device divertor leg was analyzed by the proposed method as the function of the time for the first time. This indicates that heat flux measurement with thermal probes would become a monitoring tools even for the future fusion devices. (© 2014 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Modification of the helical divertor footprint induced by the net toroidal plasma current in the LHD heliotron

In this study, the modification of the helical divertor footprint induced by the net toroidal plasma current was investigated in the Large Helical Device (LHD). The divertor footprint shifted a few centimetres when the net toroidal plasma current, as measured by Rogowski coils, was induced via the neutral-beam injection. The shift direction was related to the plasma current direction, and the shift width showed almost linear dependence on the plasma current divided by the toroidal field strength. Magnetic field line tracing calculation considering the plasma current within a simple model revealed that the modification of the edge field line structure causes the footprint shift. The predicted and experimental shift widths agreed qualitatively.

Extension of operation regimes and investigation of three-dimensional currentless plasmas in the Large Helical Device

The progress of physical understanding as well as parameter improvement of net-current-free helical plasma is reported for the Large Helical Device since the last Fusion Energy Conference in Daejeon in 2010. The second low-energy neutral beam line was installed, and the central ion temperature has exceeded 7 keV, which was obtained by carbon pellet injection. Transport analysis of the high-Ti plasmas shows that the ion-thermal conductivity and viscosity decreased after the pellet injection although the improvement does not last long. The effort has been focused on the optimization of plasma edge conditions to extend the operation regime towards higher ion temperature and more stable high density and high beta. For this purpose a portion of the open helical divertors are being modified to the baffle-structured closed ones aimed at active control of the edge plasma. It is compared with the open case that the neutral pressure in the closed helical divertor increased by ten times as predicted by modelling. Studies of physics in a three-dimensional geometry are highlighted in the topics related to the response to a resonant magnetic perturbation at the plasma periphery such as edge-localized-mode mitigation and divertor detachment. Novel approaches of non-local and non-diffusive transport have also been advanced.

Initial experiments towards edge plasma control with a closed helical divertor in LHD

A baffle-structured closed helical divertor (CHD) is being constructed in the Large Helical Device to actively control the edge plasma, which consists of ten discrete modules installed on the inboard side of the torus. At this stage, two of the ten modules have been constructed. In the initial experiments, the performance of the CHD was experimentally investigated, comparing with numerical expectations. During the continuous gas puffing discharge, it was observed that the neutral pressure in the CHD was more than ten times higher than that in the open helical divertor, which agrees well with the numerical simulation. In the high-density regime, an indication of divertor detachment was observed in the CHD, which was caused by the high recycling and the high-density state in the CHD. With a Penning discharge diagnostics, the neutral particle behaviour in the mixture gas discharge was investigated, measuring hydrogen and helium pressures simultaneously. Similar compression properties were observed between two gases in the CHD, although larger recycling character was seen in helium.

Effect of n = 3 perturbation field amplitudes below the ELM triggering threshold on edge and SOL transport in NSTX

The pulsed application of n=3 magnetic perturbation fields with amplitudes below that which triggers ELMs results in distinct, transient responses observable on several edge and divertor diagnostics in NSTX. We refer to these responses as Sub-Threshold Edge Perturbations (STEPs). An analysis of edge measurements suggests that STEPs result in increased transport in the plasma edge and scrape-off layer, which leads to augmentation of the intrinsic strike point splitting due to error fields, i.e., an intensification of the helical divertor footprint flux pattern. These effects are much smaller in magnitude than those of triggered ELMs, and are observed for the duration of the field perturbation measured internal to the vacuum vessel. In addition, STEPs are correlated with changes to the MHD activity, along with transient reductions in the neutron production rate. Ideally the STEPs could be used to provide density control and prevent impurity accumulation, in the same manner that on-demand ELM triggering is used on NSTX, without the impulsive divertor fluxes and potential for damage to plasma facing components associated with ELMs.