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The ergodic layer in the Large Helical Device (LHD) consists of stochastic magnetic fields exhibiting a three-dimensional structure that is intrinsically formed by helical coils. Spectroscopic diagnostics was employed in the extreme ultraviolet (EUV) and vacuum ultraviolet (VUV) wavelength ranges to investigate emission lines of carbon impurities in both hydrogen (H) and deuterium (D) plasmas, aiming to elucidate the impact of distinct bulk ions on impurity generation and transport in the edge plasmas of the LHD. The emission intensity of carbon CIII, CIV, CV, and CVI lines is significantly higher in the D plasma compared to the H plasma, indicating a greater sputtering rate of carbon materials in the D plasma, resulting in a higher quantity of carbon impurities originating from the divertor plates. A Doppler profile measurement of the second order of CIV line emission (1548.20 × 2 Å) was attempted using a 3 m normal-incidence VUV spectrometer in the edge plasma at a horizontally elongated plasma position. The flow velocity reaches its maximum value close to the outermost region of the ergodic layer, and the observed flow direction aligns with the friction force in the parallel momentum balance. The flow velocity increases with the electron density in H plasmas, suggesting that the friction force becomes more dominant in the force balance at higher density regimes. This leads to an increase in the impurity flow, which can contribute to the impurity screening. In contrast, the flow velocity in the D plasma is smaller than that in the H plasma. The difference in flow values between D and H plasmas, when the friction force term dominates in the momentum balance, could be attributed to the mass dependence of the thermal velocity of the bulk ions., source:Plasma 2023, 6(2), 308-321; https://doi.org/10.3390/plasma6020021, source:https://doi.org/10.3390/plasma6020021, identifier:0000-0002-1171-8603

Fusion product diagnostics based on four commercially available single-crystal chemical vapor deposition (s-CVD) diamond detectors are installed in the Large Helical Device (LHD) in order to understand energetic ion confinement. Characteristics of s-CVD diamonds were surveyed using alpha and D-T neutron sources. It is found that the energy resolutions of s-CVD diamonds for ∼ 5 MeV alpha particles and 14 MeV neutrons are 1%–3% and ∼ 1.7%, respectively. Moreover, the response of four s-CVD diamond detectors to alpha particles and D-T neutrons is almost identical. The installation positions of the diamond detectors in the vacuum vessel are searched for, based on the loss points of charged fusion products reckoned by Lorentz orbit calculations. Energy- and time-resolved measurement of fusion product flux will progress in further understanding of energetic ion confinement in LHD., source:K. Ogawa et al 2023 JINST 18 P01022, source:https://doi.org/10.1088/1748-0221/18/01/P01022, identifier:0000-0003-4555-1837

A radiative collapse predictor has been developed using a machine-learning model based on high-density plasma experiments in the Large Helical Device (LHD). Concurrently, the physical background of radiative collapse was discussed based on the distinct features extracted by a sparse modeling, which is one of the frameworks of data-driven science. Electron density, CIV and OV line emissions, and electron temperature at the plasma edge have been extracted as the key parameters of radiative collapse. Those parameters are relevant to the physical knowledge that the major cause of radiative collapse is the enhancement of radiative loss by light impurities in the plasma-edge region. Using these four parameters, the likelihood of occurrence of radiative collapse has been estimated. The behavior of plasma at the edge—in particular, the carbon impurities outside the last closed flux surface—has been evaluated using EMC3-EIRENE code for the phase with increasing likelihood, that is, the plasma is getting close to the collapse. It is shown that the radiation caused by the C3+ ion, which corresponds to the CIV emission, is enhanced in the region where electron temperature is around 10 eV., source:https://doi.org/10.1585/pfr.16.2402010

A radiative collapse predictor has been developed using a machine-learning model with high-density plasma experiments in the Large Helical Device (LHD). The model is based on the collapse likelihood, which is quantified by the parameters selected by the sparse modeling, including ne, CIV, OV, and Te,edge. The control system implementing this model has been constructed with a single-board computer to apply this predictor model to the LHD experiment. The controller calculates the collapse likelihood and regulates gas-puff fueling and boosts electron cyclotron resonance heating in real-time. In density ramp-up experiments with hydrogen plasma, high-density plasma has been maintained by the control system while avoiding radiative collapse. This result has shown that the predictor based on the collapse likelihood has the capability to predict a radiative collapse in real-time., source:http://doi.org/10.1585/pfr.17.2402042

Controlled particulate injections from the PPPL impurity powder dropper (IPD) into the Large Helical Device have demonstrated positive effects on the wall conditions on both an intra and inter-shot basis. Injections over a range of densities, input powers, pulse lengths, heating schemes, injection quantities and main ion species show conclusive evidence of improvement to plasma wall conditions. Successful injections are confirmed by both spectroscopic measurements as well as real-time visible camera signals. In 7 s long plasmas the responses include a reduction in wall recycling as well as a reduction in native impurity content as observed over the course of several discharges. For plasmas longer than 40 s, improvements to the recycling rate and increased impurity control are observed in real time as a consequence of the extended particulate injections. These experiments demonstrate the extended applicability of this solid particulate conditioning technique to the control and maintenance of the plasma wall conditions. In addition they are an important initial step in the development of the real-time boronization technique as a supplement to standard conditioning scenarios., source:R. Lunsford et al 2022 Nucl. Fusion 62 086021, source:http://doi.orig/10.1088/1741-4326/ac6ff5, identifier:0000-0003-3588-6801

In state-of-the-art stellarators, turbulence is a major cause of the degradation of plasma confinement. To maximize confinement, which eventually determines the amount of nuclear fusion reactions, turbulent transport needs to be reduced. Here we report the observation of a confinement regime in a stellarator plasma that is characterized by increased confinement and reduced turbulent fluctuations. The transition to this regime is driven by the injection of submillimetric boron powder grains into the plasma. With the line-averaged electron density being kept constant, we observe a substantial increase of stored energy and electron and ion temperatures. At the same time, the amplitude of the plasma turbulent fluctuations is halved. While lower frequency fluctuations are damped, higher frequency modes in the range between 100 and 200 kHz are excited. We have observed this regime for different heating schemes, namely with both electron and ion cyclotron resonant radio frequencies and neutral beams, for both directions of the magnetic field and both hydrogen and deuterium plasmas., source:Nespoli, F., Masuzaki, S., Tanaka, K. et al. Observation of a reduced-turbulence regime with boron powder injection in a stellarator. Nat. Phys. 18, 350–356 (2022). https://doi.org/10.1038/s41567-021-01460-4, source:https://doi.org/10.1038/s41567-021-01460-4, identifier:0000-0001-7644-751X

Three-dimensionally localization of radiative cooling due to nitrogen (N2) seeding for divertor detachment was detected experimentally. Since the localization along some magnetic field lines induces toroidal asymmetry of heat load reduction on divertor plates, it should be avoided for fusion reactors. The three-dimensionally localized structure was extracted using Principal Component Analysis (PCA) from two-dimensional radiation images measured with an InfraRed imaging Video Bolometer (IRVB). By applying PCA to 34 images each in N2 seeded plasmas with toroidally-asymmetric heat load reduction and in neon (Ne) seeded plasmas with toroidally-symmetric heat load reduction, a radiation feature in N2 seeded plasmas was found as one of the principal components (PC). The three-dimensional transport code EMC3-EIRENE indicated that the ionization in one of the divertor legs is enhanced in nitrogen seeding compared with Ne seeding due to the difference in the first ionization energy. The magnetic field lines from the divertor leg were along the extracted radiation structure and were terminated by the divertor where the heat load decreased due to the N2 seeding. These results indicate that three-dimensionally localized structure of radiative cooling was detected experimentally., source:https://doi.org/10.1016/j.nme.2022.101294, identifier:0000-0003-1586-1084

A closed helical divertor (CHD) has been designed for efficient particle control in the plasma periphery and for retaining neutral particles and impurity ions in the divertor region. The effect of impurity retention by divertor legs for the CHD configuration is investigated from the viewpoints of neutral impurity transport and force balance of impurity ions along magnetic field lines. A fully three-dimensional neutral particle transport simulation proves that the plasma on the divertor legs is effective for retaining neutral particles/impurities in the CHD region. A one-dimensional impurity ion transport analysis predicts that friction force by plasma flow from the main plasma sweep impurity ions toward the divertor plates even in high neutral density case in which a steep temperature gradient is formed. It shows that the CHD configuration is promising for enhancing LHD plasma performance by effective control of the neutral particles and the impurity ions in the plasma periphery., source:M. Shoji, S. Masuzaki, T. Morisaki, M. Tokitani, R. Sakamoto, H. Yamada, Effect of divertor legs on neutral particle and impurity retention for a closed helical divertor configuration in the Large Helical Device, Journal of Nuclear Materials, Volume 415, Issue 1, Supplement, 2011, Pages S557-S561, ISSN 0022-3115, https://doi.org/10.1016/j.jnucmat.2010.10.082., source:https://doi.org/10.1016/j.jnucmat.2010.10.082

An optimized closed divertor configuration for effective particle control in LHD is proposed from the viewpoints of the distribution of the strike points and neutral particle transport. Calculations of the distribution of the strike points indicate that 50% of the strike points locate in the inboard side of the torus in a standard magnetic configuration (Rax = 3.60 m). The ratio increases to 80% by installing target plates near lower/upper ports. A three-dimensional neutral particle transport simulation shows that installation of closed divertor components with the target plates raise the neutral pressure in the inboard side by more than one order of magnitude compared to that in the present open divertor case. The analysis of the neutral particle transport predicts that enhancement of the neutral pressure becomes moderate in outward shift configurations (Rax > 3.75 m)., source:M. Shoji, M. Kobayashi, S. Masuzaki, T. Watanabe, H. Yamada, A. Komori, Study of the effect of a closed divertor configuration on neutral particle control in the LHD plasma periphery, Journal of Nuclear Materials, Volumes 390–391, 2009, Pages 490-493, ISSN 0022-3115, https://doi.org/10.1016/j.jnucmat.2009.01.054., source:https://doi.org/10.1016/j.jnucmat.2009.01.054

The measured divertor heat flux profiles are compared to the EMC3-EIRENE simulations for two different times of an LHD discharge, corresponding to higher and lower edge temperatures. The relation between the three-dimensional magnetic field structure and the heat flux distributions on the divertor has been analysed. The modelled heat flux for the lower plasma temperature case has a better agreement with the experimental result obtained by the Langmuir probes, which shows a qualitative reproduction of the experimental profile shape. However, the heat flux distribution for the high plasma temperature case shows a different behaviour between the simulation results and the experimental measurements. The detailed analysis of the heat flux distribution for the higher temperature case which has a larger discrepancy has been performed, both quantitatively and qualitatively. The radiation of the eroded impurity from divertor target plates has a minor effect on the heat flux distribution. Non-uniform cross-field transport coefficients are used in the simulations and its impact on the heat flux distributions is discussed for the case of the high plasma temperature., source:Shuyu Dai et al 2017 Plasma Phys. Control. Fusion 59 085013, source:https://doi.org/10.1088/1361-6587/aa7599

FFHR-d1 is a conceptual design of the Large Helical Device-type fusion reactor. Several design optimizations for FFHR-d1 have been conducted under a multipath strategy. The structural design began from a radial build design of the components using a system code analysis and considering a shielding/breeding blanket design. The support structure must be sufficiently rigid to hold a constant geometric position of the radial build components. Additionally, large apertures are required for the maintenance work of in-vessel components. The shape of the support structure was carefully chosen and the analytically determined stress level was within the permissible limit. Thus, the basic design of the structural components of FFHR-d1 was established. To accelerate the design activity and achieve a consistent helical reactor system, the high-temperature superconducting joint-winding, liquid metal divertor, a supplemental helical coil addition, a novel divertor structure, and other challenging options are investigated alongside the basic design. This paper describes the structural design status of FFHR-d1 by focusing on the latest radial build design, construction scheme, and effectiveness of the novel divertor structure., source:https://doi.org/10.1585/pfr.11.2405061, identifier:0000-0003-4210-9859

A new closed helical divertor configuration for efficient particle control and reduction of the heat load on the divertor plates is proposed. The closed divertor configuration practically utilizes an ergodic layer and magnetic field line configuration on divertor legs in helical systems. For optimization of the design of the closed divertor, the distribution of the strike points is calculated in various magnetic configurations in the Large Helical Device (LHD). It suggests that the installation of the closed divertor components in the inboard side of the torus under an inward shift configuration (Rax=3.60m) is the best choice for achieving the above two purposes. This divertor configuration does not interfere with plasma heating and diagnostic systems installed in outer ports. The prospect of the closed divertor configuration to a helical fusion reactor is investigated using a three-dimensional neutral particle transport simulation code with a one-dimensional plasma fluid calculation on the divertor legs. The investigation shows efficient particle pumping from the in board side and reduction of the heat load due to the combined effect of the optimized closed divertor geometry, ergodized divertor legs, and low electron temperature in the ergodic layer. It indicates a promising closed divertor configuration for helical fusion reactors., source:To cite this article: Mamoru Shoji, Masahiro Kobayashi, Suguru Masuzaki, Akio Sagara, Hiroshi Yamada, Akio Komori & LHD Experimental Groups (2009) Design of a Closed Helical Divertor in LHD and the Prospect for Helical Fusion Reactors, Fusion Science and Technology, 56:2, 1001-1008, DOI: 10.13182/FST09-A9041, source:https://doi.org/10.13182/FST09-A9041

An additional deuterium irradiation was performed for long-term samples mounted on the first wall in the large helical device (LHD) to investigate the effect of mild baking (at 368 K) on fuel hydrogen removal from the modified surface. The efficiency of baking strongly depended on the surface conditions. In the erosion dominant area, almost all the retained deuterium was removed by baking, whereas only 17% and 13% of the retained deuterium were removed from the boron and carbon deposition areas, respectively. The results suggest that the erosion dominant area in the first wall can act as a pumping wall during the main discharge, and, in contrast, the fuel hydrogen retention can continuously increase in boron and carbon deposition areas owing to the low removal efficiency of fuel hydrogen given that the deposition continues during the plasma discharge., source:https://doi.org/10.1585/pfr.12.1302048, identifier:0000-0002-7412-3658

The first attempt to characterize the divertor footprint profile in the heliotron device LHD was done, by using a number of Langmuir probes and the multivariable analysis technique. In order to clarify the generation mechanism of the private-side peak on the footprint profile, which has not been reproduced in the modeling study, over 6000 time points were extracted by excluding time points with profile modifications due to already-known reasons. A characterization index ${r}_{2/1}$ was newly defined from the multivariable analysis result, and its dependences on upstream parameters were investigated. As a result, it was found that the footprint profile correlates with the pressure gradient at the edge inside the core region with a fixed beta, suggesting that change of the plasma pressure profile could modify the edge magnetic field structure even if the volume integral of the plasma pressure was constant., source:H Tanaka et al 2018 Plasma Phys. Control. Fusion 60 125001, source:https://doi.org/10.1088/1361-6587/aae2d7, identifier:0000-0002-8025-008X

Recently, a new limiter concept with liquid metal for a helical fusion reactor was proposed. This work present a modeling study to understand the influence of the limiter to the plasma and to find critical parameters for the liquid limiter concept. EMC3-EIRENE code is applied to a limiter configuration of a vertical column formed with plane surfaces at the inboard side of the Large Helical Device (LHD). The calculation results indicate that heat load on the helical divertor plates is significantly reduced when the limiter is inserted into the ergodic layer even without impurities. However, the influence of impurity radiation is too large in some cases. In order to achieve a large reduction of divertor flux and to avoid a large reduction of core electron temperature, parameter scans with limiter position and sputtering yield were performed. The results suggest that the limiter position is a critical parameter for acceptable degradation of the core and sufficient reduction of divertor flux., source:https://doi.org/10.1585/pfr.13.3403034, identifier:0000-0002-9117-5898

Arcing experiments were conducted in the linear plasma device Pilot-PSI, where a pulsed plasma was superimposed to a steady state plasma. The arcing was observed by a fast framing camera, and the sample was analyzed with a transmission electron microscope. Observations of glowing objects released from the sample in response to the arcing and destruction of the fuzzy layer at the edge of the arc trail without significant melting suggested that dust was formed and released from the surface in response to arcing., source:https://doi.org/10.1585/pfr.15.1205061, identifier:0000-0002-2023-9786

The Large Helical Device (LHD) started the deuterium plasma experiment on March 7, 2017. Approximately 6.4 GBq of tritium was produced in the first deuterium plasma experiment campaign and were utilized as tracers for the investigation of release behavior and the balance of tritium in the LHD vacuum vessel. To determine the tritium balance in LHD, the tritium release from the vacuum vessel was continually observed during the plasma experiment period and the vacuum vessel maintenance activities. The tritium exhaust rate was approximately 32.8% at the end of the plasma experiment. After the plasma experiment, the vacuum vessel was ventilated by room air for the maintenance activity and the tritium release from the in-vessel components was observed. The tritium release rate gradually decreased and became constant after four-month in spite of water vapor concentration. It is suggested that the tritium release mechanism from the vacuum vessel is a diffusion-limited process from the bulk. The tritium release amount during the maintenance activity for one year was approximately 5.0%. Considering the decrease of tritium decay for 1.5 years, tritium inventory in LHD was estimated to be approximately 3.66 GBq (57.2%) at the end of maintenance activity., source:https://doi.org/10.1585/pfr.15.1405062, identifier:0000-0001-9941-1958

We report the observation of arcing damage on the diagnostic shutter during the glow discharge wall conditioning in LHD. The diagnostic system has no experience of plasma discharge produced by electron or ion cyclotron resonance heating or neutral beam injection. The arc trails were observed on the aluminum surface but not on the stainless steel although both materials were exposed to the glow discharge with the same duration. The difference in work functions between two materials may be a cause to divide the conditions of arcing ignition., source:https://doi.org/10.1585/pfr.16.1202061, identifier:0000-0001-6090-5010