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• 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]

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 report summarizes the results of in situ measurements of noninductive voltages developed across prototype International Thermonuclear Experimental Reactor (ITER) magnetic coils performed at the Japan Materials Test Reactor. The voltages appear to be mainly thermoelectric in origin, with the thermal sensitivity developing gradually through irradiation and the thermal gradients supplied by nuclear heating. Possible mechanisms and circuits within the coils that can contribute to this radiation-induced thermoelectric sensitivity are described, and schemes to improve the ITER coil design are outlined. [ABSTRACT FROM AUTHOR]

Magnetic coils are indispensable for plasma position control in tokamaks. For ITER, non-inductive voltages generated under irradiation have to be reduced to well below 1 μV in order to support long pulse (3000 s) operation. In situ measurements of the differential voltage between the ends of two magnetic coils wound with mineral-insulated (MI) cable have been carried out at the JMTR fission reactor. The MI cables of the two magnetic coils had copper center conductor with diameter 0.5 and 0.8 mm, respectively, stainless steel outer sheath and MgO insulation. The measured differential voltage for the two MI cables increased with neutron fluence, reaching 4.5 and -0.7 μV in the two coils, at a fast neutron fluence of 1.26 × 1023 n/m2. The magnitude of the measured voltage can be explained by thermoelectric potentials, enabled mainly by non-uniform transmutation and displacement damages of the copper core of the cable. [Copyright &y& Elsevier]

The measurement requirements and expected performance of the International Thermonuclear Experimental Reactor (ITER) magnetics system are summarized. The constraints of a burning plasma experiment (BPX) on tokamak magnetic diagnostic system designs are listed and illustrated using the parameters and design of sensors being developed for the ITER device. Design principles to meet the constraints are discussed, with an emphasis on radiation effects, reliability, and redundancy for sensors and subsystems. The remaining research and development for the ITER system is outlined. [ABSTRACT FROM AUTHOR]

Focuses on the use of reflectometry on the International Thermonuclear Experimental Reactor to measure the density profile in the main plasma and the divertor and to provide measurements of the plasma position and shape supplementary to the magnetic diagnostics. Systems required to access the full plasma profile; Measurements of the scrape-off layer; Real-time calibration technique of the whole transmission line.

ITER magnetic diagnostics are now in their detailed design and R&D phase. They have passed their conceptual design reviews and a working diagnostic specification has been prepared aimed at the ITER project requirements. This paper highlights specific design progress, in particular, for the in-vessel coils, steady state sensors, saddle loops and divertor sensors. Key changes in the measurement specifications, and a working concept of software and electronics are also outlined. [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 concept of multiplexing thermography is applied to the design of the divertor thermography system for International Thermonuclear Experimental Reactor (ITER). The combination of the front mirror with multiellipticity and a Czerney–Turner spectrometer with a 0.2 mm pitched multichannel detector enables a spatial resolution of 3 mm and a time resolution of 20 μs above a target temperature of 300 °C to be achieved. This should be sufficient to measure ELM heat fluxes to the targets in ITER. To satisfy the measurement requirement, it is very important to keep an accurate alignment around the optical axis against movement of the vessel during the plasma discharges. Several key engineering problems, such as the survivability of components against mirror coating by redeposited divertor material, remain to be solved. Potential solutions have been identified. [ABSTRACT FROM AUTHOR]

ITER First Plasma (FP) operation aims to produce a plasma with current higher than 100 kA for a duration longer than 100 ms. However, the low values of plasma density and temperature associated with the initiation phase make it difficult to diagnose the plasma accurately. It is therefore essential to develop models for the available diagnostics for ITER FP operation in order to determine the necessary measurement ranges for plasma initiation and use them as inputs for controller development and assessment within the plasma control system (PCS). The use of accurate diagnostic models also helps to optimally prepare for, and analyse, ITER FP operation. A model developed for the Hα main chamber visible spectroscopy for FP shows that it is possible to measure Hα emission soon after breakdown for plasma temperatures higher than 3 eV and average plasma densities higher than 1017 m−3. The diagnostic model for the interferometer shows that for the proposed ITER FP scenarios, accurate measurements are possible for average plasma densities higher than 5 × 1017 m−3. The results from modelling of the hard x-ray monitor demonstrate that it can provide signals useful for runaway electron detection and can measure current levels typically down to 3 kA. Thus, these measurements can thus be used effectively by the PCS for runaway electron (RE) detection during FP operation. Improved diagnosis of the plasma state deducing, for example, parameters such as the plasma size and position, is possible by combining the measurements from the different FP diagnostics (density interferometer polarimeter and Hα main chamber visible spectroscopy), which will be available for ITER FP operation. [ABSTRACT FROM AUTHOR]

Focuses on the installation of an integrated system of correlation reflectometers on the Joint European Torus tokamak. Description of the reflectometers; Components of the diagnostic; Ability of the diagnostic to provide unique information on the spatial and temporal characteristics of plasma fluctuations; Illustration of the capabilities of the system.

A technique based on a flexible endoscope was selected as a tool for the diagnostic of dust in ITER (“The Way” in Latin). The diagnostic will consist of two tools—one for fine viewing of dust with a resolution down to a few tens of microns in a few millimeter spot and another one for dust collection. The endoscope will have to go up to 15-m deep inside the tokamak to the inspection region. Due to the specific design features of ITER, the endoscope will have to go upward on an inclined surface for inspection about 18 m away from the insertion point. In order to ensure that the endoscope gets to the desired region of inspection, it will be pushed through guide tubes having a number of bends along their length. Initial estimations of endoscope jacket materials, endoscope stiffness, and push/pull forces were defined experimentally. This article will give a brief reminder of the overall strategy for dust/erosion/tritium monitoring in ITER that is a Protection Important Activity (PIA) and the role of the dust monitor in this context. It focuses on experimental results of real-size tests inside guide tubes of the behavior of different endoscope emulators under various conditions. [ABSTRACT FROM AUTHOR]

• 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]

In the case of fusion devices with long pulses or continuous operation, conventional magnetic sensors, such as inductive sensors, have to be supplemented by steady-state magnetic field sensors. Therefore, the Hall sensors are considered as a part of the DEMO magnetic diagnostics. The paper reviews the present status of the DEMO steady-state magnetic field sensors based on the ceramics-metal Hall sensors with a sensing layer made of antimony. The mission of the Hall sensors and their operating requirements are summarized, and the design of the Hall sensor assembly is presented. Satisfactory results of initial tests of the sensors are described. [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]

A magnetically driven fast-ion loss detector (FILD) is being designed for ITER. Due to the high heat load expected, fast movements and several measurement cycles will be required during its operation. The measurement cycle characteristics (measuring and parking periods, as well as the probe insertion and retraction times) will be defined as a compromise between the thermal load acting on the probe head and the measuring requirements. In this article, dynamic simulations of the detector movement are presented and the thermal behavior of the probe head is analyzed during its operation within a complete plasma discharge. [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]

Two electronics prototypes for the ITER outer vessel steady-state magnetic field Hall sensors based on a lock-in signal processing and current spinning technique were developed. The prototypes, the first using a rectangular lock-in signal processing with sensor offset cancellation and the second employing advanced lock-in signal processing with sensor and electronics offset cancellation, were compared with a DC technique and commercial lock-in amplifier. The influence of long interconnection cables, white noise, and interfering planar Hall effect to a measurement accuracy was evaluated. The planar Hall voltage suppression by the current spinning technique was investigated. Whilst the rectangular lock-in amplifier prototype failed to meet the ITER accuracy requirement, the advanced lock-in amplifier prototype demonstrated required functionality and accuracy. [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]

A dispersion interferometer is a reliable density measurement system and is being designed as a complementary density diagnostic on ITER. The dispersion interferometer is inherently insensitive to mechanical vibrations, and a combined polarimeter with the same line of sight can correct fringe jump errors. A proof of the principle of the CO2 laser dispersion interferometer combined with the PEM polarimeter was recently conducted, where the phase shift and the polarization angle were successfully measured simultaneously. Standard deviations of the line-average density and the polarization angle measurements over 1 s are 9 × 1016 m-2 and 0.19°, respectively, with a time constant of 100 μs. Drifts of the zero point, which determine the resolution in steady-state operation, correspond to 0.25% and 1% of the phase shift and the Faraday rotation angle expected on ITER. [ABSTRACT FROM AUTHOR]

A conceptual design of a reciprocating fast-ion loss detector for ITER has been developed and is presented here. Fast-ion orbit simulations in a 3D magnetic equilibrium and up-to-date first wall have been carried out to revise the measurement requirements for the lost alpha monitor in ITER. In agreement with recent observations, the simulations presented here suggest that a pitch-angle resolution of ~5° might be necessary to identify the loss mechanisms. Synthetic measurements including realistic lost alpha-particle as well as neutron and gamma fluxes predict scintillator signal-to-noise levels measurable with standard light acquisition systems with the detector aperture at ~11 cm outside of the diagnostic first wall. At measurement position, heat load on detector head is comparable to that in present devices. [ABSTRACT FROM AUTHOR]

Performance of bismuth Hall sensors developed for the ITER steady state magnetic diagnostic was investigated for high magnetic fields in the range ±7 T. Response of the sensors to the magnetic field was found to be nonlinear particularly within the range ±1 T. Significant contribution of the planar Hall effect to the sensors output voltage causing undesirable cross field sensitivity was identified. It was demonstrated that this effect can be minimized by the optimization of the sensor geometry and alignment with the magnetic field and by the application of "current-spinning technique." [ABSTRACT FROM AUTHOR]

This paper presents the results of the R&D activities recently performed to achieve the preliminary design of electromagnetic transducers for in-vessel local magnetic field measurements at low frequency (<30 kHz) for ITER. The study is based on thermoelectrical and electromagnetic bench tests performed on a batch of 40 sensor prototypes manufactured with the low temperature cofired ceramic (LTCC) technology. Also, the outcomes of a first neutron irradiation test on LTCC sensors are presented. The development of an alternative concept of a magnetic sensor for ITER, based on mineral insulated cables winding is also reported, with suitable thermomechanical and electromagnetic numerical analyses performed to assess the reliability of such sensors within the ITER operating conditions. [ABSTRACT FROM AUTHOR]

The present paper is focused on the sensitivity analysis of the diamagnetic sensor measurements of ITER against several kinds of error sources, with the aim of compensating them for improving the accuracy in the evaluation of the energy confinement time and poloidal beta, via Shafranov formula. The virtual values of measurements at the diamagnetic sensors were simulated by the COMPFLUX code, a numerical code able to compute the field and flux values generated in a prescribed set of output points from massive conductors and generalized filamentary currents (with an arbitrary 3D shape and a negligible cross section) in the presence of magnetic materials. The major issue to face with has been to determine the possible deformations of sensors and electromagnetic sources. The analysis has been carried out considering the following cases: - deformed sensors and ideal EM (electromagnetic) sources; - ideal sensors and perturbed EM sources; - both sensors and EM sources perturbed. As regards the compensation, several formulas have been proposed, based on the measurements carried out by the compensation coils; they basically use the value of the flux density measured to compensate the effects of the poloidal eddy currents induced in the conducting structures surrounding the plasma. The static deviation due to sensor manufacturing and positioning errors has been evaluated, and most of the pollution of the diamagnetic flux has been compensated, meeting the prescribed specifications and tolerances. [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]

This paper contains the design of the beam dump for the ITER edge Thomson scattering system and mainly concerns its lifetime under the harsh thermal and electromagnetic loads as well as tight space allocation. The lifetime was estimated from the multi-pulse laser-induced damage threshold. In order to extend its lifetime, the structure of the beam dump was optimized. A number of bent sheets aligned parallel in the beam dump form a shape called a chevron which enables it to avoid the concentration of the incident laser pulse energy. The chevron beam dump is expected to withstand thermal loads due to nuclear heating, radiation from the plasma, and numerous incident laser pulses throughout the entire ITER project with a reasonable margin for the peak factor of the beam profile. Structural analysis was also carried out in case of electromagnetic loads during a disruption. Moreover, detailed issues for more accurate assessments of the beam dump's lifetime are clarified. Variation of the bi-directional reflection distribution function (BRDF) due to erosion by or contamination of neutral particles derived from the plasma is one of the most critical issues that needs to be resolved. In this paper, the BRDF was assumed, and the total amount of stray light and the absorbed laser energy profile on the beam dump were evaluated. [ABSTRACT FROM AUTHOR]

Abstract: ITER is targeting Q =10 with 500MW of fusion power. To meet this target, the plasma needs to be controlled and shaped for a period of hundreds of seconds, avoiding contact with internal components, and acting against instabilities that could result in the loss of control of the plasma and in its disruptive termination. Axisymmetric magnetic control is a well-understood area being the basic control for any tokamak device. ITER adds more stringent constraints to the control primarily due to machine protection and engineering limits. The limits on the actuators by means of the maximum current and voltage at the coils and the few hundred ms time response of the vacuum vessel requires optimization of the control strategies and the validation of the capabilities of the machine in controlling the designed scenarios. Scenarios have been optimized with realistic control strategies able to guarantee robust control against plasma behavior and engineering limits due to recent changes in the ITER design. Technological issues such as performance changes associated with the optimization of the final design of the central solenoid, control of fast transitions like H to L mode to avoid plasma-wall contact, and optimization of the plasma ramp-down have been modeled to demonstrate the successful operability of ITER and compatibility with the latest refinements in the magnetic system design. Validation and optimization of the scenarios refining the operational space available for ITER and associated control strategies will be proposed. The present capabilities of magnetic control will be assessed and the remaining critical aspects that still need to be refined will be presented. The paper will also demonstrate the capabilities of the diagnostic system for magnetic control as a basic element for control. In fact, the noisy environment (affecting primarily vertical stability), the non-axisymmetric elements in the machine structure (affecting the accuracy of the identification of the plasma boundary), and the strong component of eddy current at the start-up (resulting in a poor S/N ratio for plasma reconstruction for I p <2MA requiring a robust plasma control) make the ITER magnetic diagnostic system a demanding part of the magnetic control and investment protection systems. Finally the paper will illustrate the identified roles of magnetic control in the PCS (plasma control system) as formally defined in the recent first step of the design and development of the system. [Copyright &y& Elsevier]

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]

In ITER, mirrors will be used as plasma-viewing elements in all optical and laser diagnostics. In the harsh environment, mirror performance will degrade hampering the operation of associated diagnostics. The most adverse effect on mirror performance is caused by the deposition of impurities. It is expected that the most challenging situation will occur in the divertor. With the envisaged changes to all-metal plasma-facing components (PFCs) in ITER, an assessment of mirror performance in an existing divertor tokamak with all-metal PFCs is urgently needed. Molybdenum and copper mirrors were exposed for nearly nine months in ASDEX Upgrade which has all-tungsten PFCs. Mirrors were located at the inner wall, under the dome and in the pump duct. During exposure, the mirrors were heated to temperature in the range 145-165 °C. This was made to approach the expected level of heating due to absorption of neutrons and gammas on mirrors in the ITER divertor. After exposure, degradation of the reflectivity was detected on all mirrors. The highest reflectivity drop was measured on mirrors under the dome facing the outer strike point, reaching -55% at 500 nm. The least degradation was detected on mirrors in the pump duct, where the reflectivity was preserved in the range 500-2500 nm and the largest decrease was about -8% at 250 nm. On all contaminated mirrors carbon fraction did not exceed 50 at% while the major contaminants were metals and oxygen. The degradation of exposed mirrors underlines the necessity for urgent R&D on deposition mitigation and in situ mirror cleaning in ITER. [ABSTRACT FROM AUTHOR]

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: The nuclear heating rates of the optical mirrors of the poloidal polarimeter installed in the equatorial port plug of ITER are calculated. Since the system cannot have a sufficiently labyrinthine structure and the second mirrors are located closer to the plasma than the first mirrors due to limited space, the nuclear heating rate of the second mirrors is as high as that of the first mirrors. However, it is possible to reduce the nuclear heating rates more than 1 order if the blanket shield module provides a sufficient degree of neutron shielding. The effect of the neutron shield on the peripheral region of the upper port plug, in which the optical system of the poloidal polarimeter is installed, is also evaluated. The gamma-ray dose rate 106 s after shutdown of operations in the back flange region has been estimated to be ∼60μSv/h when the neutron shield is filled in the port plug, which is below the design threshold for the ITER maintenance phase. Further, the nuclear heating rate of the Poloidal field coil near the same upper port has been evaluated to be more than 2 orders lower than the allowable value of 1mW/cc. [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]

The International Thermonuclear Experimental Reactor will have wide angle viewing systems and a divertor thermography diagnostic, which shall provide infrared coverage of the divertor and large parts of the first wall surfaces with spatial and temporal resolution adequate for operational purposes and higher resolved details of the divertor and other areas for physics investigations. We propose specifications for each system such that they jointly respond to the requirements. Risk analysis driven priorities for future work concern mirror degradation, interfaces with other diagnostics, radiation damage to refractive optics, reflections, and the development of calibration and measurement methods for varying optical and thermal target properties. [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 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]

The next generation of large scale fusion devices—ITER/LMJ/NIF—will require diagnostic components to operate in environments far more severe than those encountered in present facilities. This harsh environment is the result of high fluxes of neutrons, gamma rays, energetic ions, electromagnetic radiation, and in some cases, debris and shrapnel, at levels several orders of magnitude higher than those experienced in today’s devices. The similarities and dissimilarities between environmental effects on diagnostic components for the inertial confinement and magnetic confinement fusion fields have been assessed. Areas in which considerable overlap have been identified are optical transmission materials and optical fibers in particular, neutron detection systems and electronics needs. Although both fields extensively use cables in the hostile environment, there is little overlap because the environments and requirements are very different. [ABSTRACT FROM AUTHOR]

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]