Mirrors inside the ITER torus vacuum are used in many of the ITER port based diagnostic systems. As example, the core Charge Exchange Recombination Spectroscopy (core CXRS) concept is based on a mirror labyrinth of 7 consecutive in-vessel mirrors. For this system, ex-vessel alignment of up to three mirrors might be needed. Because of the unique and demanding environment in the ITER diagnostic port plugs, especially in the first meter from the plasma, existing designs of adjustable mirror holders cannot be used without modification. The wish to allow re-alignment of the mirrors during maintenance of the diagnostic system, if possible by remote-handling only, adds to the complexity. With many of the ITER diagnostics being in the stage of a conceptual design and advancing towards a more detailed design, the mechanical design of alignable mirrors themselves is starting. This paper discusses mirror alignment principles for use by ITER diagnostics. The core CXRS system is used as example throughout the paper. Loads and requirements for in-vessel diagnostic mirrors are discussed. Three basic groups of alignment principles are assessed for applicability in close proximity to the ITER plasma. The strength and weaknesses of the different alignment principles are discussed and a conclusion about their general applicability in ITER is given.
The acceleration grid power supply of the ITER neutral beam injector is designed to deliver up to 200 kV to each of the five stages of the electrostatic accelerator, producing a total voltage up to -1 MV with respect to ground. In the present design, the output voltages of the dc generators are independently controlled by five closed control loops. The output voltage measurements are obtained from the difference between the grid potential measurements, referred to local ground. Therefore, the control accuracy for each stage is affected by the amplification of the measurement errors. Thus, an alternative control scheme, based on a single closed control loop and regulator, has been considered. In this paper, the performance of the two control methods in terms of systematic errors, dynamic performance, and sensitivity to load variations is evaluated and compared.
An electron dump (ED) has been developed and added to the SPIDER negative ion beam source in order to dispose of the electrons exiting from the accelerator, which are mainly generated by stripping reactions. This component prevents from having unwanted and possibly concentrated heat load from electrons on the downstream components and permits to obtain on the diagnostic calorimeter a detailed measurement of the negative ion beam. The detail design of the electron dump is here described, together with the main design choices that have driven the design process. The physics and thermo-mechanical analyses, carried out to compare the different design proposals, are also presented.
This paper describes the detailed design of the Quench Protection Circuits (QPC) for the superconducting Toroidal Field (TF) and Poloidal Field (PF) magnets of the Satellite Tokamak JT-60SA, which will be installed in Naka, Japan [1]. The nominal currents to be interrupted and the maximum reapplied voltages are 25.7 kA and 2.8 kV for the TF QPCs and 20 kA and 5 kV for PF QPCs. The innovative solution proposed in the QPC design is based on a Hybrid Circuit Breaker (CB) composed of a mechanical Bypass Switch for conducting the continuous current, in parallel to a static CB for current interruption. The main choices of the final design are presented and discussed, either to confirm or to update and complete the study performed at the conceptual design level.
With the development of heterogeneous camera networks working at different wavelengths and frame rates and covering a large surface of a vacuum vessel, the visual observation of a large variety of plasma and thermal phenomena (e.g., hot spots, ELMs, MARFE, arcs, dusts, etc.) becomes possible. In the domain of machine protection, a phenomenological diagnostic is a key element toward plasma/thermal event dangerousness assessment during real-time operation. It is also of primary importance to automate the extraction and the storage of phenomena information for further offline event retrieval and analysis, thus leading to a better use of massive image databases for plasma physics studies. To this end, efforts have been devoted to the development of image processing algorithms dedicated to the recognition of specific events. However, a need arises now for the integration of techniques developed so far in both hardware and software directions. We present in this paper our latest results in the field of real-time phenomenon recognition and management through our image understanding software platform. This platform has been validated on Tore Supra during operation and is under evaluation for other present Tokamaks and for the foreseen imaging diagnostic of ITER.
Diagnostics in ITER are supported by big structures called port plugs, the second main function of which is to ensure a sufficient shielding against neutrons and gammas. Regarding the integration of diagnostics in equatorial port plugs, a new approach is under study, which consists in installing the diagnostics in “drawers”. This paper describes the recent work which has been performed in Europe on the integration of diagnostics in drawers in the Equatorial Port Plug 1 (EPP1). First the methodology which has been followed to progress on the integration of the diagnostics in this port plug is described and the resulting arrangement of diagnostics is shown. Then a special attention is paid to the integration of the two main diagnostics of EPP1, namely the visible/infrared wide angle viewing system and the radial neutron camera. Finally the preliminary design of the drawers of EPP1, in particular the shielding modules around the diagnostics, is presented, and the preliminary results of the analyses performed to validate this design are provided.
Inductively coupled radio-frequency (RF)-based multidriver negative ion sources form the basis of the ion sources for neutral beam injectors considered for present-day fusion devices. Information on the operational behavior of such multidriver RF sources is limited. Thus, an Indian program on the development and operation of such multidriver sources has been initiated to support and enhance the existing information and thus contribute to the successful development of the large-size multidriver ion sources. Experiments initiated under this program have the objective of understanding the physics and technology of plasma production and its control in the presence of the multidriver coupling. An experimental system consisting of a two-driver-based source powered by a single 1-MHz 180-kW RF generator has been configured for the purpose. It is foreseen that, in the proposed configuration of the source, a plasma of density on the order of 10 18 m -3 in a volume about 0.5-m 3 chamber shall be produced, and under cesiated conditions (Cs vapor injected into the source), a negative hydrogen ion current on the order of 10-12 A can be extracted at 50 kV. Adequate flexibility has been incorporated in the system design to test the source performance in air or under vacuum. The experiment is supported by extensive optical-, thermocouple-, and probe-based diagnostics to provide information related to the essential plasma parameters such as uniformity, temperature, density, and impurity level.
The Helium Refrigerator System (HRS) of the Korea Superconducting Tokamak Advanced Research (KSTAR) which was designed to provide an energetic equivalent cooling power of 9 kW at 4.5 K, has been operated successfully during last three years since 2008. For the operation of the KSTAR, the all of the cold components, such as the superconducting (SC) magnets, magnet structure, thermal shields (TS), SC buslines (BL), and current leads (CL) were cooled down to the aimed cryogenic temperature by generated helium in the HRS. The supercritical helium (4.5 K and 5.5 bar) of 300 g/s was delivered for cooling of the SC magnet (PF magnet: 14, TF magnet: 16) and magnet structure and was circulated by the Poloidal Field (PF) and the Toroidal Field (TF) circulators. Thermal loads due to magnet operation are removed at the thermal damper (TD) installed in the distribution box of the HRS. And the pressure change in the magnet will directly affect to the circulators. In particular, pressure drop of the PF circulator drastically changed due to the PF magnet operation depending on the plasma generation scenario. In the 2010 campaign, the operating range of the PF magnet expanded from ± 4 kA to ± 10 kA and the plasma scenario was improved. Accordingly, the mass flow rate of the PF circuit was increased from 350 g/s to 369 g/s in order to achieve cryogenic stability. But the operating range of PF circulator exceeded the allowable range during shot #2526 shot and shot #3878. Nevertheless, the circulator successfully operated without any significant fault. Operating plan of the PF magnet that will be further expanded in the operating range may cause damages to the circulator continuously. Consequently, the influence of PF operating scenarios on the PF circulator was analyzed to investigate safety margin of the HRS.
Korea (KO) has developed a Liquid breeder blanket and participated in the Test Blanket Module (TBM) program within the International Thermonuclear Experimental Reactor (ITER) group with a Helium Cooled Molten Lithium (HCML) concept. Based on this concept, helium (He) and liquid lithium (Li) were used as a coolant and breeder, respectively. Additionally, ferritic martensitic (FM) steel was considered as a structural material. However, according to our strategy for developing a liquid breeder TBM and its more relevant DEMO concept, not only liquid lithium breeders, but also lead-lithium (PbLi) breeders were considered. An Experimental Loop for a Liquid breeder (ELLI) was constructed for the purpose of validating the design and fabrication of our electromagnetic (EM) pump; testing the effects of the magneto-hydro-dynamics (MHD); and investigating the compatibility of PbLi using structural materials such as ferritic martensitic steel. In the ELLI, Pb-15.7Li, where Li is 15.7 at % (called PbLi hereafter), is used as the breeding material, and the EM pump circulates it through the loop. The maximum operating pressure and temperature in the loop are 0.5 MPa and 550 °C, respectively. In this study, performance tests with the EM pump were carried out. During the three separate experiments, the EM pump was operated for 250 h with a speed of 0.16 m/s for corrosion tests. For a material of corrosion test, tubular-type specimens and cylindrical-type specimens were fabricated and installed in three test pots of the loop. After installing the specimens into the loop, the corrosion test was performed while the EM pump was operating with a 0.16 m/s flow rate at 340 °C for 250 h.
Steady State Tokamak (SST-1) is currently being refurbished in a mission mode at the Institute for Plasma Research with an ultimate objective of producing the first plasma in early 2012. Since Jan 2009, under the SST-1 Mission mandate, a broad spectrum of refurbishment activities have been initiated and pursued on several subsystems of SST-1. Developing sub nano-ohm leak tight joints in the magnet winding packs, developing single phased LN2 cooled thermal shields, developing supercritical helium cooled 5 K thermal shields for magnet cases, insurance of thermal and electrical isolations between various sub-systems of SST-1, testing of each of the SST-1 Toroidal Field (TF) magnets in cold with nominal currents, testing each of the modules and octants of SST-1 machine shell in representative experimentally simulated scenarios, augmentation and reliability establishment of the SST-1 vacuum vessel baking system, time synchronizations amongst various heterogeneous subsystems of SST-1, large data storage scenarios, integrated engineering testing of the first phase of the plasma diagnostics etc are some of the major refurbishment activities. Presently, the SST-1 device integration is in full swing. The cold test of the assembled SST-1 TF and PF magnets are due to begin from Dec 2011. Following the successful testing of the SST-1 superconducting magnet system and engineering validations of the machine shell, the first plasmas will be attempted in SST-1. The first plasma will be ∼ 100 kA limiter assisted with the available volt-sec and could possibly be assisted by ECCD/LHCD.
The Korean Superconducting Tokamak Advanced Research (KSTAR) device is aimed at advanced tokamak (AT) research. Three years have passed since it achieved its first plasma in 2008. Because it is a superconducting machine and is pursuing AT research, it has unique features in terms of the machine engineering and operation. The toroidal field (TF) magnet coils are made of Nb3Sn, which provide high toroidal fields up to 3.5 T, and have been fully tested. The poloidal field (PF) magnet coils, consisting of both Nb3Sn and NbTi, which have a maximum current of 25 kA in their design, were tested up to 15 kA. A thermal hydraulic analysis is being conducted for PF magnet coil operation. All plasma facing components (PFCs) are equipped with water cooled graphite tiles and have the capability of being baked up to 350 °C. A startup scenario, which considered both the effect of the ferromagnetic material in the cable in conduit conductor (CICC) jacket in the magnet coils as well as a non-ferromagnetic up-down asymmetry in the cryostat structure, was developed and demonstrated its effectiveness by the last two year's reliable operations. Passive stabilizers and In-Vessel Control Coils (IVCC) are key components to realize AT Operation in KSTAR. The segmented IVCC coils were connected to form circular coils for internal vertical control in 2010 and diverted plasmas with high elongation (κ∼1.8, δ>0.6) were achieved. A neutral beam injection (NBI) system was developed aiming at 2 MW, 300 s per ion source which meets the long-pulse requirement of KSTAR. An NBI ion source with a power of 1.7 MW at 100 kV has been commissioned. Finally, ELMy H-modes were successfully produced with 1.3 MW NBI power at a plasma current of 0.6 MA in the 2010 campaign. The first H-mode discharge (#4200) in KSTAR was achieved one year earlier than officially planned and it was done at BT=2.0 T with Ip=0.6 MA in a well-balanced double null configuration after boronization on the PFC. Successful operations in the early days of KSTAR including H-mode experiments revealed the capability of advanced and steady-state operation which is essential for the International Thermonuclear Experimental Reactor (ITER) and future fusion reactors
Electrical edge joint of a stacked high-temperature superconducting (HTS) conductor is explored for electrical joints for a demountable HTS toroidal field coil. Demountable coils would be very useful in small tokamak plasma-facing component-test machines, such as Vulcan. The demountable concept consists of forcing together the edges of stacked HTS conductor embedded in a conductive or a structural material. Numerical evaluation showed that inserting an indium film between joint surfaces or plating copper layer on joint surfaces could prevent joint resistance from increasing due to misalignment of the contact surface and would provide a compliant layer between joints. Joint resistance of the edge joint can become smaller than that of electrical lap joint when the number of the stacked HTS tape is large. We carried out a testing program of the edge joint of a stacked YBCO conductor within a copper jacket. The HTS cable has a critical current of 1600 A at 77 K in self field. The experimental results showed that joint resistance in the edge joint was higher than expectation, as was irreproducible. One potential reason for the degradation of HTS material is the damage to the tape edge in preparation of the joint region due to milling. Another potential reason was low accuracy of fabrication of joint surface, the joint surface were not parallel to another one. The joint resistance will be reduced by sophistication of fabrication process for the joint surface.
The paper describes a system for digitization and optical transmission of thermal measurements on high voltage devices in high vacuum environmental conditions and presents the tests conducted on a prototype. The system has been designed in particular to satisfy such technical requirements as to be mounted on the grounded grid of the SPIDER facility (a 100keV/60A particle accelerator) and to withstand frequent fault conditions in which the voltage of the grid transiently rises up to some tens of kV. The system is based on a circuit which samples and transmits the signals to the central acquisition system while preserving the signals and avoiding any electrical links between the high voltage device and the vacuum vessel. Moreover the system has to be designed so as to minimize the electromagnetic noise affecting the low amplitude signals from the thermocouples. The circuit design is presented, describing the layout and the electronic components for the acquisition of the thermocouple signals and for the data transmission via optical fiber. When SPIDER is operational, with up to one hour pulse duration, the circuit is powered by a battery, which is in turn recharged by the energy coming from a photovoltaic cell when SPIDER is not operational and the circuit is not acquiring. Data are digitally transmitted according to RS-232 protocol for easy interfacing to the central data acquisition system. The circuit has been tested to check its proper operation, with particular care devoted to the data transmission and the recharging phase. The results are reported and discussed.
For 60 years fusion research has been focused on Fusion for Energy (F4E) as the ultimate carbon-free solution to the world's energy problems. It is proving a worthy but difficult task. However it is relatively easy to produce high-energy fusion neutrons. The many potential applications of a 14MeV neutron source are outlined, and a range of existing designs for such a source, based on a D-T fuelled Spherical Tokamak (ST), are reviewed. It is shown that the problems of high build and operating costs and uncertainties in operating conditions, can be eased by a small device SCFNS (Super-Compact Fusion Neutron Source) of major radius ∼0.5m, which although operating at modest plasma performance can provide megawatt-level neutron output. This break-through is achieved via the effectiveness of beam-plasma fusion, which becomes dominant in these conditions. Such a device would provide a resolution of the uncertainties in fusion STs (such as start-up, ramp-up, and steady-state operation); be an effective neutron source for research, and be an ideal entry vehicle for development of more powerful neutron sources in the new objective of ‘Fusion for Neutrons’ (F4N).
Most of ITER optical diagnostics will be equipped with in-vessel metallic mirrors as plasma viewing components. These mirrors will be exposed to severe plasma environment and must withstand these conditions without change of their optical properties. This implies important research and developments on the design and manufacturing of such components. Therefore, investigations on engineering and manufacturing have been carried out on diagnostic mirrors toward the development of full-scale stainless steel and TZM (Mo-based alloy) ITER mirrors. Several-micrometer coatings of rhodium and molybdenum have been deposited on the components to ensure long-lasting of the mirrors exposed to an environment which could be dominated by charge-exchange neutral flux. Three major issues have been addressed and reported in this paper: First, investigations have been performed on the design and manufacturing of the integrated cooling system to limit the mirror optical surface deformation due to radiations from the plasma and nuclear heating. For the thermomechanical design of the mock-ups, a plasma radiation flux of 0.5 MW/m 2 and a neutron head load of 7 MW/m 3 have been considered. Second, the polishing capability of full-scale (109 mm in diameter) metallic mirrors has been demonstrated: The mock-up surface front error is lower than 0.1 μm root mean square, and the mirrors exhibit low roughness ( Ra <; 2 nm) and low surface defects (scratch width lower than 0.02 mm) after polishing. Third, the manufacturing feasibility of thick molybdenum and rhodium coating layers deposited by magnetron sputtering has been evaluated. The objective of depositing layers up to 3-5 μm thick has been achieved on the mock-ups, with spectral reflectance reaching the theoretical values and showing high reflectivity over a large spectral range (from 400 nm to 11 μm). Finally, the test campaign of the manufactured mirrors, which is being prepared in several European facilities to expose the mirrors to deuterium plasma, ELMs, neutrons, erosion, and deposition conditions, is reported.
Experiments on overdense plasma heating through the mode-coupling scheme known as “O-X-B Double Mode Conversion”, obtained launching a narrow beam of millimeter-waves at 140 GHz frequency and 400 kW power, are scheduled for the next experimental campaigns of the FTU tokamak. Such a scheme, not yet demonstrated at electron density higher than the critical one (2.4·1020m-3) for the 140 GHz ordinary mode, and consequently at such a high frequency, exploits the conversion of an ordinary polarized wave (O) into the extraordinary (X) one, followed by a subsequent conversion to Bernstein (B) waves, which are then absorbed by the plasma. In the specific case of FTU the overall efficiency of this scheme is mainly determined by the coupling efficiency between the O- and the X-wave, which can occur only for ordinary polarized radiation propagating in a very narrow angular range at the cutoff region. The simulations performed with a single ray tracing, show that the required precision in the injection of the wave into the plasma is very high and an angular deviations of ±1° with respect to the optimal injection, in either vertical or horizontal direction, implies a 50% drop in the power transmitted to X-mode. Moreover, the application of models able to take into account the real shape of the incident beam, show that the maximum reachable efficiency, under optimal wave injection, is expected not to exceed 45% of the EC power injected. The new ECH&CD launcher, now being installed in FTU, will be able to provide the angular precision required for the steering. The basic idea of a control algorithm, aimed to track in real-time the optimal angular window for the wave injection in experiments on O-X-B mode conversion, is presented in the paper. The control will use the stray gyrotron radiation as observable, which is detected by a set of sniffer probes located at different toroidal positions in the FTU vessel.
Optical lifetime of the first mirror is a critical issue for the ITER upper port plug core charge exchange spectroscopy diagnostic (cCXRS). A fast shutter is engaged to protect the mirror from depositions between measurements. The prototype shutter will be examined in a test vacuum vessel that is now under development in the Forschungszentrum Jülich, Germany. Being located near the plasma, the shutter operates under severe thermal and electromagnetic (EM) loads. The multi-field analyses conducted for the shutter are presented in the paper. Since the fast shutter can operate within 1 second, its static structural analysis should be accompanied by dynamic studies. The paper pays attention to numerical strategy used for a multi-field ANSYS modeling of a complex structure. The shutter structural performance under the service, thermal and EM loading is in line with requirements. Solution for a problem of high local thermo-stresses revealed by the analysis is proposed. Problems connected with other possible port plug - shutter layouts are discussed.
Wendelstein 7-X (W7-X) will demonstrate the possibility of a stellarator for a future fusion power plant. This stellarator fusion experiment is at present in the assembly phase at the Max-Planck-Institut für Plasmaphysik (IPP). The main advance of the static plasma is caused by the three dimensional shape of the coils. But inside the Cryostat this extravagant geometry of the coils efforts also a three dimensional contour of the main mechanical components. One of the ambitious challenges is how to build up such complex machine. The manufacturing of these complex devices have been demanded the newest manufacturing methods. At 2014 Wendelstein 7-X will be the world's largest superconducting helical advanced stellarator. The toroidal plasma vessel geometry follows exactly the three dimensional shape of the plasma. It contains the plasma with a great diameter of 11m and an average plasma diameter of 1.1 m. To control the plasma geometry it is necessary that all the 20 planar and 50 non planar coils are not only extreme narrow positioned to the Plasma Vessel but also within a tolerance of 1.5 mm to each other. To meet this requirement and to withstand the high magnetic forces a complex coil support structure was created. The Central Support Ring have to bear the coils but the different inter coil supports canalize the forces by very stiff connections on one side and sliding areas on the other side. The coils and the support structure are enclosed within the Outer Vessel with its domes and openings. The Outer Vessel, the Plasma Vessel and the ports generate the boundaries for the Cryostat. The vacuum inside provides thermal insulation of the magnet system which is cooled down to 4 K. The 254 ports secure the access to the Plasma Vessel with all the supply lines and the diagnostics. Due to the different thermal movements the Plasma Vessel, Outer Vessel and the Central Support Ring have to be supported separately. The Central Support Ring is held by 10 cryo legs. The Plasma Vessel supporting system is divided into two separate systems, allowing horizontal and vertical adjustments to centre the Plasma Vessel during thermal expansion. Beside an overview about the main components in the cryostat like the plasma vessel, the outer vessel, the ports and the different support systems this paper describes the most demanding manufacturing methods. The author delineates some disparate and special problems during the manufacturing of the components at the companies in the different European countries.
Detailed profiles of nuclear heating and radiation damage parameters were determined in ITER blanket modules at different poloidal locations. The results indicate that the nuclear parameters are sensitive to the configuration and material composition with enhanced steel heating and helium production in regions with large water content.
Neutronics analysis for a tokamak reactor GNOME was studied by a three dimensional Monte Calro simulation code; MCNP-5. Distribution of nuclear heat was obtained from the calculated and applied into in-board cooling duct designs for the GNOME reactor. The pressure drop at the duct and corresponding pumping power were evaluated to survey the operational windows of the tokamak with SiC insulated He/PbLi dual coolants blankets. Besides, the pressure drop at the SiC structure PbLi cooled divertor was also evaluated in order to confirm the feasibility of the PbLi cooled divertor.
