Search

• A mechanical platform accommodating ITER in-vessel magnetic sensors has been designed and prototyped. • The platform enables electrical connection of the magnetic sensor with cabling. • The platform is compact and remote handling compatible. • Manufacturability, assemblability and end-end electrical connectivity were successfully demonstrated. A mechanical support structure (a.k.a. "platform") has been designed to provide mechanical support and thermal conductance for the inductive magnetic sensors installed on the inner shell of ITER vacuum vessel (VV) for equilibrium and high-frequency magnetic field measurement. The platform design is modular so as to simplify the on-site installation process. It consists of a permanent and a removable section. The permanent section is welded to the VV surface, forming the base of the whole assembly and allowing electrical connection with the in-vessel electrical service. The removable section houses the sensor head, which is engaged with the permanent base through a knife-switch type connector. Alignment of the two sections is realized via two alignment pins built into the permanent base. The removable section is designed to be compatible with remote handling operation in case of replacement. Real-scale prototypes of the platform components had been successfully manufactured and assembled together according to the design. Good alignment between the two sections has been achieved. Full electrical continuity from the sensor head down to the electrical service cabling has been demonstrated in the platform prototypes. Continued R&D efforts will focus on testing some installation and operation aspects, including mechanical vibration, thermal cycling, cable clamping, etc. [ABSTRACT FROM AUTHOR]

A recent improvement of the design of the ITER outer vessel steady-state magnetic field sensors (OVSS) is presented. OVSS features a pair of Hall sensors with a sensing layer made of bismuth. Sixty OVSS will be installed on the ITER vacuum vessel outer skin to perform an absolute measurement of magnetic field. The bismuth Hall sensor measurement is temperature dependent, and the maximum allowed OVSS error of 4-mT limits the error of the measurement of the Hall sensor temperature by an on-board thermocouple to be less than 0.28 °C. This error consists of the temperature difference between the thermocouple and Hall sensors, thermocouple calibration error, and thermocouple measurement error. The improved design of the sensor housing reduces the temperature difference between the thermocouple and the Hall sensors bellow 0.04 °C. Procedures of the calibration prior installation and in situ calibration will allow calibrating the on-board thermocouple with the error less than 0.06 °C. The thermocouple measurement error limit was determined as 0.27 °C in which case OVSS meets the maximum allowed error. [ABSTRACT FROM AUTHOR]

High-frequency (HF) inductive magnetic sensors are the primary ITER diagnostic set for Toroidal Alfvén Eigenmodes (TAE) detection, while they also supplement low-frequency MHD and plasma equilibrium measurements. These sensors will be installed on the inner surface of ITER vacuum vessel, operated in a harsh environment with considerable neutron/nuclear radiation and high thermal load. Essential components of the HF sensor system, including inductive coil, electron cyclotron heating (ECH) shield, electrical cabling and termination load, have been designed to meet ITER measurement requirements. System performance (e.g. frequency response, thermal conduction) has been assessed. A prototyping campaign was initiated to demonstrate the manufacturability of the designed components. Prototypes have been produced according to the specifications. A series of lab tests have been performed to examine assembly issues and validate electrical and thermo-mechanical aspects of the design. In-situ microwave radiation test has been conducted in the MISTRAL test facility at IPP-Greifswald to experimentally examine the microwave shielding efficiency and structural integrity of the ECH shield. Low-power microwave attenuation measurement and scanning electron microscopic inspection were conducted to probe and examine the quality of the metal coating on the ECH shield. [ABSTRACT FROM AUTHOR]

This paper presents the multiple set of requirements for the ITER magnetic diagnostic systems and the current status of the various R&D activities performed by the EU partners. [ABSTRACT FROM AUTHOR]

Abstract: The technological challenges associated with preparing and installing diagnostic systems on ITER are summarized and the solutions that have been developed to overcome them are described. Areas where further developments are needed are highlighted. [Copyright &y& Elsevier]

• The laser-induced fluorescence diagnostic system for measuring the ITER divertor plasma parameters is described. • The LIF spectroscopic schemes and lasers are chosen for measurements of HeII density and temperature. • The LIF and background signals and the relative accuracies are calculated for both He and DT discharges of ITER. Laser-induced fluorescence (LIF) on ITER will be used for local measurement of helium density n He and ion temperature T i in divertor. The LIF diagnostic is combined with the divertor Thomson scattering (DTS) via common laser injection and signal collection optics. Collisional-radiative model developed for helium ions is used to estimate laser parameters required for robust measurement of laser-induced signals against plasma background radiation. The estimations demonstrate feasibility of measuring n HeII with laser parameters provided by available optical parametric oscillator pumped by Nd:YAG. New approach for measuring T i is proposed and the expected errors are estimated using numerical simulations. [ABSTRACT FROM AUTHOR]

ITER in-vessel magnetic sensors play a key role for ITER plasma operation. Each sensor is accommodated in a platform mounted on the inner surface of ITER vacuum vessel and behind the blankets. A full set of engineering analyses has been performed on the platform to assess the feasibility of the design configuration. Electromagnetic Sub-Modelling technique has been used for very detailed evaluation of the EM loads due to 8 category III plasma disruption events acting on platforms at 23 locations from a full poloidal array. The two cases yielding highest EM load were selected to complete the structural assessment. Thermal analysis was performed for a platform located where maximum heat load is expected and the structural assessment was performed considering a pessimistic input load combination of highest EM loads and highest thermal stresses over ITER operation lifetime. An optimization process to reduce the EM loads and increase the structural strength of the platform was set up and it led to positive results on the structural integrity of the platform structure with the optimized design. Future work is planned to incorporate the thermal deformation of VV surface and refine the calculation. [ABSTRACT FROM AUTHOR]

• The use of diode pumping in 2 J × 50 Hz 3 ns Nd:YAG laser improve beam quality. • Cavity dumping approach in 946 nm Nd:YAG laser allow to obtain 3 ns pulse duration. • Zero diffraction length Regenerative Amplifier allow to reach 1 J in 946 nm Nd:YAG. • Image-rotation in Regenerative Power Amplifier significantly improves beam quality. We developed high energy Q-switched 1064 nm and 946 nm Nd:YAG lasers for ITER Divertor Thomson Scattering (DTS). The 1064 nm Nd:YAG laser is designed to be used as the main DTS probing laser and features output energy of 2 J at 3 ns pulse duration. MOPA configuration with phase conjugation mirror provides divergence of 1.3 times the diffraction limit and high stability of the output beam parameters with repetition rate varied from 1 to 50 Hz. The 946 nm Nd:YAG laser designed for calibration by an auxiliary Thomson spectrum at the neighboring wavelength. The 946 nm laser consists of MO based on cavity dumping, high gain regenerative preamplifier and zero diffraction length ring cavity regenerative power amplifier with image-rotation providing 1 J × 50 Hz at 3 ns pulse duration with divergence of 1.5 the diffraction limit. As far as we know this is the highest value of the output energy obtained at 946 nm with nanosecond pulse duration until now. Both lasers use unified pump modules providing high maintainability of the system. The paper describes laser design approaches, state of the art and outlook of further developments. [ABSTRACT FROM AUTHOR]

Laser-induced fluorescence diagnostics on ITER will be used for local measurements of helium density ( n HeI ) and ion temperature ( T i ) in the divertor area. The diagnostics is combined with divertor Thomson scattering via common laser injection and signal collection optics. Multichannel filter polychromators with design similar to those developed for Thomson scattering will be used to perform spectral analysis of the fluorescent signals. Both the fluorescent signals and the background radiation are estimated using the collisional-radiative models developed for He I and He II. The estimations demonstrate the feasibility of measuring n HeI with the temporal resolution 20 ms and accuracy Δ n HeI < 20% in both the He and DT regimes. The possibilities of different approaches for the T i measurements are discussed. [ABSTRACT FROM AUTHOR]

The outer vessel steady-state sensors (OVSS) – a subsystem of the ITER magnetic diagnostics – will contribute to the measurement of the plasma current, plasma-wall clearance, and local perturbations of the magnetic flux surfaces near the wall. The diagnostic consists of a poloidal array of sixty sensors welded to the vacuum vessel outer surface. Each OVSS contains a pair of bismuth Hall sensors with the measurement axes parallel (in the poloidal plane) and normal to the vacuum vessel. These Hall sensors were successfully tested in environmental conditions corresponding to those in ITER (neutron irradiation, vacuum vessel bakeout, etc.) and have proven to be capable of measuring the magnetic field up to at least ±7 T. An AC analogue lock-in current spinning signal processing electronics has been developed for the OVSS to eliminate the spurious voltage due to Hall sensor manufacturing imperfections and the planar Hall effect. This paper presents the final design of the OVSS diagnostic, developed jointly by the ITER Organization and the Institute of Plasma Physics in Prague. [ABSTRACT FROM AUTHOR]

The ITER Plasma Control System (PCS) approaching the second phase of development is now more deeply investigating alternative solutions for the various controls aimed at operations up to 15MA with low auxiliary heating in L-mode. The control functions in the PCS are strictly linked to the performance of the ITER actuators and diagnostics. The capabilities of those systems need to be carefully validated against the control needs. System and performance requirements shall be consistent with the control schemas and where limitations or restrictions are identified, it is necessary to provide alternative solutions. To guarantee this consistency, a set of interface documents is being prepared. Those interfaces for each of the plant systems that can impact the PCS activities detail the requirements specifically needed for control and report also the functional relationship between the two systems. The PCS has also to consider areas not actively part of plasma control that might affect or limit PCS operations (i.e. forces in the superconducting coils). This paper reports the main outcome from the interfaces definition. The actuator boundaries and plant systems constraints impacting the PCS design will be presented. For the sensors the challenge is the derivation of real-time measurement requirements in relation to the separate diagnostic requirements and their respective interface with the PCS. The complex organization of data integration with the PCS will be discussed. [ABSTRACT FROM AUTHOR]

One of the ITER diagnostics having reached an advanced design stage is a diagnostic RGA for the divertor, i.e. residual gas analysis system for the ITER divertor, which is intended to sample the divertor pumping duct region during the plasma pulse and to have a response time compatible with plasma particle and impurity lifetimes in the divertor region. Main emphasis is placed on helium (He) concentration in the ducts, as well as the relative concentration between the hydrogen isotopes (mainly in the form of diatomic molecules of H, D, and T). Measurement of the concentration of radiative gases, such as neon (Ne) and nitrogen (N 2 ), is also intended. Numerical modeling of the gas flow from the sampled region to the cluster of analysis sensors, through a long (∼8 m long, ∼110 mm diameter) sampling pipe originating from a pressure reducing orifice, confirm that the desired response time (∼1 s for He or D 2 ) is achieved with the present design. [ABSTRACT FROM AUTHOR]

In a tokamak the magnetic diagnostics are key to the exploitation of the machine. They play a central role in the real-time control of fundamental plasma properties, such as the plasma shape and position, while also contributing with important data to a better understanding of the plasma physics. One of the particular challenges of the ITER magnetics diagnostic is the need to balance high system reliability with sufficient freedom to tune and improve the quality of the diagnostic physics output. This requirement calls for a design pattern where the functions related to plasma control and protection are loosely coupled with the functions related to the plasma science. This work reports on the current status of the magnetics plant system controller design and discusses some possible design solutions that address the aforementioned issue. [ABSTRACT FROM AUTHOR]

Highlights: [•] The concepts chosen for ITER Diagnostic Windows are given. [•] The description of the main components of the window assemblies is presented. [•] An overview of the R&D related to the window assemblies is given. [•] The process for their development is briefly explained. [Copyright &y& Elsevier]

Abstract: Diagnostics play an essential role for the successful operation of the ITER tokamak. They provide the means to observe control and to measure plasma during the operation of ITER tokamak. The components of the diagnostic system in the ITER tokamak will be installed in the vacuum vessel, in the cryostat, in the upper, equatorial and divertor ports, in the divertor cassettes and racks, as well as in various buildings. Diagnostic components that are placed in a high radiation environment are expected to operate for the life of ITER. There are approx. 45 diagnostic systems located on ITER. Some diagnostics incorporate direct or independently pumped extensions to maintain their necessary vacuum conditions. They require a base pressure less than 10−7 Pa, irrespective of plasma operation, and a leak rate of less than 10−10 Pam3 s−1. In all the cases it is essential to maintain the ITER closed fuel cycle. These directly coupled diagnostic systems are an integral part of the ITER vacuum containment and are therefore subject to the same design requirements for tritium and active gas confinement, for all normal and accidental conditions. All the diagnostics, whether or not pumped, incorporate penetration of the vacuum boundary (i.e. window assembly, vacuum feedthrough etc.) and demountable joints. Monitored guard volumes are provided for all elements of the vacuum boundary that are judged to be vulnerable by virtue of their construction, material, load specification etc. Standard arrangements are made for their construction and for the monitoring, evacuating and leak testing of these volumes. Diagnostic systems are incorporated at more than 20 ports on ITER. This paper will describe typical and particular arrangements of pumped diagnostic and monitored guard volume. The status of the diagnostic vacuum systems, which are at the start of their detailed design, will be outlined and the specific features of the vacuum systems in ports and extensions will be described. [Copyright &y& Elsevier]

Abstract: The measurement of the plasma current in ITER plays an outstanding role as it is part of the machine protection and is a safety-relevant measurement: it will be used in relation with regulatory limits to show that the operation remains within the safe envelope defined in the ITER license. The Continuous External Rogowski (CER) is an inductive sensor designed for current measurements and located in the casing of 3 Toroidal Field Coils (TFCs). After the completion of the design of the CER, 4 prototypes of the sensor were manufactured and R&D activities were performed under a Grant with the European Domestic Agency (F4E-GRT-012). The work was carried out between 2010 and 2011 by the ITERMAG consortium comprising 3 laboratories: CRPP (Switzerland) as leader, CEA (France) and RFX (Italy). The R&D campaign on CER prototypes consisted in the measurement of about 100 parameters to characterize the CER in terms of electrical, thermal, mechanical and also of vacuum compatibility. From these results, electromagnetic modeling of the CER response was performed. It is demonstrated that the CER fulfills ITER requirements. However, the vacuum compatibility of the prototype has to be improved and solutions to cope with this issue are proposed. [Copyright &y& Elsevier]

Abstract: The ITER Plasma Control System (PCS) requires an extensive set of about 50 diagnostic systems to measure the plasma response and about 20 actuators to act on the plasma to carry out its control functions. The specifications and real limitations of the actuators and diagnostics are being assessed as part of the ongoing conceptual design of the PCS to understand the potential impact on plasma control. The actuators include magnetic coils (central solenoid (CS), poloidal field (PF), vertical stability (VS), edge localized mode (ELM), correction coils (CC)), heating and current drive (electron cyclotron (EC), ion cyclotron (IC), neutral beam injection (NBI), and possibly lower hybrid (LH)), glow discharge cleaning, fueling and impurity gas and pellet injection, vacuum pumping, and disruption mitigation systems. Diagnostic systems are prioritized according to their role in machine protection (MP), basic control (BC), advanced control (AC), and physics studies (PS). At the conceptual design phase, detailed control algorithms do not yet need to be specified, but conceptual solutions must be chosen that satisfy the PCS requirements for control within the real constraints of the diagnostics and actuators. The feasibility of the chosen solutions must be proven either through established control schemes on existing machines or through an R&D program to develop them before they will be required on ITER. The diagnostic and actuator requirements of the PCS will evolve from first plasma through the high performance DT phase. A comparison is made of the expected requirements to control vertical stability, sawteeth, neoclassical tearing modes (NTMs), edge localized modes (ELMs), error fields, resistive wall modes (RWMs), Alfvén eigenmodes, and disruptions with the ITER baseline actuator and diagnostic specifications. [Copyright &y& Elsevier]

Abstract: ITER requires extensive diagnostic systems in order to meet the requirements for machine operation, protection, plasma control and physics studies. The realization of these systems is a major challenge not only because of the harsh environment and the nuclear requirements but also with respect to Instrumentation and Control (I&C) of all the 59 diagnostics plants. The Plant Systems I&C are mostly “in-kind”, i.e. procured by the seven ITER Domestic Agencies (DAs), while the Central I&C Systems are “in-fund”, i.e. procured by ITER Organization (IO). Standardization of Plant Systems I&C is of primary importance and has been one of the highest priority tasks of CODAC. The standards are published in the Plant Control Design Handbook (PCDH) which will be followed to ensure a homogeneous design, guarantee high availability and simplify maintenance and support future upgrades. Most important for a successful commissioning and operation of the ITER facility are the concepts of interfacing the diagnostics plant systems with CODAC and the standards for instrumentation and control which must be followed all contributing parties. In this paper, we will elaborate on the concepts of interfacing the diagnostics plant systems with CODAC and the standards that must be followed for the design. [Copyright &y& Elsevier]

Abstract: The integration of ITER in-vessel diagnostic components is an important engineering activity. The positioning of the diagnostic components must correlate not only with their functional specifications but also with the design of the major parts of ITER torus, in particular the vacuum vessel, blanket modules, blanket manifolds, divertor, and port plugs, some of which are not yet finally designed. Moreover, the recently introduced Edge Localised Mode (ELM)/Vertical Stability (VS) coils mounted on the vacuum vessel inner wall call for not only more than a simple review of the engineering design settled down for several years now, but also for a change in the in-vessel distribution of the diagnostic components and their full impact has yet to be determined. Meanwhile, the procurement arrangement (a document defining roles and responsibilities of ITER Organization and Domestic Agency(s) (DAs) for each in-kind procurement including technical scope of work, quality assurance requirements, schedule, administrative matters) for the vacuum vessel must be finalized. These make the interface process even more challenging in terms of meeting the vacuum vessel (VV) procurement arrangement''s deadline. The process of planning the installation of all the ITER diagnostics and integrating their installation into the ITER Integrated Project Schedule (IPS) is now underway. This paper covers the progress made recently on updating and issuing the interfaces of the in-vessel diagnostic components with the vacuum vessel, outlines the requirements for their attachment and summarises the installation sequence. [Copyright &y& Elsevier]

Abstract: The current status of the ITER magnetics diagnostic is summarised. The various risks are discussed on the basis of when they could be present during ITER operation and appropriate mitigation is reviewed. An implementation risk or common mode operation failure risk to the magnetic diagnostic can only be mitigated by diversity. All identified risks are being handled with ongoing research and development to minimise these risks. A reasonable level of design and manufacturing diversity has been therefore introduced in the procurement and installation plans, in order to optimise overall reliability avoiding dispersion of resources. [Copyright &y& Elsevier]

Abstract: On ITER the electrical connections to most in-vessel diagnostics, in particular the magnetic diagnostics, will require cables to be routed through the main vacuum vessel. The total in-vessel cabling requirement is predicted to be of order 80km. Due to the large quantity of cabling, the JET experience shows that the cabling has the potential to cause issues for the ITER vacuum and hence operability. Findings are presented from extensive tests to assess experimentally the outgassing of different types of mineral insulated cable, in damaged, undamaged and deliberately perforated conditions. Experience with using metal braided fibre insulated cable which has been used in-vessel on JET is given. Key points are concluded for the necessary manufacturing and installation techniques required for in-vessel cable to be used successfully on ITER. [Copyright &y& Elsevier]