Overview of recent developments in pellet injection for ITER

Pellet injection is the primary fueling technique planned for core fueling of ITER burning plasmas. Also, the injection of relatively small pellets to purposely trigger rapid small edge localized modes (ELMs) has been proposed as a possible solution to the heat flux damage from larger natural ELMs likely to be an issue on the ITER divertor surfaces. The ITER pellet injection system is designed to inject pellets into the plasma through both inner and outer wall guide tubes. The inner wall guide tubes will provide high throughput pellet fueling while the outer wall guide tubes will be used primarily to trigger ELMs at a high frequency (>15 Hz). The pellet fueling rate of each injector is to be up to 120 Pa m³/s, which will require the formation of solid D–T at a volumetric rate of ～1500 mm³/s. Two injectors are to be provided for ITER at the startup with a provision for up to six injectors during the D–T phase. The required throughput of each injector is greater than that of any injector built to date, and a novel twin-screw continuous extrusion system is being developed to meet the challenging design parameters. Status of the development activities is presented, highlighting recent progress.     

EAST ion cyclotron resonance heating system for long pulse operation

Radio frequency (RF) power in the ion cyclotron range of frequencies (ICRF) is one of the primary auxiliary heating techniques for Experimental Advanced Superconducting Tokamak (EAST). The ICRF system for EAST has been developed to support long-pulse high-β advanced tokamak fusion physics experiments. The ICRF system is capable of delivering 12 MW 1000-s RF power to the plasma through two antennas. The phasing between current straps of the antennas can be adjusted to optimize the RF power spectrum. The main technical features of the ICRF system are described. Each of the 8 ICRF transmitters has been successfully tested to 1.5 MW for a wide range of frequency (25–70 MHz) on a dummy load. Part of the ICRF system was in operation during the EAST 2012 spring experimental campaign and a maximum power of 800 kW (at 27 MHz) lasting for 30 s has been coupled for long pulse H mode operation.     

Spatially-resolved flat-field soft X-ray spectrometer on experimental advanced superconducting tokamak

A space- and time-resolved flat-field soft X-ray spectrometer with the wavelength range of 1–13 nm has been developed to study impurity behavior on the Experimental Advanced Superconducting Tokamak (EAST). Using an entrance slit, a varied line spacing grating (2400 grooves/mm at the grating center), and a charged coupled device (CCD) system, time evolution of profiles of impurity line emissions were recorded. The spectral resolution of the spectrometer is 0.006 nm at 5 nm when the width of entrance slit is set at 0.03 mm. The best spatial resolution obtained is 24.5 mm with the height of slit at 1.0 mm. The spectrometer is placed 8000 mm away from the plasma center and the observed spatial range covers 0–450 mm from the equatorial plane of EAST. The first experimental results were obtained from the recent EAST campaign. The system was shown to be capable of observing spectral lines from both intrinsic low-Z impurities (C, O, et al.) and highly ionized medium- and high-Z impurities (Fe, Cr, Ni, Cu, et al.). Spectral lines from the full wavelength range (1–13 nm) can be obtained by moving the position of the CCD. Spectra with the wavelength intervals of 1–2 nm show strong metal lines for H-mode discharges. Time evolutions of C VI (3.373 nm) and O VIII (1.897 nm) lines are presented and detail analysis is performed combining electron density intensity, D_α and soft X-ray and extreme ultraviolet (XUV) radiation intensities. Evolutions of profiles of C VI (3.373 nm) and O VIII (1.897 nm) at core plasma were also shown, indicating that the spectrometer can be applied for impurity transport studies,     

Progress in development and design of the neutral beam injector for JT-60SA

Neutral beam (NB) injectors for JT-60 Super Advanced (JT-60SA) have been designed and developed. Twelve positive-ion-based and one negative-ion-based NB injectors are allocated to inject 30 MW D⁰ beams in total for 100 s. Each of the positive-ion-based NB injector is designed to inject 1.7 MW for 100 s at 85 keV. A part of the power supplies and magnetic shield utilized on JT-60U are upgraded and reused on JT-60SA. To realize the negative-ion-based NB injector for JT-60SA where the injection of 500 keV, 10 MW D⁰ beams for 100 s is required, R&Ds of the negative ion source have been carried out. High-energy negative ion beams of 490–500 keV have been successfully produced at a beam current of 1–2.8 A through 20% of the total ion extraction area, by improving voltage holding capability of the ion source. This is the first demonstration of a high-current negative ion acceleration of >1 A to 500 keV. The design of the power supplies and the beamline is also in progress. The procurement of the acceleration power supply starts in 2010.     

A distributed timing and synchronization system for EAST

The distributed timing and synchronization system (DTSS) plays an important role in Experimental Advanced Superconducting Tokamak (EAST), which is one of the national key fusion research facilities in China. This system synchronizes each subsystem of EAST by using reference clock and trigger. A prototype DTSS module has been developed based on PXI bus and RIO (reconfigurable I/O) devices. The DTSS can provide reference clock in frequency up to 80 MHz. The trigger can be pre-defined from 1 ms to 6872 s with 10 ns accuracy. In addition, this system can acquire, process signals, and send output or command to other systems. The DTSS has been successfully applied to 2010 fall EAST experiment, and the results confirmed its accuracy and reliability. After the analysis of system requirement, the architecture of the DTSS and the technical implementation based on PXI are presented in this paper.     

FIRE,a novel concept of massive gas injection for disruption mitigation in ITER: Validation on Tore Supra

Disruptions in large size tokamak like ITER must be mitigated to reduce detrimental effects on the device. Massive impurity injection prior to disruption is a promising mitigation technique. Many injector designs have already been tested on nowadays tokamaks. A novel concept of injector has been designed and tested on Tore Supra. It is based on the use of high pressure cartridges sealed with a bursting disc. For firing, an electric arc is generated inside the cartridge close to the rupture disc. This initiates a shock wave which breaks the disc. Tests show that the opening is achieved in less than 400 μs. A detailed modeling of the fast injector operation is presented. One result is that the gas is expelled in about 2 ms and an outflow rate as large as 4.2 × 10²⁵ atoms/s can be achieved at the nozzle exit (1 cm diameter). Massive neon gas injections have be performed on stable plasma to validate the concept. The gas penetration into the plasma prior to disruption is followed by a fast camera. Comparison with a ferromagnetic-valve based injector was also carried out. It is found that the velocity of the cold front penetration into the hot plasma is larger using this new injector.     

Three-dimensional FE analysis of the thermal–mechanical behaviors in the nuclear fuel rods

In order to implement numerical simulation of the thermal–mechanical behaviors in the nuclear fuel rods, a three-dimensional finite element model is established. The thermal–mechanical behaviors at the initial stage of burnup in both the pellet and the cladding are obtained. Comparison of the obtained numerical results with those from experiments validates the developed finite element model. The effects of the constraint conditions, several operation and structural parameters on the thermal–mechanical performances of the fuel rod are investigated. The research results indicate that: (1) with increasing the heat generation rates from 0.15 to 0.6 W/mm³, the maximum temperature within the pellet increases by 99.3% and the maximum radial displacement at the outer surface of the pellet increases by 94.3%. And the maximum Mises stresses in the cladding all increase; while the maximum values of the first principal stresses within the pellet decrease as a whole; (2) with increasing the heat transfer coefficients between the cladding and the coolant, the internal temperatures reduce and the temperature gradient remains similar; when the heat transfer coefficient is lower than a critical value, the temperature change is sensitive to the heat transfer coefficient. The maximum temperature increases only 7.13% when h changes from 0.5 W/mm² K to 0.01 W/mm² K, while increases up to 54.7% when h decreases from 0.01 W/mm² K to 0.005 W/mm² K; (3) the initial gap sizes between the pellet and the cladding significantly affect the thermal–mechanical behaviors in the fuel rod; when the gap size varies from 0.03 mm to 0.1 mm, the highest temperature in the pellet increases by 19.7%, and the maximum first principal stress at the outer pellet surface decreases by 17.4%; it is critical to optimize the gap size in order to reduce the pellet–cladding mechanical interaction and avoid their contact at early stage. This study lays a foundation for the further research on the irradiation-induced mechanical behaviors in the fuel rods.     

Preliminary design and performance study of EAST liquid lithium limiter based on CPS

Lithium has the ability of H recycling suppression and impurities absorption and it can be used as plasma facing material (PFM) in tokamaks. Lithium conditioning experiments were launched on EAST, HT-7 and some other tokamaks for many years by using the methods of GDC, IRCF and evaporation. Liquid lithium has better performances in effective lifetime and heat removal aspects compared to non-liquid lithium. While, applying liquid lithium in the tokamak would cause the safety problem as the lithium can react with many substances violently and the magnetohydrodynamic behavior is difficult to be handled. EAST liquid lithium limiter (LLL) system is under developing and will be applied in EAST to study the main technologies of the liquid lithium application. The normal operation temperature of the limiter is expected as 230–550 °C under the active cooling of water. Capillary porous system (CPS) is used to prevent the lithium from splashing under large electromagnetic force by increasing the surface tension of the lithium. In order to investigate the cooling performance of the cooling design, the thermal-hydraulic analysis was done which shows that with 3 m/s flowing velocity, the water can keep the limiter under 550 °C all the time if the heat flux is lower than 0.7 MW/m². Under heat flux of 1 MW/m², the limiter should be retreated within 7 s to avoid erosion. The pressure drop of the coolant under 3 m/s is less than 40 kPa with temperature difference nearly 34 °C which meet the design requirements very well. The key manufacture process and technologies like vacuum bonding between the CuCrZr heat sink and 316L guide plate were well studied in the R&D process. The heating test on the test bench showed that the CPS can be heated efficiently by the heaters attached into the heat sink.     

A method to produce lithium pellets for fueling and ELM pacing in NSTX-U

A device for producing small, high frequency spherical droplets or pellets for lithium or other liquid metals has been developed and could aid in the controlled excitation or pacing of edge-localized plasma modes (ELMs). The Liquid Lithium/metal Pellet Injector (LLPI) could also be used to replenish lithium coatings of plasma-facing components (PFCs) during a plasma discharge. With NSTX-U having longer pulse lengths (up to 5 s), it is desirable to be able to inject lithium during the discharge to maintain the beneficial effects. Using a nozzle injector design and a surrogate to lithium, Wood's metal, the LLPI has achieved droplet diameters between 0.6 mm < d_drop < 1 mm in diameter and frequencies up to 1.5 kHz with argon gas driving the formation. This paper presents the LLPI being developed with initial results mainly using Wood's metal and some lithium, using high pressure argon to force the liquid lithium through the nozzle.     

Manufacturing and testing of W/Cu mono-block small scale mock-up for EAST by HIP and HRP technologies

ITER-like W/Cu mono-block plasma-facing components (PFCs) will be used in vertical target regions of the experimental advanced superconducting tokamak (EAST) divertor. The first W/Cu mono-block small scale mock-up with five W mono-blocks has been manufactured successfully by technological combination of hot isostatic pressing (HIP) and hot radial pressing (HRP). The joining of a W mono-block and a pure copper interlayer was achieved by means of HIP technology and the bonding strength was over 150 MPa. The good bonding between the pure copper interlayer and a CuCrZr cooling tube was obtained by means of HRP technology. In order to understand deeply the process of HRP, the stress distribution of the mock-up during HRP process was simulated using ANSYS code. Ultrasonic Nondestructive Testing (NDT) of the W/Cu and Cu/CuCrZr interfaces was performed, showing that excellent bonding of the W/Cu and Cu/CuCrZr interfaces. The thermal cycle fatigue testing of the mock-up has been carried out by means of an e-beam device in Southwest Institute of Physics, Chengdu (SWIP) and the mock-up withstood 1000 cycles of heat loads up to 8.4 MW/m² with the cooling water of 2 m/s, 20 °C, 0.2 MPa.     

Aerosol formation and hydrogen co-deposition by colliding ablation plasma plumes of lithium and lead

In a high-repetition inertial fusion reactor, along with pellet implosions, the interior of target chamber is to be exposed to high-energy, short pulses of X-ray, unburned DT and He ash particles and pellet debris. As a result, wall materials will be subjected to ablation, ejecting particles in the plasma state to collide with each other in the center of volume. The interaction dynamics of ablation plasmas of lithium and lead, candidate first wall materials, has been investigated in the deposited energy density range from 3 to 10 J/cm²/pulse at a repetition rate of 10 Hz, each 6 ns long. The plasma density and electron temperature of colliding ablation plumes have been found to vary from the order of 10⁸–10¹³ 1/cm³ and from 0.7 to 1.5 eV, respectively. The formation of aerosol in the form of droplet has been observed with diameters between 100 nm and 10 μm. Also, hydrogen co-deposition has been found to occur particularly for colliding plumes of lithium, resulting in the H/Li atomic ratio from 0.15 to 0.27 in the hydrogen partial pressure range from 10 to 50 Pa.     

Russian development of enhanced heat flux technologies for ITER first wall

Recently the ITER first wall (FW) design has been significantly upgraded to improve resistance to electromagnetic loads, to facilitate FW panel replacement and to increase FW ability to withstand higher (up to 5 MW/m²) surface heat loads. The latter has made it necessary to re-employ technologies previously developed for the now-abandoned port limiters. These solutions are related to the cooling channel with CuCrZr–SS bimetallic walls and hypervapotron type cooling regime, optimization of Be-tiles dimensions and Be to CuCrZr joining technique. A number of representative mockups were tested at high heat flux (HHF) at the Tsefey electron-beam facility to verify the thermo-hydraulic characteristics of the reference cooling channel design at moderate water flow velocities (V = 1–3 m/s, P = 2–3 MPa, T = 110–170 °C). The heat flux was gradually varied in the range of 1–10 MW/m² until the critical heat flux was registered. The mockups of hypervapotron structure demonstrated the required cooling efficiency and critical heat flux margin (1.4) at a water velocity of ≥2 m/s. Dimensions of Be armor tiles strongly affect the thermo-mechanical stresses both in the CuCrZr cooling wall and at the Be–CuCrZr interface. Results of tile dimensions optimization (variable in the range 12 mm × 12 mm × 6 to 50 mm × 50 mm × 8 mm) obtained by the HHF (variable in the range of 3–8 MW/m²) experiments are presented and compared with analysis. It is shown that optimization of the tile geometry and joining technology provides the required cyclic fatigue lifetime of the reference FW design.     

The influence of neutral beam optimization for DEMO on injector design

The requirements for neutral beam injection (NBI) on DEMO are assessed and the consequences for the design of the injectors discussed. Optimization of current drive requires NBI within a 2 m × 2 m envelope at large tangency radii. This is compatible with beamlines of 20 m length and moderate high voltage stand-off distances between injectors. However, q-profile control will necessitate at least three beamlines of different injector types and may not be compatible with shinethrough. Material irradiation studies show that, with three exceptions, there is no significant design issue for distances greater than 3 m from the tokamak wall.     

RF calculation and thermal-stress analysis of the ECRH launcher on EAST tokamak

EAST is a medium sized superconducting tokamak with major radius R = 1.8 m, minor radius a = 0.45 m, plasma current I_p ≤ 1 MA, toroidal field B_T ≤ 3.5 T and expected plasma pulse length up to 1000 s. An electron cyclotron resonance heating (ECRH) launcher for four-beam injection is being installed on EAST tokamak. Four electron cyclotron wave beams which are generated from four sets of 140 GHz/1 MW/1000 s gyrotrons will be injected into the plasma by the spherical focusing mirrors and plane mobile mirrors. The focusing mirrors are spherical to focus Gaussian beams after reflection. Four plane mobile mirrors independently steer continuously in the poloidal and toroidal direction controlled by motors. With the suitable distance between mirrors and appropriate focal length of focusing mirror, the beam radius in the resonance layer of plasma is 31.145 mm. The heat from plasma radiation and metal losses is loaded on the mobile mirror. In order to decrease the temperature and thermal stress, the inner equivalent diameter of water channels is 8 mm and the suggested water velocity is 4 m/s.     

The ITER full size plasma source device design

In the framework of the strategy for the development and the procurement of the NB systems for ITER, it has been decided to build in Padova a test facility, including two experimental devices: a full size plasma source with low voltage extraction and a full size NB injector at full beam power (1 MV). These two different devices will separately address the main scientific and technological issues of the 17 MW NB injector for ITER. In particular the full size plasma source of negative ions will address the ITER performance requirements in terms of current density and uniformity, limitation of the electron/ion ratio and stationary operation at full current with high reliability and constant performances for the whole operating time up to 1 h. The required negative ion current density to be extracted from the plasma source ranges from 290 A/m² in D₂ (D^?) and 350 A/m² in H₂ (H^?) and these values should be obtained at the lowest admissible neutral pressure in the plasma source volume, nominally at 0.3 Pa. The electron to ion ratio should be limited to less than 1 and the admissible ion inhomogeneity extracted from the grids should be better than 10% on the whole plasma cross-section having a surface exposed to the extraction grid of the order of 1 m².The main design choices will be presented in the paper as well as an overview of the design of the main components and systems.     

Application of HDF5 in long-pulse quasi-steady state data acquisition at high sampling rate

A new high sampling rate quasi-steady state data-acquisition system has been designed for the microwave reflectometry diagnostic of EAST experiments. In order to meet the requirements of long-pulse discharge and high sampling rate, it is designed based on PXI Express technology. A high-performance digitizer National Instruments PXIe-5122 with two synchronous analog input channels in which the maximum sampling rate is 100 MHz has been adopted. Two PXIe-5122 boards at 60 MSPS and one PXIe-6368 board at 2 MSPS are used in the system and the total throughput is about 500 MB/s. To guarantee the large amounts of data being saved continuously in the long-pulse discharge, an external hard-disk data stream enclosure NI HDD-8265 in which the capacity of sustained speed of reading and writing is 700 MB/s. And in RAID-5 mode its storage capacity is 80% of the total.The obtained raw data firstly stream continuously into NI HDD-8265 during the discharge. Then it will be transferred to the data server automatically and converted into HDF5 file format. HDF5 is an open source file format for data storage and management which has been widely used in various fields, and suitable for long term case. The details of the system are described in the paper.     

Model validation of GAMMA code with heat transfer experiment for KO TBM in ITER

By considering the requirements for a DEMO-relevant blanket concept, Korea (KO) has proposed a He cooled molten lithium (HCML) test blanket module (TBM) for testing in ITER. A performance analysis for the thermal–hydraulics and a safety analysis for the KO TBM have been carried out using a commercial CFD code, ANSYS-CFX, and a system code, GAMMA (GAs multicomponent mixture analysis), which was developed by the gas cooled reactor in Korea. To verify the codes, a preliminary study was performed by Lee using a single TBM first wall (FW) mock-up made from the same material as the KO TBM, ferritic martensitic steel, using a 6 MPa nitrogen gas loop. The test was performed at pressures of 1.1, 1.9 and 2.9 MPa, and under various ranges of flow rate from 0.0105 to 0.0407 kg/s with a constant wall temperature condition. In the present study, a thermal–hydraulic test was performed with the newly constructed helium supplying system, in which the design pressure and temperature were 9 MPa and 500 °C, respectively. In the experiment, the same mock-up was used, and the test was performed under the conditions of 3 MPa pressure, 30 °C inlet temperature and 70 m/s helium velocity, which are almost same conditions of the KO TBM FW. One side of the mock-up was heated with a constant heat flux of 0.3–0.5 MW/m² using a graphite heating system, KoHLT-2 (Korea heat load test facility-2). Because the comparison result between CFX 11 and GAMMA showed a difference tendency, the modification of heat transfer correlation included in GAMMA was performed. And the modified GAMMA showed the strong parity with CFX 11 calculation results.     

Mechanical design of the second ICRF antenna for EAST

In order to satisfy the requirements of heating plasma on EAST project, 3 MW ion cyclotron range of frequency (ICRF) heating system will be available at the second stage. Based on this requirement, the second ICRF antenna, has been designed for EAST. The antenna which is planned to operate with a frequency ranging from 30 MHz to 110 MHz, comprises four poloidal current straps. The antenna has many cooling channels inside the current straps, faraday shield and baffle to remove the dissipated RF loss power and incoming plasma heat loads. The antenna is supported via a cantilever support box to the external support structure. Its assembly is plugged in the port and fixed on the support box. External slideway and bellows allow the antenna to be able to move in the radial direction. The key components of the second ICRF antenna has been designed together with structural and thermal analysis presented.     

Manufacturing process and tests of a Lower Hybrid Passive–Active Multijunction launcher for long pulse experiments on Tore-Supra

A new concept of multijunction-type antenna has been developed, the Passive–Active Multijunction, which improves the cooling of the waveguides and the damping of the neutron energy (for ITER) compared to Full Active Multijunction. Due to the complexity of the structures, prototypes of the mode converters and of the Passive–Active-Multijunction launcher were fabricated and tested, in order to validate the different manufacturing processes and the manufacturer's capability to face this challenging project. This paper describes the manufacturing process, the tests of the various prototypes and the construction of the final Passive–Active-Multijunction launcher, which entered into operation in October 2009. It has been commissioned and is fully operational on the Tore-Supra tokamak, since design objectives were reached in March 2010: 2.75 MW – 78 s, power density of 25 MW/m² in active waveguides, steady-state apparent surface temperatures <350 °C; 10 cm long distance coupling.     

A digital long pulse integrator for EAST Tokamak

A digital integrator has been developed to be compatible with the long pulse plasma discharges on the Experimental Advanced Superconductor Tokamak (EAST), in which the induced signal is modulated by a chopper, and a field programmable gate array (FPGA) in the 16-bit digitizer is used to realize the digital integration in real time. After rectification and integration, the drift is almost linear and stable in controlled temperature, so a period of 50 s is used to determine the linear drift rate for drift compensation. The integration data can be directly transferred to the reflective memory (RFM) card, which is installed in the same PCI eXtensions for Instrumentation (PXI) chassis, so the data transmission can be also done in real time. The test results show that the real time data transmission rate is up to 10 kHz, the integration drift is typically less than 0.4 uVs/s and drift performance is a little worse in real long pulse discharge, which can be reduced further by using more precise data acquisition.