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Toward the 1 MV power supply system for the ITER neutral beam injector consisting of various components with oil insulation, gas insulation and vacuum insulation, design and manufacturing of dc ultra-high-voltage component have been progressed. Thermal design of long transmission line up to 100 m was established. In order to design the high-temperature water channel for the plasma grid at 1 MV potential, the resistivity of high-purity water at 180 °C was measured for the first time. With the obtained resistivity, realistic high-temperature water circulating system was designed. A high-voltage bushing to transmit multiple voltages simultaneously has been also developed through a careful experimental study on vacuum insulation characteristics with multi-gaps coaxial electrodes. With the empirical scaling the electrostatic screen was designed taking into account effect of multi-gap and expected deformation such as thermal elongation during operation. After the assembly, stable voltage holding of 240 kV at each stage was confirmed in the manufactured HV bushing.

The long-pulse acceleration of the high-power positive ion beam has been demonstrated with the JT-60 positive ion source in the joint experiment among Japan Atomic Energy Agency (JAEA), Korea Atomic Energy Research Institute (KAERI) and National Fusion Research Institute (NFRI) under the collaboration program for the development of plasma heating and current drive systems. In this joint experiment, the increase of the heat load and the breakdowns induced by the degradation of the beam optics due to the gas accumulation was one of the critical issues for the long-pulse acceleration. As a result of development of the long-pulse operation techniques of the ion source and facilities of the neutral beam test stand in KAERI, 2 MW 100 s beam has been achieved for the first time. The achieved beam performance satisfies the JT-60SA requirement which is designed to be a 1.94 MW ion beam power from an ion source corresponding to total neutral beam power of 20 MW with 24 ion sources. Therefore, it was found that the JT-60 positive ion sources were applicable in the JT-60SA neutral beam injectors. Moreover, because this ion source is planned to be a backup ion source for KSTAR, the operational region and characteristic has been clarified to apply to the KSTAR neutral beam injector.

Toward ITER, the long-pulse acceleration of MeV-class negative ion beams has been successfully demonstrated and R&Ds and the procurement of high-voltage components of 1 MV power supply system have been progressed. In the accelerator development after the production of 1 MeV beams, an extraction grid, where a water cooling is reinforced and the beam is steered by aperture offset to suppress grid heat loads, has been newly developed to extend the pulse duration of the MeV-class beams. As the result, a total grid heat load has been reduced from 20% to 10% of the total beam power. The beam pulse was successfully increased from 0.4 s to 60 s at 0.68 MeV, 100 A/m 2 and 12 s at 0.89 MeV, 156 A/m 2 beams. As for the procurement of the 1 MV power supply system, 1 MV insulating transformer as the critical component has been developed. In the mock-up, a dc 1.2 MV for 1 h with the margin of 20% was stably sustained. Through the R&Ds of the critical components, design of five power supplies procured by JAEA has been finalized, and out of which a testing power supply has been completed in June 2014.

The HV bushing has a function of a bulk head between insulation gas (SF6) of 0.6 MPa and vacuum, which is needed to insulate −1 MV in vacuum in ITER heating neutral beam (NB) injection system. Mechanical analyses were carried out to verify the mechanical integrity of the HV bushing composed of five-stage double-layered insulator columns with large brazed ceramic ring and fiber reinforced plastic (FRP) ring of which non-metallic material is required to satisfy a safety factor of ≥3.5. As for the FRP ring required sustaining the pressure load, seismic load and dead weight, it was confirmed that isotropic fiber cloth FRP rings having sufficient strength against shear stress should be used. The shape and fixation area of the Kovar sleeve were modified to lower the stress at the brazing area having vacuum leak tightness and pressure tightness against the air filled at 0.6 MPa. As a result, a design of the insulator for the HV bushing was established satisfying the requirement.

Vacuum insulation of −1 MV is a common issue for the HV bushing and the accelerator for the ITER neutral beam injector (NBI). The HV bushing as an insulating feedthrough has a five-stage structure and each stage consists of double-layered insulators. To sustain −1 MV in vacuum, reduction of electric field at several triple points existing around the double-layered insulators is a critical issue. To reduce electric field simultaneously at these points, three types of stress ring have been developed. In a voltage holding test of a full-scale mockup equipped with these stress rings, 120% of rated voltage was sustained and the voltage holding capability required in ITER was verified. In the MeV accelerator, whose target is the acceleration of a H− ion beam of 1 MeV, 200 A/m2, the gap between the grid support was extended to suppress breakdowns triggered by electric field concentration at the edge and corner of the grid support. This modification improved the voltage holding capability in vacuum, and the MeV accelerator succeeded in sustaining −1 MV stably. Furthermore, it appeared that the H− ions beam was deflected and a part of the beam was intercepted at the acceleration grid. This causes high heat load on the grids and breakdowns during beam acceleration. To suppress the direct interception, a new grid was designed with proper aperture displacement based on a three dimensional beam trajectory analysis. As a result, 980 keV, 185 A/m2 H− ion beam acceleration has been demonstrated, which is close to the ITER requirement.

The neutral beam injection (NBI) system was designed to provide plasma heating and current drive for high performance and long pulse operation of the Korean Superconducting Tokamak Advanced Research (KSTAR) device using two co-current beam injection systems. Each neutral beam injection system was designed to inject three beams using three ion sources and each ion source has been designed to deliver more than 2.0 MW of deuterium neutral beam power for the 100-keV beam energy. Consequently, the final goal of the KSTAR NBI system aims to inject more than 12 MW of deuterium beam power with the two NBI for the long pulse operation of the KSTAR. As an initial step toward the long pulse (∼300 s) KSTAR NBI system development, the first neutral beam injection system equipped with one ion source was constructed for the KSTAR 2010 campaign and successfully commissioned. During the KSTAR 2010 campaign, a MW-deuterium neutral beam was successfully injected to the KSTAR plasma with maximum beam energy of 90 keV and the L-H transition was observed with neutral beam heating. In recent 2011 campaign, the beam power of 1.5 MW is injected with the beam energy of 95 keV. With the beam injection, the ion and electron temperatures increased significantly, and increase of the toroidal rotation speed of the plasma was observed as well. This paper describes the design, construction, commissioning results of the first NBI system leading the successful heating experiments carried in the KSTAR 2010 and 2011 campaign and the trial of 300-s long pulse beam extraction.

A supercritical-water-cooled reactor (SCWR) is a high-temperature, high-pressure water cooled reactor that operates above the critical pressure of water. In order to perform efficiently the thermal design of the SCWR, it is important to assess the thermal hydraulics in rod bundles of the core. Experimental conditions of mockup tests, however, may be limited because of technical and financial reasons. Therefore, it is required to establish an analytical design technique that can extrapolate experimental data to various design conditions of the reactor. Japan Atomic Energy Agency (JAEA) has improved the three-dimensional two-fluid model analysis code ACE-3D, which was originally developed for the two-phase flow thermal hydraulics of light water reactors, to handle the thermal hydraulic properties of water in the supercritical region. In the present study, heat transfer experiments of supercritical water flowing in a vertical annular channel around a heater pin, which were performed at JAEA, were analyzed wi...

Fluid flow and heating tests on the full-scale Test Blanket Module (TBM) First Wall (FW) mock-up under the TBM-relevant operating condition were carried out in an ion beam irradiation system with a pressurized high temperature water loop in JAEA to verify the fabrication process and design of the TBM First Wall and demonstrate its heat removal capability. The TBM FW mock-up is made of reduced activation ferritic/martensitic (RAFM) steel, F82H, and its parts such cooling channels and plates were assembled with Hot Isostatic Pressing (HIP) method. In the fluid flow test, the cooling water at room temperature is supplied to 15 parallel flow paths in the mock-up. The flow distribution in each path inside the First Wall mock-up is compared with a numerical simulation. This result shows that no severe cross-sectional deformation of the entire flow path in the mock-up takes place during the fabrication process. In the heating test, the mock-up was exposed to repetitive hydrogen ion beam irradiation with the maximum heat flux of 0.5 MW/m2. No indication of joint defect of the HIP joint like a hot spot was observed during 80 irradiation cycles.

Development of Water Cooled Solid Breeder (WCSB) TBM, the primary candidate of ITER Test Blanket Module (TBM), has been performed in Japan, according to the TBM milestones, which are necessary for acceptance of the TBM in ITER for testing from the first day of plasma operation. The TBM milestones consist of milestones on safety assessment, module qualification and design integration in ITER. For the safety milestones, essential source terms were evaluated, and failure modes and effect analysis (FMEA) was performed. Based on the results of FMEA, safety assessment was performed. For the milestones of the design integration, detailed structural design of the TBM and the interface structure with the ITER test port were performed. Based on the design, performance analysis such as thermo-mechanical analysis in over-pressurization was performed. For the milestones of the qualification of fabrication technology, essential fabrication technology was developed and near full size first wall of the TBM was successfully fabricated and demonstration of the integrity in heat flux equivalent to ITER. The development of the WCSB TBM is showing steady progress toward the installation in ITER.

This paper reports R&D progress of the MeV class accelerator, beam focussing technique and a high voltage (HV) bushing toward the ITER Neutral Beam system at JAEA. A H − ion beam of 320 mA (current density: 140 A/m 2 ) was successfully accelerated up to 796 keV by fixing air leak in the ion source chamber caused by heat loading due to backstream positive ions. From analyses of a three-dimensional beam trajectory, a countermeasure against beamlet deflection due to space charge repulsion among beamlets was identified to compensate the deflection by proper aperture offset in the electron suppression grid. For development of the HV bushing, a half-size (O800 mm) mockup single stage bushing was manufactured and tested, that demonstrated the dc voltage holding capability of up to 220 kV in a condition with vacuum inside. Electrostatic analyses and design of the HV bushing are in progress, in which some design improvements are suggested to reduce electric field strength on the surface of ceramic insulator.

As part of the development of plasma-facing components (PFCs) for fusion machines, JAEA has been developing high-performance cooling tubes with pressurized water flow. Along this line, a cooling tube with a helical triangular fin on its inner surface has been recently proposed for application in a fusion DEMO to enhance heat removal. Since the fin can be machined via simple mechanical threading, this tube is called as a “screw tube”. The divertor cooling conditions for the JAEA DEMO design will be a pressure of 4 MPa and an outlet temperature of 200 °C, in order to improve the thermal efficiency of power generation. In the this study, the effect of subcooling on the critical heat flux (CHF) by the screw tube has been investigated under DEMO divertor-relevant conditions, with the local pressure of 4 MPa and the inlet coolant temperature up to 180 °C. A test sample of the screw tube is made of pure Cu instead of F82H, a candidate structural material for the DEMO divertor. The results show that the ICHF values of the screw tube are more than double the values of the smooth tube at the inlet coolant temperature of 180 °C, although the temperature rise of the cooling water by 140 K leads to a reduction in the ICHF by almost half as compared with those values at room temperature.

The paper reports recent progress of R&D towards the beam source and HV bushing of the ITER Neutral Beam System. A new magnetic configuration with a “tent” magnetic filter was tested to suppress localization of fast electrons in a large negative ion source. The result was a drastic improvement of the negative ion uniformity with a beam intensity deviation less than ±4%. This fulfills the requirement of ITER (deviation ≤±5%). In a 1 MeV accelerator development, experiment of high current beam acceleration is in progress. The H − ion beams of 0.24 A has been obtained so far, at beam energy of 800 keV. Concerning the insulator rings of the HV bushing, a new cold-isostatic press method was developed to press alumina powder to the ring form. Formation and sintering of a full-scale (1.56 m in diameter) ceramic insulator ring was successfully completed without damages or cracks in the ceramic body.

This paper reports the progress of R&D towards the ITER neutral beam (NB) system at the Japan Atomic Energy Agency (JAEA) during 2004–2006. In the accelerator development, an H−ion beam of 146 A m−2was accelerated up to 836 keV in a vacuum insulated accelerator with five acceleration stages. This is the first production of a high power density beam at MeV range energy, which is more than twice as high as those of the existing NB systems for large experimental reactors such as JT-60U and LHD. The ITER NB system requires a high voltage (HV) bushing as an electric feedthrough to the vacuum insulated accelerator from the gas insulated power supply. A full-size (1.56 m in diameter and 0.292 m in height) insulator ring made of alumina ceramic has been successfully fabricated for HV bushing, as a result of the technological development of a new cold iso-static press.Long pulse operations of JT-60U negative ion sources with a large extraction area of 0.45 × 1.1 m2are in progress. From the two ion sources, D−ion beams of 21 A (90 A m−2), 320 keV have been produced for 21 s. As a consequence, injection of 3.2 MW D0beams was achieved. Thus, the long pulse operation of the negative ion based NB system has been demonstrated with multi-tens of ampere of D−ions extracted from the ion extraction area relevant to the ITER NB design (0.6 × 1.5 m2). For long pulse operation of such powerful beams, the negative ion uniformity over the wide extraction area is an essential issue to avoid excess power loading on the accelerator grids. By adopting a new magnetic configuration with the so-called ‘tent-shaped filter’, the uniformity of the JAEA 10 A source (extraction area: 0.13 × 0.22 m2) was improved by 8%, which fulfilled the ITER requirement.

A DC 1 MV insulating transformer mock-up has been successfully developed for the ITER Neutral Beam Injector. The transformer has a newly developed composite bushing consisting of a small condenser bushing and fiber reinforced plastic (FRP) tube. The transformer mock-up has successfully fulfilled the ITER requirement of DC 1.2 MV for 3600 s insulation. In a development the high-voltage (HV) bushing, vacuum insulation characteristics of the multi-layer cylindrical electrodes have been investigated. As a result, it has been clarified that the voltage holding of multi-gap and meter-class electrode depends on the surface area of the electrode. Based on this result, dimensions and locations of the screen for the ITER HV bushing has been designed to realize a stable insulation of 1 MV in vacuum.

High power and long-pulse negative ion extractor, which is composed of the plasma grid (PG) and the extraction grid (EXG), is newly developed toward the neutral beam injector for heating and current drive of future fusion machines such as ITER, JT-60 Super Advanced and DEMO reactor. The PG is designed to enhance surface production of negative ions efficiently by applying the chamfered aperture. The efficiency of the negative ion production for the discharge power increased by a factor of 1.3 against that of the conventional PG. The EXG is also designed with the thermal analysis to upgrade the cooling capability for the long pulse operation of >1000 s required in ITER. Though the magnetic field for electron suppression is reduced to 0.75 of that in the conventional EXG due to this upgrade, it was experimentally confirmed that the extracted electron current can be suppressed to the allowable level for the long pulse operation. These results show that newly developed extractor has the high potential for the long pulse extraction of the negative ions.

A shielding blanket design in a fusion reactor such as ITER (International Thermonuclear Experimental Reactor) has been proposed to be a modular structure integrated with the first wall. In terms o...

Development of an electrostatic accelerator for 1 MeV, 1 ampere class H− ion beams has been carried out as a Proof-of-Principle (PoP) of high current accelerators required in neutral beam system of fusion reactors such as ITER. A unique feature of such fusion oriented accelerators is to utilize vacuum insulation not only inside but also outside of the accelerator for high voltage insulation, since conventional gas insulation is not applicable due to excess radiation induced conductivity in the gas. The PoP accelerator is surrounded by an insulator stack as a vacuum boundary with a vacuum gap of 50 mm all around the accelerator. One of the key technologies to achieve insulation of MV voltage was reduction of electric field strength at triple junction of the insulator (made of FRP, fiber reinforced plastic) stack. By lowering the stress to 1.7 kV/mm, the accelerator sustained 1 MV stably for more than 2 hours. By tuning the ion source with/without cesium under stable accelerator operation, the PoP accelerat...

In large negative ion sources of the neutral beam injector (NBI) for large fusion devices, such as ITER and JT-60SA, one of issues is that negative ion beam might be partially intercepted at acceleration grids due to a spatial non-uniformity of negative ion production on large extraction area. Previous experiments have shown that fast electrons emitted from filament cathodes are transported in a longitudinal direction by B × grad B drift and the spatial distribution of electron temperature (Te) is strongly related with the non-uniformity of negative ion production. In this study, a three-dimensional electron transport analysis code including detailed collision processes has been developed to clarify a physical mechanism of non-uniform Te distribution. Electron density and temperature in the analysis agree well with measurements by Langmuir probe in the JAEA 10A ion source. Then, this study has clarified that the non-uniformity of Te distribution is mainly caused by the following reasons; i) fast electrons...

Compensation methods of beamlet deflections have been studied in a three dimensional (3D) beam analysis using OPERA-3d code for 1 MeV accelerator of the ITER neutral beam injector (NBI). The beamlet deflection is caused by i) magnetic field generated by permanent magnets embedded in the extraction grid (EXG) for electron suppression and ii) space charge repulsion between the beamlets and beam groups. Moreover, the beamlet deflection is caused due to electric field distortion formed by a grid support structure. In order to compensate the beamlet deflections due to i) and ii), an aperture offset of 0.6 mm was applied in the electron suppression grid (ESG) and a metal bar with 3 mm in thickness, so-called a kerb, was attached around the aperture area at the back side of the ESG, respectively. Detailed configuration of the compensation methods was also considered so as to suppress the beam spread due to the electric field distortion and to lower electric field concentrations at the edge of the kerb. For the beamlets near the grid support structure, the beamlet deflection due to the space charge repulsion could be negated due to the electric field distortion formed by the grid support structure.

Voltage holding test on MeV accelerator indicated that sustainable voltage was a half of that of ideal quasi-Rogowski electrode. It was suggested that the emission of the clumps is enhanced by a local electric field concentration, which leads to discharge initiation at lower voltage. To reduce the electric field concentration in the MeV accelerator, gaps between the grid supports were expanded and curvature radii at the support corners were increased. After the modifications, the accelerator succeeded in sustaining -1 MV in vacuum without beam acceleration. However, the beam energy was still limited at a level of 900 keV with a beam current density of 150 A∕m(2) (346 mA) where the 3 × 5 apertures were used. Measurement of the beam profile revealed that deflection of the H(-) ions was large and a part of the H(-) ions was intercepted at the acceleration grid. This causes high heat load on the grids and the breakdowns during beam acceleration. To suppress the direct interception, new grid system was designed with proper aperture displacement based on a 3D beam trajectory analysis. As the result, the beam deflection was compensated and the voltage holding during the beam acceleration was improved. Beam parameter of the MeV accelerator was increased to 980 keV, 185 A∕m(2) (427 mA), which is close to the requirement of ITER accelerator (1 MeV, 200 A∕m(2)).

A plasma generator for a long pulse H(+)/D(+) ion source has been developed. The plasma generator was designed to produce 65 A H(+)/D(+) beams at an energy of 120 keV from an ion extraction area of 12 cm in width and 45 cm in length. Configuration of the plasma generator is a multi-cusp bucket type with SmCo permanent magnets. Dimension of a plasma chamber is 25 cm in width, 59 cm in length, and 32.5 cm in depth. The plasma generator was designed and fabricated at Japan Atomic Energy Agency. Source plasma generation and beam extraction tests for hydrogen coupling with an accelerator of the KSTAR ion source have been performed at the KSTAR neutral beam test stand under the agreement of Japan-Korea collaborative experiment. Spatial uniformity of the source plasma at the extraction region was measured using Langmuir probes and ±7% of the deviation from an averaged ion saturation current density was obtained. A long pulse test of the plasma generation up to 200 s with an arc discharge power of 70 kW has been successfully demonstrated. The arc discharge power satisfies the requirement of the beam production for the KSTAR NBI. A 70 keV, 41 A, 5 s hydrogen ion beam has been extracted with a high arc efficiency of 0.9 -1.1 A/kW at a beam extraction experiment. A deuteron yield of 77% was measured even at a low beam current density of 73 mA/cm(2).

In a five stage multi‐aperture and multi‐grid (MAMuG) accelerator in JAEA, beam acceleration tests are in progress toward 1 MeV, 200 A/m2 H− ion beams for ITER. The 1 MV voltage holding has been successfully demonstrated for 4,000 s with the accelerator of expanded gap length that lowered local electric field concentrations. This led to increase of the beam energy up to 900 keV‐level. However, it was found that beamlets were deflected more in long gaps and direct interceptions of the deflected beamlet caused breakdowns. The beamlet deflection and its compensation methods were studied utilizing a three‐dimensional multi beamlet analysis. The analysis showed that the 1 MeV beam can be compensated by a combination of the aperture offset of 0.8 mm applied in the electron suppression (ESG) and the metal bar called a field shaping plate with a thickness of 1 mm attached beneath the ESG. The paper reports these compensation methods and analytical predictions, with experimental results of the MAMuG accelerator in...

Heat load on acceleration grids by secondary particles such as electrons, neutrals, and positive ions, is a key issue for long pulse acceleration of negative ion beams. Complicated behaviors of the secondary particles in multiaperture, multigrid (MAMuG) accelerator have been analyzed using electrostatic accelerator Monte Carlo code. The analytical result is compared to experimental one obtained in a long pulse operation of a MeV accelerator, of which second acceleration grid (A2G) was removed for simplification of structure. The analytical results show that relatively high heat load on the third acceleration grid (A3G) since stripped electrons were deposited mainly on A3G. This heat load on the A3G can be suppressed by installing the A2G. Thus, capability of MAMuG accelerator is demonstrated for suppression of heat load due to secondary particles by the intermediate grids.

In JAEA, research and developments to realize high power accelerator (1 MeV, 40 AD(-) ion beams for 3600 s) for ITER have been carried out experimentally and numerically utilizing a five stage MAMuG (Multiaperture, Multigrid) accelerator. In this paper, the extension of the gap length, which is required to improve the voltage holding capability, is examined in two dimensional beam optics analyses and also from view point of stripping loss of ions. In order to suppress excess power loadings due to the direct interception of negative ions, which is issued in long pulse tests, the beamlet deflection is analyzed in three dimensional multibeamlet analyses. The necessary modifications shown above are applied to the MAMuG accelerator for coming long pulse tests in JAEA and ITER.

The J1-60 neutral beam system has been successfully operated for 4 years under a wide range of operation conditions: beam energy of 30-75 keV, beam pulse up to 6 s, injection power up to 26 MW with hydrogen beams. The maximum injection power of 26 MW was obtained at 73 keV with a two-stage accelerator. In a lower energy beam injection with a single-stage accelerator, the beam power at 38 keV reached 18 MW. The system could routinely inject a nominal power of 20 MW with high reliability. The beam energy could be changed during a beam pulse, e.g. from 40 keV to 70 keV for 1.5 s. Helium beams were injected with one of the beamlines for a simulation experiment of helium ash, and the injected power was 0.4 MW at 31 keV. The helium beams could deposit in the vicinity of the plasma center column. Helium gas in the beamline was evacuated by SF6 gas condensed cryo-sorption pumps whose pumping speed was about 800 m3/s.

The electron beam facility was constructed to test and develop plasma facing components. Maximum beam power reaches 400 kW. The facility has a plasma electron gun which is the largest one in the world. The power supply of the facility utilizes high-frequency invertors as a beam switching system. The plasma electron gun and power supply system were developed on the basis of neutral beam injectors for JT-60. Heating time ranges from 1 ms to continuous operation. Maximum head flux is over 2000 MW/m2, and heating area is up to 1800 cm2. The heating area can be changed by a number of grid apertures. The facility can produce intense heat loads at off-normal events such as disruption, and steady-state heat loads on the divertor and the first wali at normal operations.

A high proton yield beam was produced in a semicylindrical multicusp plasma generator. The dimension of the cylinder is 34 cm in diameter and 129 cm in length, and the beam extraction area is 8 cm in width and 120 cm in length. A strongly linked magnetic field configuration was utilized to confine the primary electrons in the cylinder and prevent them from flowing into the beam extraction region. As a temporary test hydrogen ion beams were extracted from the central region of the plasma generator. Ion species in the beam were measured by a Doppler‐shifted spectroscopy of Balmer‐alpha light and found to be 95% at a beam current density of 175 mA/cm2 at the filling pressure of 0.4 Pa.

A new electron gun utilizing a magnetic multipole plasma generator was designed and fabricated as the heat source of the high heat flux test facility, called JEBIS (JAERI electron beam irradiation stand). By changing the acceleration grids, this electron gun is able to produce a pencil to a sheetlike electron beams up to 4 A at 100 keV for 1 ms to continuous mode. In this electron gun, magnetic lens system is not adopted to focus the electron beam, but the space charge neutralization effect by the beam plasma produced downstream of the electron gun is utilized to prevent the blow‐up of the electron beam. In addition, high permeability metal is embedded in the first and the second grids to magnetically shield the earth field and the stray field from the beam bending magnet. It was experimentally demonstrated that wide range of heat flux from 0.2 MW/m2 to over 2000 MW/m2 can be realized at the test sample position about 1.7 m downstream of the electron gun.

Pumping characteristics of condensed SF6 layers for He gas have been measured in a small-size cryosorption pump. From the experimental results, He pressure dependence of the effective adsorption conefficient of He on condensed SF6 layers is estimated with the aid of numerical simulation.In the next stage of developing a large-scale cryosorption pump, which is a modification of the cryopump of the JT-60 neutral beam injector (NBI), the experimental results and numerical simulations are used to control the pumping performance of the SF6 layers on pumpmodules. The He pressures measured as a function of the He gas flow rate in the NBI chamber are in good agreement with the numerical calculations.

Effect of primary electron distribution on ion species yield is studied using a semicylindrical, multicusp plasma generator. Experimentally obtained, ion species yields were correlated with primary electron orbits estimated by computer simulation in several magnetic field configurations, where primary electron distributions were considered to be different. The result showed that a magnetic field configuration which produced lower proton yield had denser primary electron orbits in the neighborhood of the plasma grid. The floating potential of the plasma grid was also measured to suppose a distribution of the primary electrons. As a result, the proton yield increases with an increase of the floating potential of the plasma grid. These results lead us to confirm that the proton yield is determined by distribution of the primary electrons in the neighborhood of the plasma grid, in addition to the ratio of the plasma volume and the effective ion loss area. The highest proton yield of 93% and the lowest proton yield of 83% were obtained at the floating potential of the plasma grid of −5 and −21 V with respect to the anode potential, respectively, at an extraction current density of 150 mA/cm2.

Significant progress in the extension of pulse durations of powerful negative ion beams has been made to realize the neutral beam injectors for JT-60SA and ITER. In order to overcome common issues of the long-pulse production/acceleration of negative ion beams in JT-60SA and ITER, new technologies have been developed in the JT-60SA ion source and the MeV accelerator in Japan Atomic Energy Agency.As for the long-pulse production of high-current negative ions for the JT-60SA ion source, the pulse durations have been successfully increased from 30 s at 13 A on JT-60U to 100 s at 15 A by modifying the JT-60SA ion source, which satisfies the required pulse duration of 100 s and 70% of the rated beam current for JT-60SA. This progress was based on the R&D efforts for the temperature control of the plasma grid and uniform negative ion productions with the modified tent-shaped filter field configuration. Moreover, each parameter of the required beam energy, current and pulse has been achieved individually by these R&D efforts. The developed techniques are useful to design the ITER ion source because the sustainment of the caesium coverage in the large extraction area is one of the common issues between JT-60SA and ITER.As for the long-pulse acceleration of high power density beams in the MeV accelerator for ITER, the pulse duration of MeV-class negative ion beams has been extended by more than 2 orders of magnitude by modifying the extraction grid with a high cooling capability and a high transmission of negative ions. A long-pulse acceleration of 60 s has been achieved at 70 MW m−2 (683 keV, 100 A m−2) which has reached the power density of JT-60SA level of 65 MW m−2. No degradations of the voltage holding capability of the acceleration voltage and the beam optics due to the distortion of the acceleration grids have been observed in this power density level.These results are the longest pulse durations of high-current and high-power-density negative ion beams in the world.

A design of a high-voltage DC power supply for future neutral beam injectors (NBI) is presented together with the latest R&D results for a 100-kV, 5-A, 5-kHz inverter-type power supply. A negative-ion based NBI is used for heating and current drive in the JT-60U (JAERI Tokamak-60 upgrade) and the FER (Fusion Experimental Reactor). For a high-voltage DC power supply of 500-1000 kV, an inverter-type DC power supply has been adopted. It consists of low-voltage converters, inverters, step-up transformers, and high-voltage rectifiers. Ion sources are protected from electrical breakdowns by stopping the inverters. As the first step, the same system was used in the 100-kV, 5-A acceleration power supply of the JAERI (Japan Atomic Energy Research Institute) electron-beam irradiation stand. The power supply is based on 5-kHz inverters. After the dummy load test, the power supply operated in conjunction with a plasma electron gun, and a 100-keV, 4-A electron beam has already been extracted. >

The JT-60 JAERI (Japan Atomic Energy Research Institute) Tokamak-60 neutral beam injector (NBI) has been successfully operated for three years, delivering a neutral beam power of 20 MW at a beam energy of 75 keV. The helium-beam injection into JT-60 was performed to investigate exhaustion of fusion alpha particles from the plasma. The helium-ion beam current extracted from ion sources was 40-45 A at an energy of 31 keV. The injection power was about 0.4 MW. In the helium-beam operation, the helium gas in the beamline chamber was pumped out by a condensed SF/sub 6/ gas cryosorption pump. The JT-60 neutral beam injector will be modified to inject a 120-keV deuterium beam into the JT-60 upgrade, which will start to operate at the beginning of 1991. >

To develop plasma facing components (PFC) for the next fusion devices, JAERI has been carrying out high heat flux and high particle flux experiments on the divertor modules and candidate materials in JAERI Electron Beam Irradiation System (JEBIS). (1) To investigate the feasibility and the advantage of a saddle type divertor modules, which has unidirectional (1-D) carbon fiber reinforced carbon composites (CFCs) armour tile brazed on OFHC-copper heat sink, high heat flux experiments have been carried out under a cyclic heat load of 24.5 MW/m/sup 2/ at a duration of 30 s. After 1000 cycles, no degradation of thermal response and no defect in the module was found. (2) To reduce the residual stress around the brazed interface, we have developed small specimens with the new materials combination of W-30Cu composites heat sink and 1-D CFC armour tile and carried out the high heat flux experiments under a cyclic heat load of 15 MW/m/sup 2/ at a duration of 20 s. After 1,000 cycles, no cracks have been observed at the interface. (3) To evaluate the erosion of armour tiles by high heat flux, we have measured the erosion of CFCs and isotropic graphite up to 1100/spl deg/C under a heat flux of 1800 MW/m/sup 2/ for the duration of 1.5/spl sim/2 ms. It is clear that the erosion of carbon based materials increases with the bulk temperature and decreases with the thermal conductivity. (4) To evaluate the erosion by high particle flux, we have developed a new irradiation device, which have a high hydrogen particle flux of 10/sup 21//m/sup 2//s at 50/spl sim/100 eV.

Recent results of R&D on plasma facing components for fusion experimental reactors such as the ITER/FER (International Thermonuclear Experimental Reactor/Fusion Experimental Reactor) are presented. Plasma facing components will be exposed to severe heat loads. Thermal cycling tests for first wall mock-ups and divertor mock-ups have been carried out under specified heat flux conditions of ITER/FER in high heat flux test facilities at JAERI. Tests at a stationary heat flux of 0.2 MW/m/sup 2/ for first wall mock-ups, exposed in the normal heat flux part, have confirmed their durabilities and reliabilities against the cyclic heat loads. In experiments at a stationary heat flux of 0.6 MW/m/sup 2/, with exposure also in the high heat flux part, only the radiatively cooled first wall mock-up was tested. A ceramic sleeve covered with pure titanium between the armor and the cooling structure was damaged before reaching a stationary heat flux of 0.6 MW/m/sup 2/. The authors have also confirmed the integrity and the durability of the bonds in the thermal cycling test of brazed carbon-fiber-composite/copper divertor mock-ups at a simulated stationary surface heat flux of 10 MW/m/sup 2/ for 1000 cycles. >

Plasma Facing Components will be exposed to severe heat loads during the normal operation, and will severely be eroded by evaporation and particle emission during disruptions in next generation fusion devices such as International Thermonuclear Experimental Reactor (ITER). In this paper, we present recent major RD (1) thermal cycling experiments on a 1 m long divertor mock-up, (2) disruption simulation tests on B/sub 4/C doped CFCs and tungsten, (3) tensile and bending tests of beryllium/copper alloy duplex structures.

In order to realize neutral beam systems in International Thermonuclear Experimental Reactor whose target is to produce a 1 MeV, 200 A/m(2) during 3600 s D(-) ion beam, the electrostatic five-stages negative ion accelerator so-called "MeV accelerator" has been developed at Japan Atomic Energy Agency. To extend pulse length, heat load of the acceleration grids was reduced by controlling the ion beam trajectory. Namely, the beam deflection due to the residual magnetic field of filter magnet was suppressed with the newly developed extractor with a 0.5 mm off-set aperture displacement. The new extractor improved the deflection angle from 6 mrad to 1 mrad, resulting in the reduction of direct interception of negative ions from 23% to 15% of the total acceleration power, respectively. As a result, the pulse length of 130 A/m(2), 881 keV H(-) ion beam has been successfully extended from a previous value of 0.4 s to 8.7 s. This is the first long pulse negative ion beam acceleration over 100 MW/m(2).

The first neutral beam (NB) injection system of the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak was partially completed in 2010 with only 1∕3 of its full design capability, and NB heating experiments were carried out during the 2010 KSTAR operation campaign. The ion source is composed of a JAEA bucket plasma generator and a KAERI large multi-aperture accelerator assembly, which is designed to deliver a 1.5 MW, NB power of deuterium at 95 keV. Before the beam injection experiments, discharge, and beam extraction characteristics of the ion source were investigated. The ion source has good beam optics in a broad range of beam perveance. The optimum perveance is 1.1-1.3 μP, and the minimum beam divergence angle measured by the Doppler shift spectroscopy is 0.8°. The ion species ratio is D(+):D(2)(+):D(3)(+) = 75:20:5 at beam current density of 85 mA/cm(2). The arc efficiency is more than 1.0 A∕kW. In the 2010 KSTAR campaign, a deuterium NB power of 0.7-1.5 MW was successfully injected into the KSTAR plasma with a beam energy of 70-90 keV. L-H transitions were observed within a wide range of beam powers relative to a threshold value. The edge pedestal formation in the T(i) and T(e) profiles was verified through CES and electron cyclotron emission diagnostics. In every deuterium NB injection, a burst of D-D neutrons was recorded, and increases in the ion temperature and plasma stored energy were found.

Spatially non-uniform electron energy distribution function (EEDF) in an arc driven negative ion source (JAEA 10A negative ion source: 10 A NIS) is calculated numerically by a three-dimensional Monte Carlo kinetic model for electrons to understand spatial distribution of plasma production (such as atomic and ionic hydrogen (H(0)∕H(+)) production) in source chamber. The local EEDFs were directly calculated from electron orbits including electromagnetic effects and elastic∕inelastic collision forces. From the EEDF, spatial distributions of H(0)∕H(+) production rate were obtained. The results suggest that spatial non-uniformity of H(0)∕H(+) productions is enhanced by high energy component of EEDF.

The first neutral beam injector (NBI-1) has been developed for the Korea Superconducting Tokamak Advanced Research (KSTAR) tokamak. The first long pulse ion source (LPIS-1) has been installed on the NBI-1 for an auxiliary heating and current drive of KSTAR plasmas. The performance of ion and neutral beam extractions in the LPIS-1 was investigated initially on the KSTAR NBI-1 system, prior to the neutral beam injection into the main plasmas. The ion source consists of a magnetic bucket plasma generator with multipole cusp fields and a set of prototype tetrode accelerators with circular apertures. The inner volume of the plasma generator and accelerator column in the LPIS-1 is approximately 123 L. Design requirements for the ion source were a 120 kV/65 A deuterium beam and a 300 s pulse length. The extraction of ion beams was initiated by the formation of arc plasmas in the LPIS-1, called the arc-beam extraction method. A stable ion beam extraction of the LPIS-1 was achieved up to 85 kV/32 A for a 5 s pulse length and 80 kV/25 A for a 14 s pulse length. An optimum beam perveance of 1.15 µperv was observed at an acceleration voltage of 60 kV. Neutralization efficiency was measured by a water-flow calorimetry (WFC) method using a calorimeter and the operation of a bending magnet. The full-energy species of ion beams were detected by using the diagnostic method of optical multichannel analyzer (OMA). An arc efficiency of the LPIS was 0.6–1.1 A/kW depending on the operating conditions of arc discharge. A neutral beam power of ∼1.0 MW must be sufficiently injected into the KSTAR plasmas from the LPIS-1 at a beam energy of 80 keV.

A multiaperture multigrid accelerator called "MeV accelerator" has been developed for neutral beam injection system of international thermonuclear experimental reactor. In the present work, long pulse H(-) ion beam acceleration was performed by the MeV accelerator equipped with new water-cooled grids. At present, the pulse length was extended to 5 s for the beams of 750 keV, 221 mA, and 10 s for the beams of 600 keV, 158 mA. Energy density, defined as products of beam energy (keV), current (mA), and pulse (s) divided by aperture area (m(2)), increased more than one order of magnitude higher compared with original MeV accelerator without water cooling in its grids. At higher energy and current, the grid was melted by beam deflection. Due to this grid melting, breakdowns occurred between the grids, and hence, the pulse length was limited. Beam deflection will be compensated by aperture displacement in next experiment.

In order to realize a high power negative ion accelerator (1 MeV, 40 A D for 3600 s) in a neutral beam injector (NBI) of ITER, numerical approaches have been carried out in parallel to negative ion acceleration tests in JAEA. To investigate negative ion beam trajectories more precisely, a three dimensional (3D) beam analysis has been progressed by combination of a commercial 3D beam code and numerical codes developed in JAEA [1]. In a negative ion acceleration test for long pulses up to 30 s in JAEA, molten parts were observed around apertures of the grids. In order to examine whether or not direct interception of negative ions caused the meltdown, the beamlet deflection was examined in the 3D beam analyses. The calculation model was composed of a plasma grid, an extraction grid (EXG), three intermediate acceleration grids and a grounded grid (GRG), simulating a JAEA MAMuG accelerator used in the long pulse test. The H negative ion beams were extracted from fifteen apertures (14 mm in a diameter) drilled in a lattice pattern of 3 x 5. In the beam path in the accelerator, there are transverse magnetic filter fields generated by permanent magnets installed in the ion source and dipole magnetic fields generated by electron suppression magnets installed in the EXG. Those were precisely simulated in the present analyses and the beamlets were deflected due to the space charge repulsion and the magnetic fields. The 3D beam analyses were carried out utilizing OPERA-3d supplied by Vector Field Co. Ltd. As an input of OPERA-3d for initial positions of beam particles, a 2D beam code developed in JAEA, BEAMORBT, was utilized. The BEAMORBT calculates beam emitter surfaces according to a Child-Langmuir law. In order to take into account the stripping loss of ions, the background gas density in the accelerator was calculated utilizing a 3D Monte Carlo gas flow code developed in JAEA. The figure below shows power density around one aperture on the GRG surface and a picture of the GRG surface. In the calculation, the acceleration voltage was set to a typical value in this test, 600 kV. The extracted negative ion density was 140 A/m. A H2 gas pressure was 0.17 Pa in the ion source and 0.1 Pa at the accelerator exit. The stripping loss of ions was 34 %. The beamlets were deflected upward by the filter field and to left by the dipole magnetic field and space charge repulsion. The deflection angle at accelerator exit was 3.5 mrad upward and 11 mrad to left. This deflected beam caused high power loading of over 20 kW/cm along the aperture edge. This shows a good agreement with the molten parts. Thus the grid melt could have been caused by the beam deflection followed by the direct interception of negative ions. Toward a next long pulse test, an offset aperture grid to compensate the beamlet deflection will be installed.

At JAEA, as the Japan Domestic Agency (JADA) for ITER, a MAMuG (multi-aperture multi-grid) accelerator has been developed to perform the required R&D for the ITER neutral beam (NB) system. As a result of countermeasures to handle excess heat load to the ion source by backstreaming positive ions, H− ion beam current was increased to 0.32 A (the ion current density of 140 A m−2) at a beam energy of 796 keV. This high power beam acceleration simulated the ITER operation condition maintaining the perveance (H− ion current density/beam energy3/2) of the ITER accelerator. After the high power beam operation, the pulse length was successfully extended from 0.2 to 5 s at 550 keV, which yielded a 131 mA H− ion beam as an initial test of the long pulse operation. A test of a single-aperture single-gap (SINGAP) accelerator was performed at JAEA under an ITER R&D task agreement. The objective of this test was to compare two different accelerator concepts (SINGAP and MAMuG) at the same test facility. As a result, the MAMuG accelerator was defined as the baseline design for ITER, due to advantages in its better voltage holding and less electron acceleration. In three-dimensional beam trajectory analyses, the aperture offset at the bottom of the extractor was found to be effective for compensation of beamlet deflection due to their own space charge. It has been analytically demonstrated that these compensated beamlets can be focused at a focal point by adopting the aperture offset at the final grid of the accelerator.

Procurement of a dc −1 MV power supply system for the ITER neutral beam injector (NBI) is shared by Japan and the EU. The Japan Atomic Energy Agency as the Japan Domestic Agency (JADA) for ITER contributes to the procurement of dc −1 MV ultra-high voltage (UHV) components such as a dc −1 MV generator, a transmission line and a −1 MV insulating transformer for the ITER NBI power supply. The inverter frequency of 150 Hz in the −1 MV power supply and major circuit parameters have been proposed and adopted in the ITER NBI. The dc UHV insulation has been carefully designed since dc long pulse insulation is quite different from conventional ac insulation or dc short pulse systems. A multi-layer insulation structure of the transformer for a long pulse up to 3600 s has been designed with electric field simulation. Based on the simulation the overall dimensions of the dc UHV components have been finalized. A surge energy suppression system is also essential to protect the accelerator from electric breakdowns. The JADA contributes to provide an effective surge suppression system composed of core snubbers and resistors. Input energy into the accelerator from the power supply can be reduced to about 20 J, which satisfies the design criteria of 50 J in total in the case of breakdown at −1 MV.

The H(-) ion accelerator RD to realize the international thermonuclear experimental reactor neutral beam is ongoing at Japan Atomic Energy Agency (JAEA). The required performance for the prototype MeV accelerator developed at JAEA is 1 MeV, 500 mA (current density of 200 A/m(2)) H(-) ion beam at the beamlet divergence angle of less than 7 mrad. Up to 2005, 836 keV, 146 A/m(2) H(-) ion beam was successfully accelerated as the highest record of the current density at MeV class energy beams. In the present work, high current negative ion beam acceleration test was performed by increasing the beam extraction apertures from 3 x 3 (9 apertures) to 3 x 5 (15 apertures). By fixing the air leak at the source chamber due to backstream ions as well as the improvement of voltage holding capability by a new fiber reinforced plastic insulator ring, the performance of the MeV accelerator was improved. So far, H(-) ion beam of 320 mA was successfully accelerated up to 796 keV with the beam divergence angle of 5.5 mrad. The accelerated drain current including the electron reaches close to the power supply limit for the MeV test facility. The heat flux by the backstream ion during the above beam acceleration was estimated to be 360 W/cm(2). The Cs leakage to the accelerator during the test campaign (Cs total input of 5.0 g) was 0.26 mg (7.0 microg/cm(2)). This is considered to be the allowable level from the viewpoint of voltage holding.