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The fast ignition realization experiment (FIREX) project is progressing. The new short pulse laser system, LFEX laser, has been completely assembled and one of the four beamlets is now in operation. A fast-ignition experiment was performed using this single short pulse combined with the Gekko XII implosion laser. The energy of the GXII implosion laser was about 2 kJ and the pulse width was 1.5 ns. The energy of the LFEX laser was increased upto 800 J and two pulse durations 5 and 1.6 ps were compared. Targets were deuterated plastic shells with gold cones. It was found that the neutron yield was increased by a factor of 30 as a result of the fast electron-induced heating in LFEX 1.6 ps shot. The estimated coupling efficiency between the LFEX laser pulse and the compressed fuel was low (less than 5%). This may be due to pre-plasma formed by light arriving at the target before the main laser pulse. Further investigations and attempts to overcome these problems are now in progress. © 2011 Elsevier B.V.

An interesting high-density operational regime with an internal diffusion barrier (IDB) has been extended to the helical divertor configuration in the Large Helical Device. The IDB is characterized...

The Large Helical Device (LHD) is a heliotron-type device employing large-scale superconducting magnets to enable advanced studies of net-current-free plasmas. The major goal of the LHD experiment ...

Transport characteristics of the stochastic magnetic boundary of the Large Helical Device (LHD) are investigated, based on three-dimensional Monte-Carlo Braginskii-type fluid model code, EMC3, coup...

Heliotron reactors have several features suitable for a fusion power plant, such as no need for current drive, no plasma current disruptions, suitability for steady-state operation, and a wide spac...

Deuterium ion irradiations with an ion with energy of 1.7 keV were conducted for boron–titanium (B–Ti) film prepared by electron beam evaporation and hot pressed titanium-boride, TiB 2 . The amount of retained deuterium was measured for these materials using a technique of thermal desorption spectroscopy. The amount of deuterium retained in TiB 2 was comparable with that in B–Ti. Desorption peaks of deuterium in B–Ti were 470 K and 620 K, corresponding to a desorption in the low temperature regime observed in boron (B) and a desorption in titanium (Ti), respectively. The desorption peaks in TiB 2 were 620 K and 750 K, which correspond to the desorption in Ti and that in the high temperature regime in B, respectively. The desorption temperature in B–Ti was approximately 100 K lower than that in TiB 2 . This difference is discussed based upon chemical bindings and amorphous/crystal structures of B–Ti and TiB 2 . Irradiation of helium ion with energy of 5 keV was conducted for B–Ti after the deuterium ion irradiation. The amount of retained deuterium decreased and the desorption temperature shifted to the lower temperature regime, as the helium ion fluence increased. The shift to the low temperature regime is due to the enhancement of amorphous structure of B in B–Ti.

Configuration optimization is examined for the heliotron fusion energy reactor FFHR in order to find sufficient clearances between the ergodic region outside the nested magnetic surfaces and blankets at the inboard side of the torus. The standard configuration of FFHR, which is similar to that of LHD, has a relatively large major radius of the helical coils in order to satisfy this requirement. It has been found, as an alternative design, that equivalent clearances are obtained with a shorter major radius both by employing a lower helical pitch parameter and splitting the helical coils in the poloidal cross-section at the outboard side. Splitting the helical coils also provides another configuration that ensures magnetic well formation in the fairly large nested magnetic surfaces with outward shifted configurations. Optimization is being carried out for these configurations by adjusting the pitch modulation parameter to improve the particle confinement (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

Fuel layering of a cryogenic target with a conical laser guide such as the FIREX target is complicated because of its non-spherical symmetry appearance. To simplify the layering, a foam layer is pl...

This article reviews 10 years of engineering and physics achievements by the Large Helical Device (LHD) project with emphasis on the latest results. The LHD is the largest magnetic confinement device among diversified helical systems and employs the world's largest superconducting coils. The cryogenic system has been operated for 50,000 h in total without any serious trouble and routinely provides a confining magnetic field up to 2.96 T in steady state. The heating capability to date is 23 MW of NBI, 2.9 MW of ICRF and 2.1 MW of ECH. Negative-ion-based ion sources with the accelerating voltage of 180 keV are used for a tangential NBI with the power of 16 MW. The ICRF system has full steady-state operational capability with 1.6 MW. In these 10 years, operational experience as well as a physics database have been accumulated and the advantages of stable and steady-state features have been demonstrated by the combination of advanced engineering and the intrinsic physical advantage of helical systems in LHD. Highlighted physical achievements are high beta (5% at the magnetic field of 0.425 T), high density (1.1 × 1021 m−3 at the central temperature of 0.4 keV), high ion temperature (Ti of 5.2 keV at 1.5 × 1019 m−3), and steady-state operation (3200 s with 490 kW). These physical parameters have elucidated the potential of net-current free helical plasmas for an attractive fusion reactor. It also should be pointed out that a major part of these engineering and physics achievements is complementary to the tokamak approach and even contributes directly to ITER.

Superdense plasma with a highly peaked electron density profile was obtained in the reduced recycling discharge in the Large Helical Device (LHD). During the discharge, a core region with a density as high as ∼5 × 1020 m−3 and a temperature of ∼0.8 keV is maintained by an Internal Diffusion Barrier (IDB) with a steep density gradient. In spite of such a high density at the core region, no serious impurity accumulation has been observed. According to the numerical calculation with the EMC3–EIRENE code, downstream parallel flow in the ergodic region plays an important role to screen the influx of impurities with its strong friction force. The divertor flux distribution and global recycling properties do not change so much as against the change in the dense core plasma.

Progress of design studies on a heliotron reactor is reported. Recent experimental achievements in large helical device (LHD) are building up the physical basement. Integrated designs on LHD-type energy reactors have improved the base design of force free helical reactor and revealed economical feasibility. Large helical device is exploring a new horizon of net current free helical plasmas. Inherent characteristics demonstrate reactor relevant performances individually; steady state (1 h), high beta (5%), high density (1 × 10 21 m −3 ), etc. In particular, discovery of internal diffusion barrier provides a novel attractive scenario of super high density operation. It greatly facilitates physical requirements as well as mitigates engineering demands. New methods have been proposed to access and operate ignited plasmas; a long rise-up over 300 s to reduce the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling to handle the thermally unstable plasma at high-density operations.

Large helical device, the largest superconducting stellarator, has been operated for the research of fusion plasma since 1998. The toroidal field of almost 3 T is produced by a pair of pool-cooled helical coils, in the innermost layers of which a normal-zone had been induced several times at the bottom of the coil at higher currents than 11.0 kA. Since the field is not the highest there, the local cooling conditions are probably deteriorated by bubbles gathered by buoyancy. In order to improve the cryogenic stability by subcooling, an additional cooler with two-stage cold compressors was installed at the inlet of the coil in 2006. The inlet and outlet temperatures of the coils were successfully lowered to 3.2 K and 3.8 K, respectively, with a mass flow of 50 g/s. In spite of a half charging rate to reduce AC losses, a normal-zone was induced near the top of the coil at 11.45 kA. It propagated to one side and stopped near the inner equator, where the field is the highest. In comparison with the stability tests with a model coil, the local temperatures of the innermost layers near the top is considered to have been raised up to almost the saturated temperature of 4.4 K by charging. The excitation method was revised to waiting cool-down at 11.0 kA, and the excitations up to 11.5 kA have been attained.

This paper presents the control, data acquisition, data analysis and remote participation facilities of the Large Helical Device (LHD), which is designed to confine the plasma in steady state. In LHD the plasma duration exceeds 3000 s by controlling the plasma position, the density and the ICRF heating. The “LABCOM” data acquisition system takes both the short-pulse and the steady-state data. A two-layer Mass Storage System with RAIDs and Blu-ray Disk jukeboxes in a storage area network has been developed to increase capacity of storage. The steady-state data can be monitored with a Web browser in real time. A high-level data analysis system with Web interfaces is being developed in order to provide easier usage of LHD data and large FORTRAN codes in a supercomputer. A virtual laboratory system for the Japanese fusion community has been developed with Multi-protocol Label Switching Virtual Private Network Technology. Collaborators at remote sites can join the LHD experiment or use the NIFS supercomputer system as if they were working in the LHD control room.

The impurity transport properties in the ergodic layer of LHD are analyzed by the 3D edge transport code as well as 1D simple model, which is used to illustrate the essential transport terms in the analysis. It is found that as the plasma density increases the edge surface layers, very edge region of the ergodic layer, enters friction dominant regime, resulting in impurity retention. It is considered that the cause for the retention is both temperature drop and the flow acceleration in the edge surface layers. The edge surface layers can provide effective retention of impurities coming from divertor as well as the first wall, because of the geometrical advantage of the edge region of LHD. (© 2008 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

"A high density regime with an internal diffusion barrier (IDB) has been extended to the helical divertor (HD) configuration in the Large Helical Device (LHD). Avoidance of the local enhancement of neutral pressure is necessary to enable IDB formation, which is consistent with earlier works by using the Local Island Divertor (LID) with efficient active pumping. The central pressure reached 1.3 times atmospheric pressure, where ne(0) = 6 × 1020 m?3 and Te(0) = 660 eV. The plasmas with an IDB are located in the plateau collisionality regime. The significant impurity effect has not been observed throughout the discharges in spite of the existence of a negative radial electric field. A central pressure limiting event is observed in the plasmas with an IDB using the HD. During this event which is referred to as the core density collapse (CDC), particles are flushed out from the core on the time scale of a few hundreds of microseconds. The suppression of the Shafranov shift by vertical elongation (κ) is effective to mitigate CDC. At κ = 1.2, the central β value is increased up to 6.6% at 1 T."

Minimization of the external heating power to access self-ignition is advantageous to increase the reactor design flexibility and to reduce the capital and operating costs of the plasma heating device in a helical reactor. In this work we have discovered that a larger density limit leads to a smaller value of the required confinement enhancement factor, a lower density limit margin reduces the external heating power and over 300 s of the fusion power rise-up time makes it possible to reach a minimized heating power. While the fusion power rise-up time in a tokamak is limited by the OH transformer flux or the current drive capability, any fusion power rise-up time can be employed in a helical reactor for reducing the thermal stresses of the blanket and shields, because the confinement field is generated by the external helical coils.

Material probes were installed at the inner walls along the toroidal and poloidal directions in LHD from the sixth to eighth experimental campaign in LHD. The boronization was three times conducted in each campaign. Gas species, H2, He, Ar and Ne, were employed in both main and glow discharges during the campaign. After each campaign, gas retention and boron deposition behavior was examined to clarify the change of wall surfaces and plasma wall interactions. The amounts of retained gases were observed to be large in the vicinity of anodes used for the glow discharges. The amount of retained helium was observed to be relatively large, only one or two orders of magnitude smaller than that of hydrogen. The amounts of retained argon and neon were about two orders of magnitude smaller than that of helium. The thickness of deposited boron was large in the vicinity of both anodes and gas inlets. The toroidal uniformity of the boron deposition increased to about 70% in the eighth campaign by adjusting the gas flow rate. Understanding for the plasma wall interactions in LHD was deepened by the present study.

The force-free helical-type fusion reactor, FFHR, is proposed on the basis of the engineering achievements and confinement properties of the experimental fusion device of LHD. The outputs of the thermal power and electric power are optimized to 3 and 1 GW, respectively. Total weight of the superconducting (SC) coils and their supporting structures of the FFHR are estimated to be 18,000 t. An equivalent refrigeration capacity of 98 kW is necessary for coping with different plant loads. Mass-flow rate of the main circulation compressors is 9.5 kg/s and their power consumption is 29 MW. The FFHR is used for the co-generation system of electricity and hydrogen. The pressurized hydrogen of 100 t per day can be produced, when the stem electrolyzer of 150 MW class is applied. Electric power consumption of the hydrogen liquefaction with 100 t per day is estimated to be 26 MW.

The divertor performance of LHD is studied for the two configurations, LID and HD. It is shown that the both divertor configurations play important roles for obtaining high performance plasmas in LHD : the large pumping capability of the LID to keep the low edge density in the IDB-SDC plasma, the large wetted area and the flexibility of strike point sweep of HD to reduce the power load on the divertor plates in long pulse operations. The possible effect of the ergodic layer on impurity retention in divertor is discussed by using the 3D edge transport modelling. It is found that the drag force exerted by the plasma flow can dominate over the thermal force, providing the impurity retention effect. The further changes needed to improve the current divertor configurations are discussed. New divertor designs for the future upgrade of LHD and for a LHD-type reactor are presented.

Achieving steady-state plasma operation at high plasma temperatures is one of the important goals of worldwide magnetic fusion research. High temperatures of approximately 1–2 keV, and steady-state plasma sustainment operations have been reported. Recently the steady-state operation regime was greatly extended in the Large Helical Device (LHD). A high-temperature plasma was created and maintained for 54 min with 1.6 GJ in the 2005FY experimental programme. The three-dimensional heat-deposition profile of the LHD helical divertor was modified, and during long-pulse discharges it effectively dispersed the heat load using a magnetic axis swing technique developed at the LHD. A sweep of only 3 cm in the major radius of the magnetic axis position (less than 1% of the major radius of the LHD) was enough to disperse the divertor heat load. The steady-state plasma was heated and sustained mainly by hydrogen minority ion heating using ion cyclotron range of frequencies and partially by electron cyclotron of fundamental resonance frequency. By accumulating the small flux of charge-exchanged neutral particles during the long-pulse operation, a high energy ion tail which extended up to 1.6 MeV was observed. This is the first experimental evidence of high energetic ion confinement of MeV range in helical devices. The long-pulse operations lasted until a sudden increase in radiation loss occurred, presumably because of metal wall flakes dropping into the plasma. The sustained line-averaged electron density and temperature were approximately 0.8 × 1019 m−3and 2 keV, respectively, at a 1.3 GJ discharge (#53776) and 0.4 × 1019 m−3and 1 keV at a 1.6 GJ discharge (#66053). The average input power was 680 kW and 490 kW, and the plasma duration was 32 min and 54 min, respectively. These successful long operations show that the heliotron configuration has a high potential as a steady-state fusion reactor.

The transport properties of the edge region in LHD have been investigated in order to clarify the divertor/SOL function of the heliotron type device. The momentum loss, mainly through the friction of counter-flows induced by the ergodic field lines, breaks the pressure conservation along flux tubes. This results in no high recycling regime even at high density operation, ¯ n » 7 £ 10 19 m i3 . The momentum loss is found to be higher than the case of W7-AS. This is because of the higher ratio of perpendicular and parallel transport scale length, » 10 i4 , and of the momentum loss due to the friction of counter-flows, which is more eective in the ergodic layer than the island divertor. In the heliotron configuration, a large temperature drop from LCFS to divertor, an order of magnitude, is easily realized due to the long connection length in the ergodic layer. This is certainly a favourable feature for future reactor in terms of reduction of heat power load at the divertor.

The fuel layering process of a cryogenic target for the Fast Ignition Realization EXperiment (FIREX) project has been studied. A foam shell method is proposed as a fuel layering technique for this target design. The difficulty of the fuel layering comes from the aspherical target symmetry. In the case of the foam shell method, liquid fuel is directly infiltrated into a foam shell though a fuel feeder and is soaked up into the foam layer by capillarity. The fuel is then solidified and an ideal cryogenic target is formed. To date, the cryogenic system for the demonstration of the fuel layering was fabricated and subsequently modified to improve its cool-down performance. A dummy foam target has been utilized to study the fuel layering process using H 2 instead of D 2 and DT fuels. Liquid H 2 is supplied into the shell through a feeder with a 20 μm inner tip diameter. The solid H 2 quantity remaining in the shell was controlled by regulating both H 2 pressure and target temperature during solidification.

The Large Helical Device (LHD) is not only the largest stellarator for the research of fusion plasma near a reactor region but also the largest superconducting system. Availability higher than 98% has been achieved in the long-term continuous operation both in the cryogenic system and in the power supply system. It is due to the robustness of the systems and efforts of maintenance and operation. One big problem is the shortage of cryogenic stability of a pair of pool-cooled helical coils. Composite conductors had been developed to attain sufficient stability at high current density. However, it was revealed that a normal-zone could propagate below the cold-end recovery current by additional heat generation due to the slow current diffusion into a thick pure aluminium stabilizer. Besides, a novel detection system with pick-up coils along the helical coils revealed that normal-zones were initiated near the bottom of the coil where the field is not the highest. Therefore, the cooling condition around the innermost layers, the high field area, will be deteriorated at the bottom of the coil by bubbles gathered by buoyancy. In order to raise the operating currents, methods for improving the cryogenic stability have been examined and stability tests have been carried out with a model coil and small coil samples. We selected a method to lower temperatures of the coil and an additional cooler has been installed at the inlet of the coil. The outlet temperatures of the coil have been successfully lowered to 3.8 from 4.4 K of the saturated temperature, as planned.

The performance of net-current free heliotron plasmas has been developed by findings of innovative operational scenarios in conjunction with an upgrade of the heating power and the pumping/fuelling capability in the Large Helical Device (LHD). Consequently, the operational regime has been extended, in particular, with regard to high density, long pulse length and high beta. Diversified studies in LHD have elucidated the advantages of net-current free heliotron plasmas. In particular, an internal diffusion barrier (IDB) by a combination of efficient pumping of the local island divertor function and core fuelling by pellet injection has realized a super dense core as high as 5 × 10 20 m -3 , which stimulates an attractive super dense core reactor. Achievements of a volume averaged beta of 4.5% and a discharge duration of 54 min with a total input energy of 1.6 GJ (490 kW on average) are also highlighted. The progress of LHD experiments in these two years is overviewed by highlighting IDB, high β and long pulse.

In local island (m/n = 1/1) divertor discharges in the large helical device a stable super dense core plasma develops when a series of pellets are injected. A core region with a density as high as 4.6 × 1020 m−3 and a temperature of 0.85 keV is maintained by an internal diffusion barrier with a very high density gradient. In a study of island dynamics, we find that an externally imposed large island (m/n = 1/1) as large as 15% of the minor radius is healed when beta at the island exceeds a critical value.

Since the foundation of National Institute for Fusion Science (NIFS) in 1989, the primary mission of the applied superconductivity and cryogenic researches has been focused on the development of the large helical device (LHD): the largest fusion experimental apparatus exclusively utilizing superconducting technologies. The applied superconductivity and cryogenics group in NIFS was organized to be responsible for this activity. As a result of extensive research activities, the construction of LHD was completed in 1997. Since then, the LHD superconducting system has been demonstrating high availability of more than 97% during eight years operation and it keeps proving high reliability of large-scale superconducting systems. This paper describes the extensive activities of the applied superconductivity and cryogenic researches in NIFS during and after the development of LHD and the fundamental researches that aim at realizing a helical-type fusion reactor.

Helical coils of the large helical device are pool-cooled superconducting magnets. Although they were designed to be cryostable at 13.0 kA in saturated helium at 4.4 K, propagation of a normal zone has been observed several times at higher currents than 11.0 kA. The main cause is additional heat generation due to the slow current diffusion into a thick pure aluminum stabilizer. Methods for improving the cryogenic stability of the coils have been examined. Considering the withstanding voltage of the coil and the possibility of air mixing, the method of being subcooled has been selected. Design values of the mass flow and the inlet temperature are 50 g/s and 3.1 K, respectively, by utilizing the surplus power of an existing helium refrigerator. The average temperature of the helical coil is expected to be lowered to 3.5 K. In order to estimate the effect, cryogenic stability tests have been carried out with a model coil. According to the test results and comparison of the difference of the magnetic field between the model coil and the helical coil, the operating current is expected to be increased to 12.0 kA that corresponds to 3.0 T at the major radius of 3.6 m.

This paper is devoted to the analysis of the possible experiments with the D + D in Large Helical Device geometry and dynamics of fusion products. The transfer of energy from the fusion product to background plasma particles is discussed and the separation of the fusion products with the use of the drift resonances is considered.

A new divertor configuration called helical X point divertor (HXD) for a heliotron type fusion reactor FFHR is proposed, which has relatively wide target plates near two X points helically rotating along the torus. It has a clear divertor magnetic configuration with enough space to install blanket modules all around the torus. From the numerical calculation, it has been found that the wetted area on target plates for the divertor flux is sufficiently large to receive its high heat load. In this new divertor configuration, there exists another attractive feature to enhance the divertor function, with existence of intrinsic magnetic islands in the edge ergodic region.

The large helical device is a heliotron device with L = 2 and M = 10 continuous helical coils and three pairs of poloidal coils, and all of coils are superconductive. Since the experiments started in 1998, the development of engineering technologies and the demonstration of large-superconducting-machine operations have greatly contributed to an understanding of physics in currentless plasmas and a verification of the capability of fully steady-state operation. In recent plasma experiments, the steady state and high-beta experiments, which are the most important subjects for the realization of attractive fusion reactors, have progressed remarkably and produced two world-record parameters, i.e. the highest average beta of 4.5% in helical devices and the highest total input energy of 1.6 GJ in all magnetic confinement devices. No degradation has been observed in the coil performance, and stable cryogenic operational schemes at 4.4 K have been established. The physics and engineering results from the LHD experiment directly contribute to the design study for a D-T fusion demo reactor FFHR with a LHD-type heliotron configuration.

This paper is devoted to the analysis of the new physics mechanism of the removal of cold alpha particles from the fusion helical reactor. Earlier, we have proposed some mechanisms, such as the drift island motion and estafette of drift resonances to remove the test particles, namely the cold alpha particles from the helical device of a LHD type. In that time, we kept in mind the cold alpha particles with the energy W = 350 keV. Now, we concentrate our study on the removal of the helium ash, i.e. cold alpha particles with the energy W = 35 keV while the main plasma ions are confined.

One of the characteristics of the heliotron-type magnetic configuration is that it has an intrinsic divertor structure (helical divertor). Particle control using a helical divertor configuration, to achieve improved confinement and sustainment of steady-state high-performance plasmas, is a major experimental goal in the Large Helical Device (LHD), the largest heliotron-type superconducting device, and it needs to be demonstrated on the route to the design of the heliotron-type fusion reactor. The LHD scrape-off layer (SOL) in the intrinsic helical divertor configuration has a unique magnetic field line structure consisting of stochastic regions, residual islands, whisker structures, and laminar layers contrasting with the "onion-skin"-like magnetic field line structure in poloidal divertor tokamak SOLs. Since the first experimental campaign in LHD in 1998, studies aiming at understanding the edge plasma properties in the "open" helical divertor configurations have been conducted experimentally and theoretically. In this paper, the helical divertor studies in the LHD are reviewed, and the future experimental plan is shown.

A repetitive pellet injector has been developed and pellet refuelling experiments have been launched on the Large Helical Device (LHD) in a new regime characterized by high density steady state operation. In order to suppress the disturbance to the core plasma as much as possible, we employ relatively small and slow pellets. The size and velocity of the pellets are typically 2.5?mm in diameter and 250?m?s?1, respectively. The pellets were injected into neutral beam heated plasmas with repetition frequency of 10?Hz. The density rise per pellet is 1.5 ? 1019?m?3 and the pellet penetration depth, which is normalized by the plasma minor radius, is typically less than 0.5. Quasi-steady-state operation was achieved at plasma parameters of , Te = Ti = 1.0?keV. Repetitive pellet fuelling showed improved energy confinement compared with massive gas puff fuelling in the high density regime in common with previous transient pellet injection experiments, and it can sustain such an improved confinement property.

The divertor/SOL transport of the Local Island Divertor (LID) configuration in the Large Helical Device (LHD) is studied. Since the LID configuration has an inherent three dimensional magnetic structure, the 3D edge transport code, EMC3-EIRENE, is applied in order to elucidate the 3D divertor transport physics. The main observations reported here are: 1. The effective temperature cooling due to the integrated perpendicular energy loss along the long connection length. 2. The significant plasma flow acceleration caused by the upstream source that is attributed to the geometrical significance of the LID configuration. 3. The effect of the flow acceleration on the impurity screening considering a parallel transport along the SOL.

The long-pulse plasma discharge experiment is an important experiment in the Large Helical Device, which has a superconducting coil system and the capability of steady-state operation. The experiment of long-pulse plasma discharge was carried out using mainly ion cyclotron range of frequencies heating. The maximum plasma duration is 31 min and 45 s, and the total injected heating energy reached 1.3 GJ. Swing of the magnetic axis is adopted as an effective method to scatter the local heat load on the dive rtor plate during the discharge. The plasma was terminated abruptly by the influx of metallic impurities accompanied by a spark in the vacuum vessel.

The high compression of 600 times liquid density and the recent fast heating of a compressed core to 1-keV temperature have provided proof-of-principle of the fast ignition concept, and these results have significantly contributed to approve first phase of the Fast Ignition Realization EXperiment (FIREX) project. The goal of FIREX-I is to demonstrate fast heating of a fusion fuel up to the ignition temperature of 5–10 keV. Although the fuel size of FIREX-I is too small to ignite, sufficient heating will provide the scientific viability of ignition-and-burn by increasing the laser energy thereby the fuel size. Based on the result of FIREX-I, the decision of the start of FIREX-II to achieve ignition-and-burn can be made. The FIREX program is under the collaboration of the Institute of Laser Engineering and the National Institute for Fusion Science.

Helical coils of the Large Helical Device are pool-cooled superconducting magnets, in which propagation of a normal-zone has been observed several times at about 86% of the nominal current of 13.0 kA. It is planned to improve the cryogenic stability by lowering the inlet temperature. In order to estimate the effect, the cryogenic stability of a model coil of the helical coil was examined in saturated and subcooled helium. Liquid helium is supplied from the bottom of the model coil, and it is exhausted through the winding to the current-leads tank. The inlet helium is subcooled by a pre-cooler. A normal zone was initiated by a heater on the conductor at the bottom of the coil. In saturated helium of 4.4 K and 0.12 MPa, the minimum current to propagate over the next turn varies from 10.7 to 11.2 kA in the four cases that are without or with additional thermal shields, and before or after being subcooled. The difference is considered to be caused by the change of quality of saturated helium inside the winding or by the change of the wetted condition of the conductor surface. The minimum currents are higher at the lower temperatures in subcooled helium. It is raised up to 11.7 kA at 3.5 K of the temperature inside the winding. The propagation velocity at each minimum current is almost same. Namely, the propagation velocities at the same current are slower at the lower temperature in subcooled helium

Steady state divertor operation with high performance plasmas (ne ~ 0.7 × 1019 cm−3, Ti ~ 2 keV) was demonstrated for half an hour in the Large Helical Device (LHD), the superconducting helical device (R = 3.6–3.9 m, a = 0.6 m, B = 3 T, l/m = 2/10). The high performance plasmas have been sustained with an averaged heating power of 680 kW and achieved an injected energy of 1.3 GJ. This required both advanced technological integration of heating systems and divertor heat flux control. In particular, optimization of divertor heat flux distribution along the divertor leg trace on divertor plates and real-time magnetic axis sweeping (R = 3.67–3.7 m) have allowed LHD to access a steady state regime with a margin of safety for the actively cooled divertor plates. The distribution of divertor heat load along the traces was investigated with calorimetric measurements and it was found that there was a localized heat load connected with the loss of high-energy ions produced by ion cyclotron radio frequency near-fields. Orbit analysis shows that the behaviour of high-energy ions is qualitatively in good agreement with the experimental result. Long-pulse discharges were terminated by radiation collapse due to penetration of metallic flakes into the plasma.

Self-sustained detachment has been obtained in the Large Helical Device (LHD). Strong hydrogen gas puffing of ~200 Pa m3 s−1 after a density feedback phase detaches the plasma from the divertor plates with high reproducibility. High electron density of over 1 × 1020 m−3 is sustained without gas puffing until the heating beam stops and a high-density flat top for 2 s has been demonstrated. Throughout the self-sustained detachment phase, the minor radius of the hot plasma column shrinks to ~90% of the last-closed-flux-surface, which corresponds to the rational surface. This new state has been named the 'Serpens mode', for self-regulated plasma edge 'neath the last-closed-flux-surface. Global energy confinement of the Serpens mode is compared with the international stellarator scaling 1995 (ISS95) and the recently established scaling for high-density LHD plasmas (HD scaling), where shrinking confinement volume and shallow penetration of the heating beams are taken into account. Although the energy confinement of the Serpens mode seems deteriorated compared with ISS95, as in the case of high-density attached plasmas, it is consistent with the HD scaling. This suggests that the energy confinement properties of detached plasmas in LHD are similar to those in high-density attached plasmas.

In the Large Helical Device, the structure of localized oscillations at the m/n = 2/1 rational surface during neutral beam (NB) injection in the counter direction (counter-NB injection) has been investigated in detail using an electron cyclotron emission (ECE) diagnostic. Here m and n are poloidal and toroidal mode numbers, respectively. Under the condition of increasing plasma current in the counter direction and nearly constant pressure, the oscillations are observed at larger magnitudes of plasma current (−20 kA T−1). In the NB reversal experiments, where the direction of the NB is changed from the co- to the counter-direction and vice versa, the oscillations were observed only for the former case. These results suggest that the profile and magnitude of the plasma current play an important role in the mode onset. Analysis based on the Mercier criteria shows that for a flat or a hollow current profile an interchange mode becomes more unstable at higher negative plasma current, showing a tendency similar to the experimental observations. In addition, radial displacement due to the localized oscillations evaluated from electron temperature perturbation profiles with the ECE diagnostic agrees well with that predicted by a three-dimensional ideal magnetohydrodynamic stability code.

The hydrogen particle balance of the plasma-wall system in the large helical device (LHD) is analyzed, using a zero dimensional model for plasma particles, neutrals in vessel and hydrogen inventory in wall. Based on the measurement of neutral gas pressure, plasma density and the pumping speed of the cryo-pumping system, it is found that the hydrogen retained in the wall desorbes with short and long time constant. The short term desorption is of order of 10 21 atoms with a time constant of a few minutes, which is much smaller than the wall pumping for one shot, 10 22 atoms. In a long time scale of about one experimental day, the wall absorbs significantly large amounts of hydrogen, up to 10 24 atoms. One of the possible reasons for the large wall pumping is a carbon deposition layer on the first wall surface. The effect of hydrogen retention on density control is also discussed.

In the Large Helical Device (LHD), boronization using glow discharge with a mixture of helium and diborane gas was conducted in the 6th experimental campaign. In this campaign, material probes made of silicon and 316L stainless steel were placed on the plasma-facing wall in the toroidal direction. The thickness of the deposited boron and retained amounts of discharge gases (hydrogen and helium) were investigated. The toroidal distribution of the boron thickness significantly depended on the positions of the anodes and the diborane-inlet nozzles. The thickness of boron film and the atomic concentration of boron at the wall close to the anode and the nozzle were 400 nm and 80 at.%, respectively. At the wall far from the anodes and the nozzles, the boron film was very thin, only several nm. The trapped amount of oxygen had a tendency to increase with the thickness of boron film. The retained amount of helium at the wall close to the anodes was one order of magnitude larger than that far from the anodes. The retained amount of hydrogen at the wall with a thick boron film was observed to be smaller than that at the wall with a thin boron film.

Boron–titanium (B–Ti) was deposited on stainless steel substrate by evaporation of titanium and boron followed by annealing. The boron concentration at the surface region was comparable with the titanium concentration. Thermal desorption spectrum of deuterium was obtained after deuterium ion irradiation at room temperature, and the amount of retained deuterium and desorption temperatures of deuterium were measured. For comparison, titanium (Ti) and boron (B) films were deposited on the stainless steel substrate and the thermal desorption spectra were similarly obtained. The amount of retained deuterium in B–Ti was approximately 1/3 or 1/2 of that in Ti or B, respectively. The peak temperatures in B–Ti were 470 and 620 K, which were lower than those in B, 500 and 720 K. These results suggest that the use of B–Ti may reduce fuel hydrogen retention and recycling in fusion experimental devices, compared with the case of boron.

The apparatus for cryogenic target experiments of the Fast Ignition Realization EXperiment (FIREX) project has been developed at the National Institute for Fusion Science (NIFS). To achieve the target specifications for the project, the foam shell method has been proposed. The Institute of Laser Engineering (ILE), Osaka University and NIFS have started a cooperative study of the method. NIFS is responsible for the demonstration of the fuel layering process. The first cool-down test of the apparatus was conducted. Its cool-down performance is described in this paper.

A long-pulse plasma discharge of more than 30 min duration was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8 × 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. The total injected heating energy was 1.3 GJ. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by sweeping the magnetic axis inward and outward. Causes limiting the long pulse plasma discharge are discussed. An ion impurity penetration limited further long-pulse discharge in the 8th experimental campaign (2004).

Density and power scan experiments have been carried out at various magnetic field strengths on the Large Helical Device (LHD), to investigate the temperature and magnetic field dependences of the thermal diffusivity. In the moderate density regime, the thermal diffusivity shows gyro-Bohm-like parameter dependences. In the high-density and low-temperature regime, on the other hand, the thermal diffusivity increases with the square root of the local electron temperature and decreases with a 1.2 power of the magnetic field strength. The turning point temperature where the weak temperature dependence changes to the gyro-Bohm type is also found to increase with a 0.8 power of the magnetic field strength. Based on the experimental results, dimensionally correct scalings for the energy confinement time, τE, and the plasma stored energy for the high-density LHD plasmas have been derived. Compared with the gyro-Bohm model (τE ∝ (n/P)3/5, where n and P denote the density and the heating power, respectively) and the international stellarator scaling 1995 (τE ∝ n0.51P−0.59), our scaling has a weaker positive dependence on the density together with the mitigated power degradation as τE ∝ (n/P)1/3, owing to the weak temperature dependence of the thermal diffusivity.

In the Large Helical Device (LHD), two different divertor configurations, i.e. helical divertor (HD) and local island divertor (LID), are utilized to control the edge plasma. The HD with two X-points is an intrinsic divertor for heliotron devices, accompanied with a relatively thick ergodic layer outside the confinement region. Edge and divertor plasma behavior from low density to high density regimes is presented, referring to the divertor detachment. The effect of the ergodic layer on the edge transport is also discussed. On the other hand, the LID is an advanced divertor concept which realizes a high pumping efficiency by the combination of an externally induced magnetic island and a closed pumping system. Experimental results to confirm the fundamental divertor performance of the LID are presented.

A long-pulse plasma discharge for more than 30 min. was achieved on the Large Helical Device (LHD). A plasma of ne = 0.8× 1019 m−3 and Ti0 = 2.0 keV was sustained with PICH = 0.52 MW, PECH = 0.1 MW and averaged PNBI = 0.067 MW. Total injected heating energy was 1.3 GJ, which was a quarter of the prepared RF heating energy. One of the keys to the success of the experiment was a dispersion of the local plasma heat load to divertors, accomplished by shifting the magnetic axis inward and outward.

The Large Helical Device is a heliotron device with L = 2 and M = 10 continuous helical coils with a major radius of 3.5–4.1 m, a minor radius of 0.6 m and a toroidal field of 0.5–3 T, which is a candidate among toroidal magnetic confinement systems for a steady state thermonuclear fusion reactor. There has been significant progress in extending the plasma operational regime in various plasma parameters by neutral beam injection with a power of 13 MW and electron cyclotron heating (ECH) with a power of 2 MW. The electron and ion temperatures have reached up to 10 keV in the collisionless regime, and the maximum electron density, the volume averaged beta value and stored energy are 2.4 × 1020 m−3, 4.1% and 1.3 MJ, respectively. In the last two years, intensive studies of the magnetohydrodynamics stability providing access to the high beta regime and of healing of the magnetic island in comparison with the neoclassical tearing mode in tokamaks have been conducted. Local island divertor experiments have also been performed to control the edge plasma aimed at confinement improvement. As for transport study, transient transport analysis was executed for a plasma with an internal transport barrier and a magnetic island. The high ion temperature plasma was obtained by adding impurities to the plasma to keep the power deposition to the ions reasonably high even at a very low density. By injecting 72 kW of ECH power, the plasma was sustained for 756 s without serious problems of impurities or recycling.

"It is very important to demonstrate the ability to sustain the plasma in a steady state on the Large Helical Device (LHD), which has external helical magnetic coils and is a superconducting device. The long pulse discharge experiment was carried out using the ion cyclotron range of frequencies (ICRF) heating mainly. The plasma discharge of 31 minutes and 45 seconds was achieved by a total injected heating energy of 1.3GJ. Swing of the magnetic axis to scatter the local heat load on the divertor plate was one of the key methods for the steady state operation. The repetitive hydrogen pellet injection was tried successfully to fuel the minority hydrogen ions for long pulse operation."