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Fusion Engineering (SOFE), 2011 IEEE/NPSS 24th Symposium on

Date 26-30 June 2011

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Displaying Results 1 - 25 of 161
  • ITER disruption mitigation requirements and development of gas cartridge concept

    Publication Year: 2011 , Page(s): 1 - 4
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1114 KB) |  | HTML iconHTML  

    ITER is designed to withstand a certain number of full scale plasma disruptions. Each disruption event induces excessive thermal loads, electromagnetic (EM) loads and runaway electrons onto the vacuum vessel and in-vessel components; the consequences of unmitigated events are extremely serious in terms of component lifetimes. This paper describes the preliminary requirements for the disruption mitigation system (DMS), possible approaches to mitigate the disruption in ITER, the schemes compatible with ITER operation and an on-going pre-conceptual study on gas cartridge DMS, which is one of the massive gas injection DMS schemes. View full abstract»

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  • Behavior of metallic impurity in divertor configuration of Large Helical Device

    Publication Year: 2011 , Page(s): 1 - 6
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    Numerical transport study predicts that the edge surface layer in ergodic layer of Large Helical Device (LHD) has a favorable capability of impurity screening for materials of not only divertor plates but also vacuum vessel. In order to demonstrate the theoretical prediction, the density of iron originating in the LHD vacuum vessel made of stainless steel, which is not covered by carbon plates like tokamaks, is accurately determined with its radial profile using a space-resolved extreme ultraviolet (EUV) spectrometer, of which absolute intensity calibration is done with bremsstrahlung continuum. For the purpose effective intensity coefficients are precisely calculated for iron ions based on a collisional-radiative model. The iron ion density profiles of Fe14+, Fe15+, Fe22+ and Fe23+ are then evaluated with the radial emissivity profile reconstructed from chord-integrated profile and the effective intensity coefficient. The ratio of iron density to electron density integrated over the whole plasma volume can be finally calculated by fitting the iron density profile using one-dimensional impurity transport code. Thus, the analysis on the ratio gives a typical value of 8×10-7 in experimental campaign at last year. The entirely small value of the iron density demonstrates the theoretical prediction. The radial structure of transport coefficients are also obtained from the impurity transport code, showing a large inward convection velocity. View full abstract»

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  • Achievements on engineering and manufacturing of ITER first mirrors mock-ups

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2261 KB) |  | HTML iconHTML  

    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 which implies important research and developments on their design and manufacturing. Therefore investigations on engineering and manufacturing have been carried out on diagnostic mirrors towards the development of full-scale stainless steel and TZM (Mo-based alloy) ITER mirrors. Several micrometers in thickness of rhodium and molybdenum reflective coating layers have been deposited on the components to insure long-lasting of the mirrors exposed to an environment that could be dominated by neutral flux (charge-exchange). Three major issues have been addressed and reported in this paper: First, investigations have been performed on the design and manufacturing of the mirror integrated cooling system, so that the optical surface deformation due to radiations from the plasma and nuclear heating is limited. For the thermo mechanical design of the mock-ups, plasma radiation flux of 0,5 MW/m2 and neutron head load of 7 MW/m3 have been considered. Secondly, the polishing capability of full-scale (109 mm in diameter) metallic mirrors has been demonstrated: the mock ups 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. Thirdly, the manufacturing feasibility of molybdenum and rhodium thick coating layers deposited by magnetron sputtering has been evaluated. The objective of depositing layers up to 3 μm to 5 μm thick has been achieved on the mock-ups, with spectral performances 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, E- - LMs, neutrons, erosion and deposition conditions, is reported. View full abstract»

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  • ITER Radiological and Environmental Monitoring Systems conceptual design elements

    Publication Year: 2011 , Page(s): 1 - 5
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (637 KB) |  | HTML iconHTML  

    A state of the art Radiological and Environmental Monitoring Systems (REMS) is being developed by the ITER Organization, with the engineering support of Fusion for Energy. The primary functions of REMS are to provide health and radiological monitoring for workers as well as area environmental monitoring for the public and thus to assist in the protection against ionizing radiation during ITER operations, including maintenances, and decommissioning. The Radiologically-Controlled Facilities at ITER, to be monitored by REMS are the Tokamak Building, the Tritium Plant Building, the Hot Cell Facility, the Radwaste Facility and the Personnel Access Control Building. Starting from the safety functions assigned to the systems and the requirements imposed by French legislation, embedded in the ITER Preliminary Safety Report, this paper describes the identified REM systems, sub-systems, architectures, safety classification and equipments, and provides the actual status of its design. View full abstract»

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  • Optical transmission of thermal measurements from high voltage devices in high vacuum conditions

    Publication Year: 2011 , Page(s): 1 - 5
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    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. View full abstract»

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  • Advances on the high speed ignitor Pellet Injector (IPI)

    Publication Year: 2011 , Page(s): 1 - 6
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    The control of the density profile during the initial plasma current rise is a critical issue to optimize ohmic and fusion heating rates of Ignitor plasmas. Simulations performed with the NGS ablation model, for the reference ignition plasma parameters (ne0 ≅ ni0 ≅ 1021 m-3, Te0 ≅ Ti0 ≅ 11 keV), indicate that deuterium pellets of a few mm (≤ 4 mm) in size injected at 3-4 km/s from the low field side should ensure adequate deep fuelling. ENEA and ORNL are collaborating on the development of a four barrel, two-stage pneumatic injector for the Ignitor experiment, featuring two innovative concepts: (i) the proper shaping of the propellant pressure pulse to improve pellet acceleration, and (ii) the use of fast closing (~9 ms) valves to eliminate the need of large expansion volumes for propellant gas removal. Two independent sub-systems have been built. The ENEA equipment, including four independent two-stage guns (TSG) and pulse shaping valves, the gas removal system, and the associated controls and diagnostics, has been built and thoroughly tested at CRIOTEC, prior to being shipped to ORNL. The ORNL apparatus consists of the cryostat and pellet diagnostics, with related control and data acquisition system. Integration of the two subsystems (except for the gas removal system) has been readily achieved. The present arrangement accommodates both a TSG and a standard propellant valve on each barrel, allowing seamless switching between standard and high-speed operation on any or all gun barrels. Previous joint experiments at ORNL, demonstrated that the two systems match properly, while their respective control systems interface correctly. The injector performed outstandingly, showing excellent repeatability. These preliminary results indicate that, for the same peak pressure, the IPI has the potential of achieving higher speed performance, as compared with those attained- - by previous high speed injectors, such as the Single Pellet Injector (SPIN) installed on the FTU in the early 90's. Launching sequences at moderate propellant pressure and speeds up to ~2.6 km/s were performed with all four barrels; however, it was not possible to observe intact pellets at speeds above 2 km/s. The analysis of experimental data indicate that, at high speeds, the pellets may spin and hit the wall of the too narrow conduit crossing the diagnostics. This hypothesis is corroborated by numerical simulations. Following this analysis, the inner diameter of this guide tube has been enlarged, and a new target diagnostic has been implemented to test the dispersion of pellet trajectories immediately downstream of the diagnostics. The results of the latest experimental campaign are reported. View full abstract»

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  • A comparative study of different low-z liner materials in an ablation-dominated electrothermal mass accelerator for fusion fueling

    Publication Year: 2011 , Page(s): 1 - 6
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    A low-z ablation-dominated capillary with an ablation-free extension barrel is a concept that provides a plasma flow sufficient to propel fuel pellets into the tokamak fusion plasma chamber. The acceleration barrel is made from a non-ablating material to eliminate mixing the propelling plasma with any impurities evolving from the barrel ablation. A capillary discharge computer code, ETFLOW, has been developed to model plasma flow and acceleration of pellets for fusion fueling in magnetic confinement fusion reactors. The code incorporates a set of governing equations for both the capillary and the acceleration tube and ideal and non-ideal conductivity models. The joule heating in the energy conservation equation is only valid in the capillary section. The pellet momentum and kinetic energy are included in the governing equations of the barrel, with the addition of the effect of viscous drag terms. The capillary generates the plasma from the ablation of low-z liner materials "sleeves" inside the capillary. The acceleration of the pellet starts in the extension tube when the pressure of the plasma flow from the capillary reaches the release limit. The code results show exit velocities in excess of 2km/s for source/barrel systems with low-Z liner materials in the source and loaded with 5, 20, 45, and 80 mg pellets. An increase in the length of both the source and the acceleration barrel increases the pellet exit velocity with the limitation of slowdown effects for plasma expansion and cooling off inside the barrel. View full abstract»

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  • Obtaining data on fast neutron energy spectrum by a method of differential cross-sections in recording recoil nuclei

    Publication Year: 2011 , Page(s): 1 - 5
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    A mathematical approach to processing the equipment spectrum of the detector is offered which is based on the detection of neutron radiation by a method of recoil nuclei. The basis of the approach is the principle of a probabilistic energy distribution of nuclear recoil. This makes it possible to determine the contribution of neutrons of different energies and, also, to evaluate the contribution of every reaction to the equipment spectrum. The mathematical approach is shown in terms of neutron-carbon nuclei interaction. The equipment spectra are calculated for different energies of neutrons and from this the energy spectra of neutrons are obtained. The given approach significantly enhances the potential of the detection of neutrons of any recoil nuclei and makes possible, within the size of the differential cross-section data base, the linear spectrum of neutron radiation to be sorted out. View full abstract»

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  • New IQ demodulator development in diagnostic equipment for fusion energy research

    Publication Year: 2011 , Page(s): 1 - 4
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    New IQ demodulator, which will be used in the reflectometry system, was developed for the Alcator C-Mod experiments. Because the demodulator itself is a 4 mm × 4 mm integrated circuit chip and has very wide operational bandwidth (0.7 GHz - GHz), it can be mounted on a circuit board with its low frequency IQ outputs connecting directly to gain control and dc offset adjustment circuits. This whole circuitry is realized on a 4"×2.5" printed circuit board and can be well tailored to meet the requirements for any type of data collection system. Since the RF and LO ports are operated at very narrow bandwidth in this application, single stub match circuits are designed to improve the ports matching as well as serve as a single pole filter. View full abstract»

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  • Simulation of the infrared views of the upper port VIS/IR imaging system of ITER

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2187 KB) |  | HTML iconHTML  

    In the majority of nuclear fusion experiments equipped with carbonaceous Plasmas Facing Components (PFCs), infrared imaging diagnostics are routinely used for monitoring the surface temperature of the Plasma Facing Components exposed to high heat fluxes. However, future fusion machines, such as ITER, will be equipped with metallic PFCs. As a consequence, due to their low emissivity the evaluation of the surface temperature will have to take into account the multiple reflections of the light coming from hot regions. In order to assess the total infrared (IR) flux collected by the sensor located at the Upper Port VIS/IR System of ITER, realistic simulations of the infrared views of the ITER tokamak have been developed using a Monte Carlo ray-tracing code (SPEOS CAA V5 Based). The simulation includes the thermal emission, the Bremsstrahlung radiation and the reflections thereof inside the real complex 3D geometry (from CATIA) of the vessel. The IR simulated images in conditions of plasma operation are presented and the contribution of the reflected flux for two divertor configurations (CFC and Tungsten) is analyzed. Finally, the achievable measurement accuracy on the surface temperature of the PFCs from the upper port VIS/IR imaging system is discussed. View full abstract»

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  • Structural analysis of an optimally designed spherical tokamak centerpost

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2647 KB) |  | HTML iconHTML  

    The realization of commercialized fusion power will involve the development of new materials that can withstand the uniquely harsh nuclear fusion environment. Of particular interest are those materials that are closest to the plasma. The combination of thermal loading, neutron damage, material sputtering and redeposition provide uniquely hostile conditions under which no material testing has yet occurred. An experimental Fusion Nuclear Science Facility (FNSF) is required that will create the environment that simultaneously achieves high energy neutrons and high ion fluence necessary in order to bridge the gaps from ITER to the realization of a fusion nuclear power plant. One concept for achieving this is a high duty cycle spherical tokamak (ST) [1]. The centerpost is a critical component of the spherical tokamak design, as it controls the size of the entire reactor. The centerpost will experience significant thermal loading and thermal gradients from Ohmic heating, nuclear heating, and water cooling. Nuclear heating will also produce embrittlement and swelling in the centerpost. In addition to thermal loads, the centerpost must be designed to carry mechanical loads produced from the various magnetic fields (TF, PF, plasma currents), both steady-state and transient. The centerpost temperature must remain low enough to permit water cooling, and stresses must remain low enough so that the centerpost remains structurally sound. This study will focus on the stress analysis of a centerpost optimized to reduce the thermal gradients in the cross-section. View full abstract»

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  • Building block support structure for HELIAS Stellarator reactors

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2106 KB) |  | HTML iconHTML  

    An engineering design study is being performed on an enhanced version of a “Helical Advanced Stellarator” (HELIAS) reactor which is based on a fourfold up-scaled W7-X. The coil sizes, their maximal conductor field of 12.3 T, and their operation conditions are similar to those of the ITER TF coils which means that many of the ITER magnet technologies can be applied. The mechanical structure of the first design iteration step has been sub-divided into manageable building block panels for easier manufacture and assembly. The screwed panels between the coils are dismountable which opens up the possibility to separate the torus for maintenance as an alternative to the usually considered exchange of divertor and blanket components through the ports. View full abstract»

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  • A new four-barrel pellet injection system for the TJ-II stellarator

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2529 KB) |  | HTML iconHTML  

    A new pellet injection system for the TJ-II stellarator has been developed/constructed as part of a collaboration between the Oak Ridge National Laboratory (ORNL) and the Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas (CIEMAT). ORNL is providing most of the injector hardware and instrumentation, the pellet diagnostics, and the pellet transport tubes; CIEMAT is responsible for the injector stand/interface to the stellarator, cryogenic refrigerator, vacuum pumps/ballast volumes, gas manifolds, remote operations, plasma diagnostics, and data acquisition. The pellet injector design is an upgraded version of that used for the ORNL injector installed on the Madison Symmetric Torus (MST). It is a four-barrel system equipped with a cryogenic refrigerator for in situ hydrogen pellet formation and a combined mechanical punch/propellant valve system for pellet acceleration (speeds ~100 to 1000 m/s). On TJ-II, it will be used as an active diagnostic and for fueling. To accommodate the plasma experiments planned for TJ-II, pellet sizes significantly smaller than those typically used for the MST application are required. The system will initially be equipped with four different pellet sizes, with the gun barrel bores ranging between ~0.5 to 1.0 mm. The new system is almost complete and is described briefly here, highlighting the new features added since the original MST injector was constructed. Also, the future installation on TJ-II is reviewed. View full abstract»

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  • The development of argon arc brazing with Cu-based filler for ITER thermal anchor attachment

    Publication Year: 2011 , Page(s): 1 - 4
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (942 KB) |  | HTML iconHTML  

    Thermal anchor is the key components of ITER magnet supports to maintain the low temperature for the normal operation of superconducting coils. During the advanced research of ITER thermal anchor attachment, dozens of brazing filler and several kinds of brazing technique have been developed and investigated. Argon arc brazing with Cu-based filler was chosen as the principal method for the attachment of thermal anchor. The brazing temperature of Cu-based filler is over 1000°C,but heat input is relatively low for shallower heating depth of argon arc brazing. Lower heat input is good for the control of brazing deformation. It is no need to clean after brazing because for argon arc brazing there is no brazing flux used. The test result shows that Cu-based alloy have the best mechanical properties at both room temperature and 77K for high brazing temperature. Detail of argon arc brazing with Cu-based filler have been investigated in this article. View full abstract»

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  • A similarity study on absorption/desorption cycles using ZrCo-H2 for ITER hydrogen getter material

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1184 KB) |  | HTML iconHTML  

    Consecutive absorption/desorption cycles of the ZrCo-H2 system were studied to simulate the real ITER hydrogen getter system. ZrCo getter was used in this study instead of the depleted uranium (DU) getter material which was recently considered as the hydrogen getter in ITER. In a cyclic PCI measurement the high-pressure Sievert apparatus seems impractical to describe the equilibrium state of the ZrCo-H2 system in detail, especially for the desorption stage. This high-pressure Pressure-Composition Isotherm (PCI) apparatus, however, shows a cause-and-effect well, from the previous getter state to the following state in presenting hydriding/dehydriding performance. In case of the ZrCo-H2 system or in case of the DU-H2 system, having multiple getter bed battery, a similar affection by the previous getter status might be related and a similar aspect could be shown to should consider further in ITER design, for example a need for control logic, from PCI measurements using a high-pressure Sievert apparatus. View full abstract»

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  • Manifold concept design for ITER Gas Injection System

    Publication Year: 2011 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1281 KB) |  | HTML iconHTML  

    The main functions of ITER Gas Injection System (GIS) are to provide gas fuelling for plasma, wall conditioning operation, and neutral beam injectors. Dedicated manifold, which contains independent tubes for H2/D2, H2, T2, 4He/3He, N2/Ne, Ar and evacuation, is the key part of gas injection lines. It shall deliver gases from the tritium plant to the various fuelling systems. This paper presents an overview of GIS manifold design, especially introduces the solution of penetration structure and routing of manifold from concept design point of view. View full abstract»

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  • Integrated procedure for halo current reconstruction in ITER

    Publication Year: 2011 , Page(s): 1 - 5
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    The paper describes an integrated procedure for the reconstruction of halo currents expected in ITER. The procedure is based on simulation and reconstruction codes, partially adapted from existing codes and partly developed specifically for this purpose. The ultimate goal of this integrated procedure is the assessment of the performance and the possible optimization of the halo current diagnostic system foreseen in ITER. The preliminary results indicate that an alternative distribution of the present layout of halo current sensors could improve the reconstruction capability of this diagnostic system. View full abstract»

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  • R&D on in-vessel dust and tritium management in ITER

    Publication Year: 2011 , Page(s): 1 - 5
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1760 KB) |  | HTML iconHTML  

    In a tokamak, plasma-wall interactions can result in production of dust. During operation, the tritium present in the Vacuum Vessel (VV) can then be trapped in the in-vessel materials but also in dust. The vacuum vessel represents the first confinement barrier to this radioactive material. In the event of a postulated accident involving ingress of steam into the VV, hydrogen could in principle be produced by chemical reaction with hot metal and dust. If the ingress of air into the VV is also postulated, reaction of air with hydrogen and/or dust cannot be completely excluded and could lead to a possible explosion which could challenge the VV tightness. In order to prevent such accidents and their radiological consequences, limitations on the accumulation of dust and tritium in the VV and on the air ingress are imposed. Correlatively, ITER has defined a strategy for the control of in-vessel dust and tritium inventories based on both measurement and removal techniques. In this context, this paper reports on the status of tasks under F4E responsibility aiming at developing some of the measurement systems and necessary R&D for the validation of the ITER strategy. View full abstract»

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  • Structural analysis of a prototype fast shutter for ITER cCXRS diagnostic

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2034 KB) |  | HTML iconHTML  

    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. View full abstract»

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  • Precision signal conditioning and front-end electronics for temperature and field measurements in SST-1 TF magnets

    Publication Year: 2011 , Page(s): 1 - 6
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1451 KB) |  | HTML iconHTML  

    As a part of refurbishment of Steady-state Super-conducting Tokamak-1 (SST-1) at Institute for Plasma Research, India, all sixteen Toroidal Field (TF) magnets of SST-1 have been individually tested in cold with nominal current in a dedicated experimental cryostat. The magnets were cooled down to 4.5 K using either super critical or two phase helium, after which they were charged upto 10 kA of transport current. Precise temperature and field measurements in the experimental configuration were mandatory. Temperatures were required to be measured accurately at several locations in the magnet and hydraulic circuits, where as the field measurements were carried out at few predefined locations on the magnet case. Highly accurate in-house modular signal conditioning electronics had been developed for accurate temperature and field measurements. This paper describes the scheme of the measurement diagnostics, precautions taken for enhanced reliability and long term offset stability in cryogenic environment, and the results of TF magnet tests. View full abstract»

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  • Twin-screw extruder and pellet accelerator integration developments for ITER

    Publication Year: 2011 , Page(s): 1 - 4
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1051 KB) |  | HTML iconHTML  

    The ITER pellet injection system consisting of a twin-screw frozen hydrogen isotope extruder, coupled to a combination solenoid actuated pellet cutter and pneumatic pellet accelerator, is under development at the Oak Ridge National Laboratory. A prototype extruder has been built to produce a continuous solid deuterium extrusion and will be integrated with a secondary section, where pellets are cut, chambered, and launched with a single-stage pneumatic accelerator into the plasma through a guide tube. This integrated pellet injection system is designed to provide 5 mm fueling pellets, injected at a rate up to 10 Hz, or 3 mm edge localized mode (ELM) triggering pellets, injected at higher rates up to 20 Hz. The pellet cutter, chamber mechanism, and the solenoid operated pneumatic valve for the accelerator are optimized to provide pellet velocities between 200-300 m/s to ensure high pellet survivability while traversing the inner wall fueling guide tubes, and outer wall ELM pacing guide tubes. This paper outlines the current twin-screw extruder design, pellet accelerator design, and the integration required for both fueling and ELM pacing pellets. View full abstract»

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  • Investigation of the radioactivity inside the KSTAR vacuum vessel after shutdown by using gamma-ray spectrometry

    Publication Year: 2011 , Page(s): 1 - 5
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (1107 KB) |  | HTML iconHTML  

    Activation products from structural materials inside the Korea Superconducting Tokamak Advanced Research (KSTAR) vacuum vessel at 1 month after shutdown were investigated via gamma-ray spectroscopy using a high-purity germanium (HPGe) detector. To carry out this work, the HPGe detector system was placed on the midplane of port P inside the vacuum vessel, and counts were accumulated for a total time of 220,000 seconds (~2.5 days). Many gamma activities of structural materials with half lives of a minimum of a week were measured. Some of the detected isotopes were from photonuclear reactions via processes such as (γ, n), (γ, 2n), (γ, p), (γ,np), (γ,3He), etc. due to high energy x-rays produced during the operation of KSTAR. Some of them resulted in the same final radioactive isotope. Activation products generated by photonuclear reactions in the structural materials were determined although each reaction could not always be uniquely identified by means of the gamma-ray spectroscopy technique. This investigation has shown that runaway electrons with energies up to ~ 20 MeV were produced in the 2010 KSTAR operation. View full abstract»

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  • New developments for real time plasma control system on HL-2A

    Publication Year: 2011 , Page(s): 1 - 5
    Cited by:  Papers (1)
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (869 KB) |  | HTML iconHTML  

    With more and more complicated codes and control algorithm are being introduced into the plasma control system (PCS) which is a real-time, basic and complex system in HL-2A, it is becoming more and more necessary to reorganize the PCS in order to make it more flexible and effective. Some progress has been made recently for this purpose. A real time operation system that is based on Ubuntu Linux patched by Xenomai code becomes the platform for the plasma control algorithm instead of the DOS operation system at present and the platform for the plasma shape identification code in the future. The real time network consists of reflective memory (RFM) cards take the place of the signal cables for data transmitting in real time plasma control. The real time data acquisition system for plasma control is separated as a single crate where NI RT system deployed. The details of the new developments for HL-2A are described in this paper. View full abstract»

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  • Progress on the integration of ITER diagnostics equatorial port plugs in Europe

    Publication Year: 2011 , Page(s): 1 - 6
    Save to Project icon | Request Permissions | Click to expandQuick Abstract | PDF file iconPDF (2327 KB) |  | HTML iconHTML  

    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. View full abstract»

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  • Temporal and spatial PFC temperature profiles in KSTAR 2010

    Publication Year: 2011 , Page(s): 1 - 4
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    In- and outboard Plasma Facing Components (PFCs) of KSTAR (Korea Superconducting Tokamak Advanced Reasearch) have been fully installed in 2010 for D shaped diverted plasmas. Before the start of plasma operation, the PFCs were baked up to 200 °C by hot nitrogen gas circulation system to remove impurities including water. The surface temperature of the PFC tiles was monitored by 200 thermo-couple sensors during the plasma operation (plasma shot or Glow Discharge Cleaning(GDC)), and the temporal and spatial (poloidal) temperature profiles are obtained. Depending on the heat flux on each tile, the surface temperature shows time-dependency. After 1-hour morning He GDC, the temperature of the PFCs at inboard side has reached at 40 °C. After an H-mode shot, the temperature of divertor tiles around the striking points was substantially increased. The time-averaged total heat flux after an specific H-mode in 2010 was estimated to be approximately 10kW/m2. View full abstract»

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