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• Saturated-wave theory for snubber modeling is modified to fix three conceptual errors. • The equivalent resistance and inductance of the snubber are recalculated. • Calculating the thickness of eddy-current region no longer relies on external circuit. • The current rise of the model is nearly the same as the experiment with about 10 % over-estimated peak shorting current. Core snubber is a common device used for breakdown protection in auxiliary heating systems. It absorbs energy from the high-voltage power supply to protect the load. The existing snubber models are based on saturation-wave theory, but the theory still has several conceptual problems. In this paper, an improved saturated-wave theory is presented. We consider that the eddy current region of the tape needs to be layered and that there is an unsaturated region in the tape which enables the snubber to be equated to an inductor. Then we divided the working process of snubber into 4 stages and rederived the circuit model of the snubber. The comparison with the existing simulation model shows that the new model in this paper is closer to the experimental data, especially in the rising phase of short-circuit current. It can more accurately analyze the conditions within a few microseconds after breakdown, and give a useful reference for the design of breakdown protection and power supply systems. [ABSTRACT FROM AUTHOR]

• The main thermal-hydraulic parameters of the ITER IBED PHTS have been assessed under selected operational scenarios. • The Fathom model of the ITER IBED PHTS has been revised and updated. • Three scenarios are presented, i.e. one that simulates hot standby conditions and the other two that simulate the normal operation. • The study confirms the correct operation of the system while highlighting some potential criticalities. One of the key elements for the success of ITER is the efficient removal of the thermal power generated by the deuterium-tritium fusion reaction within the confined plasma. The Integrated Blanket ELM/VS Coils and Divertor (IBED) Primary Heat Transport System (PHTS) is a pressurized, closed-loop cooling system designed to supply cooling water to the in-Vessel components. To support process design of the IBED PHTS, a thermal-hydraulic research campaign has been conducted to analyse the proper operation of the system during selected operational scenarios. The work has been carried out using the commercially available thermal-hydraulic software AFT Fathom, performing steady-state hydraulic and heat transfer analyses of complex hydraulic systems. The present paper summarizes the research activity, describing the computational model used and the scenarios investigated, as well as the results obtained and their critical discussion. [ABSTRACT FROM AUTHOR]

Highlights • The ITER TBM Program plans to test DEMO-relevant tritium breeding blanket systems. • Each TBS is formed by various sub-systems with different functions and technologies. • For each TBS sub-system, description of its functions and its main components. • Identification of the ITER services required for operating each TBS sub-system. • Helium and water coolant systems, tritium extraction systems, Pb-16Li systems, I&C. Abstract Six Test Blanket Systems (TBS) will be installed and operated in ITER in the framework of the ITER Test Blanket Module (TBM) Program. After a short description of each TBS and of the main operating parameters, this paper gives the main features of the various TBSs sub-systems, in particular, the cooling systems (helium, water and/or lithium-lead), the tritium system for extracting the tritium from the breeder, the coolant purification system, the tritium accountancy system, and several instrumentation & control systems. This paper addresses also the main services that are needed to operate a TBS such as the secondary water cooling system, the chilled water system to cool the helium circulators, the liquid and gaseous nitrogen used in the tritium extraction process, the compressed air needed to operate the various valves present on the circuits, and the electrical power supply for all the active electrical components such as pumps, circulators and heaters. Other important aspects are the required interfaces with other ITER plant systems such as those involving air and process gas detritiation, power supply, radiological and environmental monitoring, remote handling, measurements, and services. [ABSTRACT FROM AUTHOR]

• STEAM will host the mock-up of the EU-DEMO Steam Generators. • STEAM primary loop has been investigated with RELAP5/Mod3.3 system code. • DEMO normal operations have been studied including pulse-dwell transitions. • Different primary loop management strategies have been numerically simulated. • Setting the primary average temperature minimizes the loop operative fluctuations. The Water Cooled Lithium Lead Breeding Blanket (WCLL BB) is a key candidate for the driver blanket of the European DEMO reactor, progressing toward its Conceptual phase by the end of 2027. To assess different water and lithium-lead technologies for the WCLL BB and Balance of Plant (BoP) systems, the Water-thermal-HYDRAulic (W-HYDRA) experimental platform is under development at the ENEA Brasimone Research Centre. Among the facilities constituting the new W-HYDRA multipurpose infrastructure, STEAM is going to experimentally investigate the DEMO WCLL BoP thermal-hydraulics, focusing on the Steam Generator (SG) of the Primary Heat Transfer Systems (PHTS), to qualify its performances and suitability under its unconventional operation. The paper aims at supporting the thermal-hydraulic characterization of the Steam Generator mock-up during the sudden power variations typical of a pulsed fusion reactor. The analyzed selected scenario is the operational transient dwell-pulse-dwell, which determines high thermal cycling and correspondent high thermo-mechanical stresses on the primary side components. Two control logics with their relative drawbacks have been analyzed with a RELAP5/Mod3.3 1-D model, the first regulating the primary side average temperature, the second monitoring the minimum one. The comparison of the two systems highlighted that neither approach leads to hazardous conditions for the facility. However, while the average temperature controller is characterized by reduced thermal stresses on the components, the minimum temperature controller is characterized by higher thermal gradients for the primary loop. Both methodologies will be tested in the dedicated experimental campaign, aiming at yielding insights and evaluations concerning control strategies applicable to the DEMO reactor. [ABSTRACT FROM AUTHOR]

An Electron Cyclotron (EC) Heating and Current Drive (HCD) system will be installed in the EU DEMO Tokamak to provide plasma heating and stabilization by coupling up to 130 MW of mm-wave power into specific regions of the plasma. The mm-waves will be generated by gyrotrons and propagated through a quasi-optical multi-beam mirror system (MBMS), and subsequently through corrugated waveguides, to launchers located at dedicated Equatorial Ports (EPs). The present design of the launcher features two mid steering antennas (MSAs) for plasma stabilization each consisting of three-beam layouts and steering mirrors positioned at the upper and lower locations of the EP. The proposed concept for the steering mirrors operates on the pantograph principle, a mechanical assembly of parallel linkages, which translates the movement produced by pneumatic actuators into an exact angular rotation of the mirror around an offset virtual axis. This design uses flexures with specific structural compliance as mechanical joints resulting in the desired movement of the linkage system while avoiding the tribological difficulties associated with mechanisms operating under vacuum. This paper describes the present status of the DEMO EC HCD steering mirror as well as the analyses performed to validate this design. [ABSTRACT FROM AUTHOR]

The neutral beam test facility, which is under construction in Padova, Italy, is developing the ITER full-scale ion source for the ITER heating neutral beam injectors, referred to as the SPIDER experiment, and the full-size prototype injector, referred to as MITICA. The SPIDER control and data acquisition system (CODAS) has been developed and its construction will start in 2014. Slow control and data acquisition will be based on the ITER CODAC core system software suite that has been designed to facilitate the integration of ITER plant systems with CODAC. Fast control and data acquisition will use solutions specific to the test facility, as the corresponding concepts are not ready-to-use in the ITER design. The ITER hardware catalog for fast control has been taken into consideration. The software development will be based on the integration of MDSplus and MARTe, two framework software packages that are well known in the fusion community, targeting data organization and fast real-time control, respectively. The paper revises the system requirements and the system design and shows the results already achieved in terms of system integration. In addition, the paper will report the experience in the usage of different cooperating software frameworks and in the integration of industrial procured plant systems. [ABSTRACT FROM AUTHOR]

• Work completed by the Heating and Current Drive Workpackage for DEMO during the pre-conceptual design phase (2014-2020) • The Electron Cyclotron System is the baseline for DEMO • Description of the Electron Cyclotron System • Description of the Ion Cyclotron System • Description of the Neutral Beam Heating System The European DEMO is a pulsed device with pulse length of 2 hours. The functions devoted to the heating and current drive system are: plasma breakdown, plasma ramp-up to the flat-top where fusion reactions occur, the control of the plasma during the flat-top phase, and finally the plasma ramp-down. The EU-DEMO project was in a Pre-Concept Design Phase during 2014-2020, meaning that in some cases, the design values of the device and the precise requirements from the physics point of view were not yet frozen. A total of 130 MW was considered for the all phases of the plasma: in the flat top, 30 MW is required for neoclassical tearing modes (NTM) control, 30 MW for burn control, and 70 MW for the control of thermal instability (TI), without any specific functions requested from each system, Electron Cyclotron (EC), Ion Cyclotron (IC), or Neutral Beam (NB) Injection. At the beginning of 2020, a strategic decision was taken, to consider EC as the baseline for the next phase (in 2021 and beyond). R&D on IC and NB will be risk mitigation measures. In parallel with progresses in Physics modelling, a decision point on the heating strategy will be taken by 2024. This paper describes the status of the R&D development during the period 2014-2020. It assumes that the 3 systems EC, IC and NB will be needed. For integration studies, they are assumed to be implemented at a power level of at least 50 MW. This paper describes in detail the status reached by the EC, IC and NB at the end of 2020. It will be used in the future for further development of the baseline heating method EC, and serves as starting point to further develop IC and NB in areas needed for these systems to be considered for DEMO. [ABSTRACT FROM AUTHOR]

Experimental Advanced Superconducting Tokamak (EAST) is one of the important steady-state divertor Tokamak physical experiment platforms in the world. The main functions of the high-temperature and high-pressure Divertor Primary Heat Transfer System (DIV PHTS) are to provide high-pressure cooling water for the divertor to remove the heat load deposited on the divertor under the plasma operation mode and to provide high-temperature nitrogen gas for the divertor system to achieve vacuum condition during the baking mode. The stability and reliable operation of the DIV PHTS play an important role in the EAST tokamak plasma physics experiment. The high-temperature and high-voltage electrical insulation break are one of the important components of the system. The main function of the Electrical Insulation Break (EIB) is to break the closed circuit of the heat transfer system to avoid eddy current during plasma discharge affecting the magnetic field distribution characteristics of the tokamak. On the other way, it also needs to meet the requirement, there is no leak of the cooling water and nitrogen gas during cooling and baking operation mode. This paper proposes the design of an electrical insulating break (EIB) to meet the design requirements of EAST tokamak at first. Then the electric field analysis is carried out by using a finite element model to evaluate its electrical insulation characteristics. And the sealing structural characteristics are verified by applying the finite element method under two design modes. Finally, the sealing test result under the high-temperature baking and high-pressure cooling modes are presented to validate the design. [ABSTRACT FROM AUTHOR]

The paper describes commissioning of the inductive heater based on numerical simulation of inductive heating in the hypervapotron sample and test channel. The HEFEL loop is being developed to research a high heat flux cooling channels applicable in fusion devices. The Box Scraper hypervapotron channel geometry represents a primary focus of research. Conditions in the first wall of tokamak are simulated by induction heating of the tested hypervapotron channel. The exact heat flux of tested sample is difficult to measure, thus the numerical simulation of induction heating and thermohydraulic model of the cooling channel are an important part of HEFEL loop commissioning. ANSYS Maxwell 3D and ANSYS Fluent coupling is used for optimalization of the induction heating to achieve the highest heat fluxes using selected induction heating source. [ABSTRACT FROM AUTHOR]

The Experimental Advanced Superconducting Tokamak (EAST) had been under major upgrading from 2020 to 2021 in China. In this campaign of major upgrading, the new lower tungsten divertor is successfully installed to achieve the targets of long pulse and high β H-mode plasma operation. High temperature and high pressure Divertor primary heat transfer system (DIV PHTS) is also developed to remove the plasma deposition heat load on the Divertor. This paper presents a description of the design and construction of DIV PHTS for EAST Tokamak. Starting from the requirements specified in the Technical Specifications of DIV PHTS, the process to meet the Divertor cooling and baking requirement is designed and validated by the supported thermal-hydraulic analysis. Then, the piping stress analysis under different types of load combinations is performed to verify the structural design requirements. General layout design is also performed based on the component requirement and the allocated space in the EAST Tokamak hall. Finally, the procurement and test of the system main equipment is also given. [ABSTRACT FROM AUTHOR]

Abstract: The neutral beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design development of the components between the injectors and the tokamak; the so-called ‘front end components’: the drift duct consists of the NB bellows and the drift duct liner, the vacuum vessel pressure suppression system box (VVPSS box), the absolute valve, and the fast shutter. These components represent the key links between the ITER tokamak and the vessels of the NB injectors. The design of these components is demanding due to the different loads that these components will have to stand. The paper will describe the different design solutions which have to be implemented regarding the primary vacuum confinement, the power handling capability and the remote maintenance operations. The sizes of the components are determined by the large cross section of the neutral beam. The power handling capability is driven by the anticipated re-ionization of the neutral beam and the electromagnetic fields in this region. The drift duct bellows (with an inner diameter of 2.5m) shall guarantee a leak tight vacuum enclosure during the vertical and radial displacements of the ITER vacuum vessel. The conductance of the VVPSS box must be maximized in the available space. The absolute valve remains a challenging development. The total leak rate through the valve must be ≤1×10−8 Pam3/s when the valve is closed. Due to the radiation environment, the seals of the gate valve will be metallic. An R&D program has been launched to develop a suitable metallic seal solution with the required dimensions. The maximum allowed closing time for the fast shutter shall be less than 1s. For all these components the leak tightness will be guaranteed by a welded lip seal and the mechanical stability by bolted structures. [Copyright &y& Elsevier]

Abstract: Establishing adequate technical and physical boundary conditions for a sustained nuclear fusion reaction is a challenging task. Phased feedback control and monitoring for heating, fuelling and magnetic shaping is mandatory, especially for fusion devices aiming at high performance plasmas. Technical and physical interrelations require close collaboration of many components in sequential as well as in parallel processing flows. Moreover, handling of asynchronous, off-normal events has become a key element of modern plasma performance optimisation and machine protection recipes. The manifoldness of plasma states and events, the variety of plant system operation states and the diversity in diagnostic data sampling rates can hardly be mastered with a rigid control scheme. Rather, an adaptive system topology in combination with sophisticated synchronisation and process scheduling mechanisms is suited for such an environment. Moreover, the system is subject to real-time control constraints: response times must be deterministic and adequately short. Therefore, the experimental tokamak device ASDEX Upgrade employs a discharge control system DCS, whose core has been designed to meet these requirements. In the paper we will compare the scheduling schemes for the parallelised realisation of a control workflow and show the advantage of a data-driven workflow over a managed workflow. The data-driven workflow as used in DCS is based on signals connecting process outputs and inputs. These are implemented as real-time streams of data samples. Consequently, real-time signal management forms the foundation of DCS. The paper explains the principal features such as tagged samples, signal groups, algorithmic blocks and processes as well as scheduling schemes which allow DCS control applications to be defined as self-contained modular building blocks glued together by a software framework. By virtue of this sound foundation, DCS is a mature but still evolving system for reliable, distributed control of an entire tokamak device coordinating and monitoring 20 diagnostic systems, 14 magnetic power supplies, 5 heating systems with a total power of more than 25MW, 8 gas fuelling channels, a pellet injector and a killer gas gun. [Copyright &y& Elsevier]

Highlights • TBR and nuclear heating have been studied for the last version of DCLL BB for DEMO. • Upgrades of this baseline are analyzed under neutronic point of view. • The impact on nuclear performances of different FW designs is addressed. • Two FCI designs with different strategies for their neutronic simulation also affect the TBR result. • The divertor compositions is also demonstrated to have impact on TBR. • NBI integration consequences on TBR and NH are evaluated. Abstract On the way towards a comprehensive design of DEMO, step by step all the systems and components must be introduced as their definition or refinement progresses, in order to demonstrate the viability of a design on larger scale, i.e. leaving fewer margins to undetermined questions. Among the EUROfusion Programme, new aspects have been recently fixed or further developed as the Divertor, the First Wall (FW) and the Flow Channel Inserts (FCI) designs. Furthermore, the integration of Heating and Current Drive (H&CD) systems, as the Neutral Beam Injector (NBI), has started. The introduction or modification of these systems and components could seriously jeopardize the nuclear behaviour of an initially validated Breeding Blanket (BB) DEMO concept, since many neutronics criteria - among others - could be no more fulfilled. Since the design of DEMO is a continuous upgrade under iterative process, as the advances push on, most of the studies have to be repeated to demonstrate that criteria are still respected in a fully integrated design. The consequences of these upgrades over the neutronic responses are addressed in this paper. Among others, the influence on Tritium Breeding Ratio (TBR) of a new design of detached FW protecting BB from high heat fluxes is investigated. The impact of different typologies of FCIs is assessed also according to the degree of detail in the neutronic description. The divertor composition also reveals to have strong impact on responses apparently not related with its design, as the tritium production in the BB. Besides, the integration of NBI minimizing its invasiveness in the BB is verified by neutronic analyses concerning the main BB functions: fuel breeding and heat generation. Accordingly, TBR and Nuclear Heating (NH) are assessed. The study is performed for a Dual Coolant Lead Lithium (DCLL) BB DEMO although can be extrapolated to other BB concepts. [ABSTRACT FROM AUTHOR]

Highlights • Investigation of the influence of homogenization on nuclear heating and shielding. • Heterogeneous model with detailed description of the BSM cooling system. • Homogeneous model simplified by removing pipes and introducing mixed materials. • Nuclear heating was calculated for several components and on a fine mesh for both models. • Homogenization tends to underestimate the local heat deposition and overestimates shielding. Abstract The electron cyclotron-heating upper launcher (ECHUL) will be installed in four upper ports of the ITER tokamak thermonuclear fusion reactor. Each ECHUL is able to deposit 8 MW power into the plasma for plasma mode stabilization via microwave beam lines. These beam lines and other components needed for the guidance of the microwaves are mounted into the upper port plug. In order to protect these components and to mitigate radiation leaving the reactor, several shield blocks are also included in the port plugs design. The ECHUL components which are closest to the fusion plasma are heated by the gamma and neutron radiation and therefore will need to be actively cooled. This paper reports on the influence of material homogenization by comparing results obtained in the neutronic analyses performed for the cooling system of the ECHUL blanket shield module (BSM). A small reduction in shielding performance of the BSM was found in the case of a heterogeneous description of the BSM compared to the standard approach of homogenizing the materials of the BSM by removing pipes and cooling circuits for the neutronic analyses. [ABSTRACT FROM AUTHOR]

• An ICRH antenna has been built for the largest stellarator in the world, Wendelstein 7-X. This is also the first time that an ICRH antenna is constructed for such a stellarator. • The antenna is installed in Wendelstein 7-X and is ready for operations. • Main aims are: heating of stellarator plasmas, plasma startup (especially at low magnetic field, where ECRH cannot be used), fast particle generation to test the confinement properties of the optimized magnetic configuration of Wendelstein 7-X and wall conditioning. The ICRH antenna for the stellarator Wendelstein 7-X (W7-X) aims at delivering RF power levels up to about 1.5 MW in the frequency range 25–38 MHz with pulse lengths up to 10 s. The antenna was constructed and tested in the institute IEK-4 of the Research Centre Jülich, and subsequently installed at W7-X. The paper will review the physics design, the construction and installation in W7-X, and the commissioning plans for the ICRH system in the coming months. [Display omitted] [ABSTRACT FROM AUTHOR]

The discharge scenario query system, designed and deployed since 2014, is an important part of the EAST remote discharge scenario management system (RDSMS) Chai et al. [ 1 ]. Parameter analysis and extraction are the key function for classifying and querying scenarios. On EAST, the plasma control system (PCS) can parse the pre-set parameters line-by-line from future shots (taken from pre-set files in PCS server) or from past shots (taken from the MDSplus Manduchi et al. [ 2,3 ] tree) through a two-step conversion while the discharge parameters can be taken from the MDSplus tree directly without any conversion. To help with searching through enormous stock of past and future scenarios that are archived at EAST, we take advantage of the parsing procedure of the PCS and MDSplus programming by developing a “Parsing Program” to extract the pre-set and discharge parameters. This paper introduces the storage structure and parsing procedure of pre-set files, going into some detail for the latter. The other parameters, on the other hand, need little explanation as they are taken directly from the MDSplus tree. The pre-set parameters and the data analysis of the completed shot are archived in the MySQL database for future advanced scenario queries. In order to provide a cross platform tool, a remote graphical user interface has been built in form of a dynamic web page to query and evaluate parameters as well as provide scenario verification via auditors. This web-based discharge scenario query system will ultimately lead to improved work efficiency of technicians, researchers, scientists, users in general. [ABSTRACT FROM AUTHOR]

The guard limiter (GL) with graphite tiles of 4.6 GHz Lower Hybrid Antennas (LHA) has experienced the discharges in EAST tokamak. Much more hot spots were observed on the surface of GL tiles during the 2015–2016 campaigns. The serious ablations of GL induced by the hot spots limit the superconductor current, the plasma density and the heating power at higher parameter operations. In this paper, the GL issues are introduced with the original design. The previous improvements on the GL tiles are discussed with the ablation problems. In particular, the hotspots and the heat affected zones are discussed to predict the GL profile in accordance with the physical estimation. A trimmed region with a bevel angel of 11° is applied to avoid the heat load deposition at millimeter magnitude, which is caused by the fast electrons hitting due to the LHA heat driving. As a result, the GL ablations are confirmed to be reduced to the minimum. Moreover, it indicates that the plasma hitting would contribute most on the tiles fracture in the mid plane of the GL. In addition, new design activities on the tungsten GL are summarized in the engineering design. The new GL would be expected in the year 2017. [ABSTRACT FROM AUTHOR]

• Insulation deterioration of coils under varying weather conditions is investigated and actions were taken to revive machine operation. • Sequence and consequences of fall in the insulation resistance of coils due to moisture condensation is studied. • Different techniques are employed to recover and maintain the insulation resistance of the different coils under high humidity range. Insulation resistance (IR) of electromagnetic (EM) coils show a tremendous variation with the changing weather condition. This effect has a major impact on the operation of the ADITYA-U tokamak operation schedule. The failure of EM coil insulation can lead to terrible failure of the coil system and subsequently, the component and respective power system. Series of circumstances are driven to check their impact on the relative humidity (RH) and simultaneously their effect on the IR of the EM coils has been observed. Different techniques are employed to control the humidity and temperature of the experiment hall to maintain the IR of the different EM coils within the experimental limits. In this paper, different factors have been reviewed which affect and change the IR values of the EM coils. Experience gained during years of operation during insulation strength degradation and its recovery is presented in this work. [ABSTRACT FROM AUTHOR]

• A second 1.2 MW/53 keV high-energy neutral beam (NBI-2) procured, manufactured and installed on TCV. • NBI-2 is designed to probe plasma physics issues of higher plasma density, fast ion plasma interactions with static and dynamic fields and plasma rotation. • The higher NBI-2 energy together with NBI-1 (1.3 MW/28 keV) greatly enhances the operational space for fast ion studies and plasma heating with up to 2.5 MW NBH power. The TCV tokamak continues to leverage its unique shaping capabilities, flexible heating systems and modern control system to provide a stepladder approach for extrapolations of experimental results from different machines to ITER and DEMO. TCV's configurational flexibility has recently been enhanced by the installation of two heating neutral beams, new dual frequency gyrotrons, removable divertor gas baffles together with upgrades of its diagnostic infrastructure. Plasma regimes with high plasma pressure, a wider range of temperature ratios and significant fast-ion population are now attainable on TCV. A 1.3 MW/28 keV heating beam (NBI-1) has been operated on TCV from 2015 introducing the direct ion auxiliary heating. A second 1MW/50–60 keV high-energy neutral beam (NBI-2) has since been procured, manufactured and installed on TCV in July 2021. NBI-2 injected near anti-co-linearly to NBI-1 is designed to probe plasma physics issues of higher plasma density, fast ion plasma interactions with static and dynamic fields and plasma rotation. The higher NBI-2 energy greatly enhances the operational space for fast ion studies. NBI-2 commissioning is presently ongoing in parallel with TCV experiments using both NBIs. This paper summarises the experience with intensive use and improvements of the NBI-1, together with the status of NBI-2 commissioning. [ABSTRACT FROM AUTHOR]

In the last experimental campaign (OP1.2b: 7/2018 to 10/2018) of the Wendelstein 7-X stellarator (W7-X), the first of two neutral beam injectors (NBI) was commissioned with 2 plasma sources. The safe operation of the NBI requires the exclusion of hazards to personnel and minimization of system hazards. The NBIcontrol system consists of a standard PLC based on PCS7 from Siemens AG and a separate safety controller based on PNOZmulti from Pilz GmbH. The requirements of the NBI safety control are derived from an analysis of the specific personal and plant hazards in accordance with the international standards IEC 61508 / IEC61511. Particular challenges are the usage of high voltage, open high frequency and high power sub systems, and explosive gases that are distributed over different locations. Firstly, the paper describes the procedure followed by the hazard assessment and for determining solutions for safe operation. Secondly, the paper, describes the hardware structure of the NBI safety system. Lastly details of individual technical solutions, e.g. for access to the high voltage area of the NBI injector, are described. [ABSTRACT FROM AUTHOR]

A power supply (20 V, 10 kA) for powering the filamentary cathode has been procured, interfaced and integrated with the centralized control system of Large Volume Plasma Device (LVPD). Software interface has been developed on the standard Modbus RTU communication protocol. It facilitates the dashboard for configuration, on line status monitoring, alarm management, data acquisition, synchronization and controls. It has been tested for stable operation of the power supply for the operational capabilities. The paper highlights the motivation, interface description, implementation and results obtained. [ABSTRACT FROM AUTHOR]

• The size of the gaps around the ITER equatorial and upper ports has associated a certain tolerance due to manufacturing and installation. • Gap size variability has an impact on the nuclear responses. • MCNP simulations show the extent of the impact of the gap variability. • A novel modelling approach, which was developed ad-hoc and used to manipulate the size of the port gaps via displacements of the geometry universes, is described. The size of the gaps around the ITER equatorial and upper ports has associated a certain tolerance due to manufacturing and installation. This variability has an impact on the nuclear responses at magnets and port interspace maintenance locations. This effect is not usually considered as the ITER neutronic reference models represent the nominal gaps configuration. This paper aims to estimate the effect of the gap size variability on nuclear heating in the magnets during plasma operation and shut-down dose rate at the port interspace maintenance location taking advantage of a set of MCNP simulations with different gap configurations. The effect has been shown to produce increases of up to 30% in the worst-case configurations. The novel modelling approach, which was developed ad-hoc and used to manipulate the size of the port gaps via displacements of the geometry universes, is also described here. This approach led to important time savings, and it may be used in future studies with similar requirements. [ABSTRACT FROM AUTHOR]

The ITER Diagnostic, Heating and Current Drive systems might appear, on the face of it, to have very different control requirements. There are approximately 45 diagnostic systems, including magnetic sensors for plasma position and shape determination, imaging systems in the IR and visible, Thompson scattering for electron temperature and density, neutron detectors and collective scattering for alpha particle density and energy distribution. The H&CD systems encompass Electron Cyclotron Heating, using 24 1MW, 170 GHz gyrotrons and 5 steerable launchers to deliver 20 MW to the plasma, Ion Cyclotron Heating, using 8 3MW, 40–55 MHz sources and two multi-element launchers to deliver 20 MW to the plasma, and 2 Negative Ion Neutral Beam Injectors, each of which can deliver up to 16.5 MW of 1 MeV beams to the plasma. Although there are substantial differences in the needs for protection, when handling multi-MW heating systems, and in data throughput for many diagnostics, the formal processes needed to translate system requirements into Instrumentation and Control are identical. Due to the distributed procurement of ITER sub-systems and the need to integrate as painlessly as possible to CODAC, the formal processes, together with a substantial degree of standardization, are even more than usually essential. Starting from the technical, safety and protection, integration and operation requirements, a loop of functional analysis and signal listing is used to generate the controller configuration and the conceptual architecture. These elements in their turn lead to the physical and software design. The paper will describe the formal processes of control system design and the methods used by the ITER project to achieve the standardization of systems engineering practices. These have been applied to several use-cases covering all operation relevant phases such as plasma operation, maintenance, testing and conditioning. There are a number of running contracts that are developing the control designs, including on which is preparing detailed diagnostic use-cases to help designers out in the field. The ITER International Organization will reinforce the methodology by holding regular User Meetings, in which successes and problems can be shared. Also, the distribution of CODAC Cores Systems and the holding of Hands-On Workshops is intended to help users in the Domestic Agencies and their suppliers to buy in to the standardization and to build compatible I&C systems, which will integrate seamlessly into CODAC when they arrive on the ITER site. [ABSTRACT FROM AUTHOR]

A control system of a flowing liquid lithium (FLiLi) limiter for the Experimental Advanced Superconducting Tokamak (EAST) was developed and implemented. The control system is not only able to control the direct current (DC) electromagnetic pump and heating power but can also set scanning parameters, receive the shot number, acquire the temperature, etc. The system consists of multifunctional LAN eXtensions for Instrumentation (LXI) instrument, temperature-acquisition module, programmable DC power supply, and programmable logic controller (PLC). The multi-range DC power supply is programmed to meet the operational requirements of the DC electromagnetic pump. The LXI instrument and temperature-acquisition module are used to obtain temperature data. The PLC is adopted to control the temperature of the FLiLi limiter. A safety interlock and protection function was developed for the FLiLi limiter control system. The software was designed by using LabVIEW to achieve data interaction between multiple protocols. The FLiLi limiter control system can acquire experimental data at a speed of 100 S/s and store it for later analysis. The control system was successfully applied to a FLiLi limiter to study the interaction between plasma and a fixed wall in the EAST campaign. This paper presents the framework, the implementation details, and results of the control system. [ABSTRACT FROM AUTHOR]

The Steady State Superconducting Tokamak (SST-1) consists of many distributed and heterogeneous plant/experiment systems viz. Water-Cooling, Power Supplies, Cryogenics, Vacuum, Magnets, Auxiliary-Heating sources, Diagnostics, Front End Electronics (FEE) & Data Acquisition systems, having their own data acquisition & control systems and control & monitor by Central Control System (CCS) during the machine operation. With distributed computing and interdependent systems, it is essential that all the data/event acquired must be with disciplined & precise time-base, so as to make the co-relation of the data/event from various plant and experiment systems easy. Hence it is important to have accurate and precise Time Synchronization in place. The two systems fulfill the requirement of the time synchronization in SST-1. The VME based Timing System (TS) provides synchronization amongst various experiment systems during the plasma discharges and works as discharge control system (DCS) while the GPS based Time Synchronization System (TSS) caters the requirement of synchronization during the continuous operation of various plant systems by feeding a central clock to all the plant systems. This paper presents the Time Synchronization System of SST-1, the results of the integrated testing and engineering validation with various SST-1 subsystems. [ABSTRACT FROM AUTHOR]

This paper presents the comparison over the nuclear analysis of the European DEMO with HCLL blanket carried out with the TRIPOLI-4 ® Monte Carlo code and the JEFF-3.1.1 nuclear data library and with the MCNP5 Monte Carlo code and the FENDL-2.1 nuclear data library. The MCNP5 analysis was conducted firstly by ENEA with a detailed 3D model describing all the HCLL blanket internal structures. This MCNP5 model was converted into TRIPOLI-4 ® representation for performing the nuclear analysis at CEA with the objective to demonstrate consistency between both analyses. A very good agreement was obtained for all of the relevant nuclear responses (neutron wall loading, tritium breeding ratio, nuclear heating, neutron flux distribution, etc.), validating CEA’s nuclear analysis approach, based on TRIPOLI-4 ® Monte Carlo code and JEFF-3.1.1 nuclear data library, for the European DEMO. [ABSTRACT FROM AUTHOR]

• NBI Heating systems are known as the most efficient type of plasma heating of tokamak. • Ion production sources are known to be a major component of NBI heating systems. • The design and optimization of ion generation systems are very effective in upgrading NBI heating systems. Neutral beam injection (NBI) is known as one of the important methods of plasma auxiliary heating and current drive in fusion reactors. Optimization of the ion sources, as one of the main components of NBI systems, with high current and minimum divergence, can be very effective in order to reach more optimal dimensions of NBI systems. A high efficiency ion source designed using a helicon plasma source for neutral beam injection system (NBI) of Damavand Tokamak. The extraction system has the ability to extract a positive hydrogen ion beam with the current and voltage of 7A and 4.5 kV, respectively. In this paper, an ion source with optimized helicon plasma and optimized beam extraction system is developed using the COMSOL software. All parameters, including RF power, magnetic field, gas pressure, antenna length, and RF frequency, for the production of helicon plasma and ion beam extraction system with optimal geometry are investigated in the most effective way, in order to reduce the beam divergence and beam current increment. Finally, the helicon plasma with density of 4.12 × 10 18 m − 3 , electron temperature of 2.8eV and hydrogen beam extraction system consists of three Accel-Decel grid, each grid with 280 apertures of 3.5mm diameter, has been designed for neutral beam injection system of Damavand Tokamak. This ion production system has an extractor grid with very large transparency (60%) and the minimum single beamlet divergence of 27 mrad. [ABSTRACT FROM AUTHOR]

The lower carbon divertor has been used since 2008 in EAST, and many significant physical results, like the 410 s long pulse discharge and the 32 s H-mode operation, have been achieved. As the carbon divertor will still be used in the next few years while the injected auxiliary heating power would be increased gradually, it’s necessary to evaluate the operational capability of the carbon divertor under the heat loads during future operation. In this paper, an actual time-history heat load during type-I ELMy H-mode from EAST experiment, as one of the most severe operating environment in tokamak, has been used in the calculation and analysis. The finite element (FE) thermal and mechanical calculations have been carried out to analysis the stress and deformation of the carbon divertor during the heat loads. According to the results, the main impact on the overall temperature comes from the relative stable phase before and after the type-I ELMs and local peak load, and the transient thermal load such as type-I ELMy only has a significant effect on the surface temperature of the graphite tiles. The carbon divertor would work with high stress near the screw bolts in the current operational conditions, because of high preload and conservative frictional coefficient between the bolts and heatsink. For the future operation, new plasma facing materials (PFM) and divertor technology should be developed. [ABSTRACT FROM AUTHOR]

In the framework of the Broader Approach Agreement for the construction of the JT-60SA tokamak, ENEA provides 9 of the 18 toroidal field (TF) coils of the JT-60SA magnet system. The 9 coils are being manufactured by ASG superconductors in Genoa under the supervision of ENEA in collaboration with the JT-60SA European home team. The manufacturing is composed of two main steps: one concerning winding pack assembly and impregnation, and the other devoted to the integration into the casing structure and associated final coil preparation. This paper describes the results of the validation activities set-up for the integration phase. Specifically, welding of casing components has been retained particularly critical for at least three reasons: (i) during welding the WP may be damaged by the intense heating; (ii) distortion caused by heating may determine incorrect coil geometry and then field errors; and (iii) flaws may reduce structural strength and then the overall lifetime of the machine. Similarly, final embedding has been demonstrated on a 1 m long mock-up of the coil. Main results and lessons learned are here described. [ABSTRACT FROM AUTHOR]

SIRHEX (“Surface Infrared Radiation Heating Experiment”) is a small-scale experimental facility at KIT, which has been built for testing and qualifying high heat flux radiation heaters for blanket specific conditions using an instrumented water cooled target. This paper describes the SIRHEX facility and the experimental set-up for the heater tests. The results of a series of tests focused on reproducing homogeneous surface heat flux densities up to 500 kW/m 2 will be presented and the impact of the heater performance on the design of the First Wall test rig will be discussed. [ABSTRACT FROM AUTHOR]

The existing tritium resource available for future fusion power plant D-T operation is severely limited. It is essential to breed tritium to maintain the continuous consumption in the D-T plasma so as to sustain the required fusion power. A minimum criterion of calculated tritium breeding ratio (TBR) >1.1 is adopted to ensure tritium self-sufficiency, allowing a 10% safety margin to cover calculation uncertainties. In a DEMO reactor, there are various factors impacting the availability and effectiveness of fusion generated neutrons for production of tritium. In this paper, the study focused on the impact of first wall design including materials and first wall thickness, and blanket coverage aspects such as divertor geometry and the provision of equatorial heating and diagnostic ports. [ABSTRACT FROM AUTHOR]

For the ECRH heating system of the stellarator Wendelstein 7-X, two remote steering antennas are developed and manufactured. The principle of remote steering antennas is based on the imaging characteristics of corrugated rectangular waveguides, which is well understood and can accurately be simulated. Several details, however, require deeper investigation. The antenna needs a miter-bend and a 24 mm gap. The positions of these elements need to be chosen carefully to reduce losses and stray radiation. The antennas are manufactured from copper by electroforming. This allows to integrate all components, including the corrugated inner walls and the cooling channels, in one vacuum-tight piece. This paper reviews the design process of the remote steering antennas for W7-X as well as technological issues and experimental results from test pieces. [ABSTRACT FROM AUTHOR]

The ITER Blanket System and the Divertor are the main components which directly face the plasma. Being the first physical barrier to the plasma, they have very demanding design requirements, which include accommodating: (1) surface heat flux and neutronic volumetric heating, (2) electromagnetic loads, (3) nuclear shielding function, (4) capability of being assembled and remote-handled, (5) interfaces with other in-vessel components, and (6) high heat flux technologies and complex welded structures in the design. The main functions of the Blanket System have been substantially expanded and it has now also to provide limiting surfaces that define the plasma boundary during startup and shutdown. As regards the Divertor, the ITER Council decided in November 2013 to start the ITER operation with a full-tungsten armour in order to minimize costs and already gain operational experience with tungsten during the non-active phase of the machine. This paper gives an overview of the design and technology qualification of the Blanket System and the Divertor. [ABSTRACT FROM AUTHOR]

In ITER, each heating neutral beam injector (HNB) will deliver about 16.5 MW heating power by accelerating a 40 A deuterium negative ion beam up to the energy of 1 MeV. The ions are generated inside a caesiated negative ion source, where the injected H 2 /D 2 is ionized by a radio frequency electromagnetic field. The SPIDER test bed, currently being manufactured, is going to be the ion source test facility for the full size ion source of the HNBs and of the diagnostic neutral beam injector of ITER. The SPIDER beam source comprises an ion source with 8 radio-frequency drivers and a three-grid system, providing an overall acceleration up to energies of about 100 keV [1] . SPIDER represents a substantial step forward between the half ITER size ion source, which is currently being tested at the ELISE test bed in IPP-Garching, and the negative ion sources to be used on ITER, in terms of layout, dimensions and operating parameters. The SPIDER beam source will be housed inside a vacuum vessel which will be equipped with a beam dump and a graphite diagnostic calorimeter. The manufacturing design of the main parts of the SPIDER beam source has been completed and many of the tests on the prototypes have been successfully passed. The most complex parts, from the manufacturing point of view, of the ion source and the accelerator, developed by galvanic deposition of copper are being manufactured. The manufacturing phase will be completed within 2015, when the assembly of the device will start at the PRIMA site, in Padova (I). The paper describes the status of the procurement, the adaptations operated on the design of the beam source for the fabrication, with particular emphasis to the engineering development to enable the fulfillment of the tight requirements set in the technical specifications. Moreover the tests performed on the prototypes are reported. [ABSTRACT FROM AUTHOR]

MITICA is a prototype of the heating neutral beam (HNB) Injectors for ITER, built with the purpose of validating the injector design and optimizing its operation. Its goal is to produce a focused beam of neutral particles (H or D) with energy up to 1 MeV and power of 16 MW for 1 h. MITICA includes a Radio Frequency (RF) Plasma Source for the production of negative ions, a multi-stage electrostatic accelerator (up to 1 MV and 40 A), a neutralizer, a residual ion dump and a calorimeter. A transverse magnetic field in the Ion source and accelerator, including both a long-range component and a local component is crucial for obtaining the required Ion current and accelerator efficiency. The long-range component is produced by the current flowing through the plasma grid (PG) and related bus-bars. The PG current distribution and the uniformity of the resulting magnetic field have been optimized by detailed finite element (FEM) models. Hollow volumes in the thick copper part of the PG among beamlet groups allow a more uniform PG current distribution and a consequently uniform magnetic field in front of the grid. The paper describes in detail the PG geometry optimization procedure and the related magnetic and thermo-structural FEM analyses. [ABSTRACT FROM AUTHOR]

The Duct Liner (DL) for the ITER Heating Neutral Beam (HNB) is a key component in the beam transport system. Duct Liners installed into equatorial ports 4 and 5 of the Vacuum Vessel (VV) will protect the port extension from power deposition due to re-ionisation and direct interception of the HNB. Furthermore, the DL contributes towards the shielding of the VV and superconducting coils from plasma photons and neutrons. The DL incorporates a 316L(N)-IG, deep-drilled and water cooled Neutron Shield (NS) whose internal walls are lined with actively cooled CuCrZr Duct Liner Modules (DLMs). These Remote Handling Class 2 and 3 panels provide protection from neutral beam power. This paper provides an overview of the preliminary design for the ITER HNB DL and focusses on critical features that ensure compatibility with: high heat flux requirements, remote maintenance procedures, and transient magnetic fields arising from major plasma disruptions. The power deposited on a single DLM can reach 300 kW with a peak power density of 2.4 MW/m 2 . Feeding coolant to the DLMs is accomplished via welded connections to the internal coolant network of the NS. These are placed coaxially to allow for thermal expansion of the DLMs without the use of deformable connections. Critically, the remote maintenance of individual DLMs necessitates access to water connections and bolts from the beam facing surface, thus subjecting them to high heat flux loads. This design challenge will become more prevalent as fusion devices become more powerful and remote handling becomes a necessity. The novel solutions implemented to overcome this are detailed and their performance scrutinised. The designs presented include tungsten caps that protect bolted and remotely welded connections by radiating heat away, and explosion bonded CuCrZr/Stainless Steel panels designed to reduce by a factor of 10 the eddy current torques caused by plasma disruptions. [ABSTRACT FROM AUTHOR]

The neutral beam system for ITER consists of two heating and current drive injectors plus a diagnostic neutral beam injector. The proposed physical plant layout allows for a possible third heating injector to be installed later. For correct operation of the beam source, and to avoid deflections of the charged fraction of the beam, the magnetic field along the beam path must be very low. To minimize the stray ITER field in critical areas (ion source, acceleration grids, neutralizer, residual ion dump), a Magnetic Field Reduction System will envelop the beam vessels and the high voltage transmission lines to ion source. This whole system comprises the Passive Magnetic Shield, a set of thick steel plates, and the Active Correction and Compensation Coils, a set of coils carrying currents which depend on the tokamak stray field. This paper describes the status of the coil design, terminals and support structures, as well as a description of the calculations carried out. Most coils are suitable for removal from their final position to be replaced in case of a fault. Conclusions of the chosen design highlight the strategy for the system feasibility. [ABSTRACT FROM AUTHOR]

The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components, which in turn is related to the amount of mobilised dust present in the Vacuum Vessel. Periodically or on request, the IVVS scanning probes will be deployed into the Vacuum Vessel in order to acquire both visual and metrological data on plasma facing components (blanket, divertor, heating/diagnostic plugs, and test blanket modules). Recent design changes made to the six IVVS port extensions implied the need for a substantial redesign of the IVVS integrated concept – including the scanning probe and its deployment system – in order to bring it to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the concept design for IVVS as well as of the various engineering analyses and R&D activities carried out in support to this design: neutronic, seismic and electromagnetic analyses, probe actuation validation under environmental conditions. [ABSTRACT FROM AUTHOR]

The Karlsruhe Advanced Technologies Helium Loop (KATHELO) has been designed for testing divertor modules as well as qualifying materials for high heat flux, high temperature (up to 800 °C) and high pressure (10 MPa) applications. The test section inlet temperature level is controlled using a process electrical heater. To cope with the extreme operating conditions, a special design of this unit has been proposed. In this paper the conceptual design of the unit will be presented and the impact of the coupling between the cold and hot helium gas on the overall efficiency of the loop will be investigated. The detailed thermal-hydraulic analysis of the feed through of the hot helium into the low temperature pressure vessel using ANSYS CFX will be presented. The impact of the design choices on the overall energy budget of the loop will be analyzed using RELAP5-3D. [ABSTRACT FROM AUTHOR]

The paper summarizes the current status of neutronics at ITER and a first set of proposals for experimental programmes to be conducted in the early operational life-time of ITER are described for the more crucial areas. These include a TF coils heating benchmark, a streaming benchmark and streaming measurements by activation on ITER itself. Also on ITER the measurement of activated water from triton burn-up should be planned and performed. This will require the measurement of triton burn-up in DD phase. Measurements of neutron flux in the tokamak building during DD operations should also be carried out. The use of JET for verification of shut down dose rate estimates is desirable. Other facilities to examine the production and behaviour of activated corrosion products and the shielding properties of concretes to high energy (6 MeV) gamma-rays are recommended. [ABSTRACT FROM AUTHOR]

The DEMO Active Maintenance Facility (AMF) would be used for the storage, handling and processing of In-Vessel Components (IVC) throughout their time on site, the only exception being the time that they are installed in the vessel. It is anticipated that all handling operations associated with used components will have to be carried out using remote handling techniques. During plasma operations the In-Vessel Components are exposed to high levels of neutron activation. This activation results in high radiation dose rates and decay heating. This presents a significant problem for Remote Handling Equipment (RHE) in the AMF. The high dose rates require the equipment to be sufficiently radiation tolerant to allow it to work reliably for long periods. The decay heating requires forced cooling of newly removed IVC's while they are in storage. The duration of the storage is dependent on the decay heating reducing to a level that has been nominally set at <50 °C without active cooling in room temperature air. This paper summarises the progress made in 2012 on the conceptual design of the AMF and its facilities. The layout and proposed function of the main areas will be described along with the principles applied. The design of the AMF has evolved from a simple representation of the required facilities in 2011 to a concept that can be developed to support maintenance of DEMO. [ABSTRACT FROM AUTHOR]

• The theory of the measurement of the diamagnetic flux in EAST is introduced in this paper. • A good response between the diamagnetic flux measurements and rapid stored energy drop induced by ELMs. • Provide a simple and reliable method to study the plasma anisotropy in EAST. The measurement of the diamagnetic flux for poloidal beta and energy content of EAST based on diamagnetic loops is described in this article. The error sources can be well compensated with the diamagnetic loop and the compensation loop signals. And the maximum error of vacuum flux under experiments without plasmas is about 0.11 mWb, which meet the typical experimental requirements for high β p discharges (β p ∼ 2.5). For ohmic discharges, the stored energy determined from the diamagnetic flux measurement W d i a is in good agreement with that from the equilibrium reconstruction W m h d. After the auxiliary heating power are injected into plasmas, the increment of the energy content is observed to verify the heating effects of different heating systems. With the upgrade of 2021, the recent campaign has shown a good response between the diamagnetic flux measurements and rapid stored energy drop induced by edge localized modes (ELMs). The preliminary experimental results indicate that plasma anisotropic degree decreases with increasing density under operation conditions with high density and high heating power. [ABSTRACT FROM AUTHOR]

Abstract: The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs. [Copyright &y& Elsevier]

Abstract: In Europe the work on the specification and design of a demonstration power plant (DEMO) is being carried out by EFDA in the power plant physics and technology (PPP&T) programme. DEMO will take fusion from experimental research into showing the potential for commercial power generation. During the fusion reaction, components in the tokamak become highly activated. The estimated dose rate levels after shutdown (zero decay time) due to 60dpa accumulation in steel (blanket) and 30dpa (divertor) are 13.1–17.4kGy/h (blanket); 8.8–11.6kGy/h (divertor) [1], much higher than those to be encountered at ITER. Upon removal from the tokamak, components would be transported to the hot cell facility with attention to minimizing the spread of activated dust and tritium contamination. It is proposed to use a sealed cask of ∼20tonnes, running on air castors with 50% lifting capability redundancy. Due to the number and complexity of the routes taken by this transporter it would have to be an un-tethered semi-autonomous system. This poses some technical challenges, including providing sufficient battery capacity, reliable guidance and a fail safe un-tethered control system. The mass of the components being moved is assumed here to range from a few tonnes to in excess of one hundred tonnes. Before the removed in-vessel components can be processed in the hot cell, they would require a period of cooling, approximately 2.5 years, to allow dose rate and decay heating to reduce. This reduces the decay heating level to ∼1–1.5kW/m3 and a contact dose rate to ∼250Sv/h, which is more suited to dexterous man-in-the-loop remote handling (RH). During maintenance, many components would require replacement or refurbishment. The hot cell facility would have to provide all the associated RH functions and operate fully remotely. With no human intervention, an accordingly robust RH recovery system would be required. This paper describes the first steps being taken towards the design of the DEMO hot cell. It will show a comparison of the current DEMO in-vessel maintenance concepts from a hot cell perspective, describe a proposed ex-vessel transport system, and summarize the facilities that have been identified as required within the hot cell, examine current RH technology and discuss the identified critical development issues. [Copyright &y& Elsevier]

Abstract: The ITER neutral beam cell will contain up to three heating neutral beams and one diagnostic neutral beam, and four upper ports. Though manual maintenance work is envisaged within the cell, when containment is breached, or the radiological protection is removed the maintenance must be conducted remotely. This maintenance constitutes the removal and replacement of line replaceable units, and their transport to and from a cask docked to the cell. A design of the remote handling system has been prepared to concept level which this paper describes including the development of a beam line transporter, beam source remote handling equipment, upper port remote handling equipment and equipment for the maintenance of the neutral shield. This equipment has been developed complete the planned maintenance tasks for the components of the neutral beam cell and to have inherent flexibility to enable as yet unforeseen tasks and recovery operations to be performed. [Copyright &y& Elsevier]

Abstract: The ITER neutral beam (NB) injectors are used for heating and diagnostics operations. There are 4 injectors in total, 3 heating neutral beam injectors (HNBs) and one diagnostic neutral beam injector (DNB). Two HNBs and the DNB will start injection into ITER during the hydrogen/helium phase of ITER operations. A third HNB is considered as an upgrade to the ITER heating systems, and the impact of the later installation and use of that injector have to be taken into account when considering the installation and assembly of the whole NB system. It is assumed that if a third HNB is to be installed, it will be installed before the nuclear phase of the ITER project. The total weight of one injector is around 1200t and it is composed of 18 main components and 36 sets of shielding plates. The overall dimensions are length 20m, height 10m and width 5m. Assembly of the first two HNBs and the DNB will start before the first plasma is produced in ITER, but as the time required to assemble one injector is estimated at around 1.5 year, the assembly will be divided into 2 steps, one prior to first plasma, and the second during the machine second assembly phase. To comply with this challenging schedule the assembly sequence has been defined to allow assembly of three first injectors in parallel. Due to the similar design between the DNB and HNBs it has been decided to use the same tools, which will be designed to accommodate the differences between the two sets of components. This reduces the global cost of the assembly and the overall assembly time for the injector system. The alignment and positioning of the injectors is a major consideration for the injector assembly as the alignment of the beamline components and the beam source are critical if good injector performance is to be achieved. The theoretical axes of the beams are defined relative to the duct liners which are installed in the NB ports. The concept adopted to achieve the required alignment accuracy is to use the main rail of the overhead crane associated with offset tooling when necessary. The overhead crane is used for the assembly of the components, and the final positioning of the beamline components and the beam source will be adjusted with respect to laser targets referring to the optimum beam axis and source position. This paper describes the installation tasks and the alignment and positioning solutions and the complexity of operations within the NB cell. Particular constraints on the HNB installation sequence due to the planned testing of the 1MV high voltage supply are also described. [Copyright &y& Elsevier]

Abstract: The In-Vessel Viewing and Metrology System (IVVS) is a fundamental tool for the ITER machine operations, aiming at performing inspections as well as providing information related to the erosion of in-vessel components. Periodically or on request, the IVVS probes will be deployed into the Vacuum Vessel from their storage positions (still within the ITER primary confinement) in order to perform both viewing and metrology on plasma facing components (blanket, divertor, heating/diagnostic plugs, test blanket modules) and, more generically, to provide information on the status of the in-vessel components. In 2011, the IO proposed to simplify and strengthen the six IVVS port extensions situated at the divertor level. Among other important consequences, such as the relocation of the Glow Discharge Cleaning (GDC) electrodes at other levels of the machine, this major design change implied the need for a substantial redesign of the IVVS plug, which took part to an on-going effort to bring the integrated IVVS concept – including the scanning probe and its deployment system – to the level of maturity suitable for the Conceptual Design Review. This paper gives an overview of the various design and R&D activities in progress: plug design integration, probe concept validation under environmental conditions, development of a metrology strategy, the whole supported by a nuclear analysis. [Copyright &y& Elsevier]

Abstract: This paper deals with Remote Handling activities foreseen on the Blanket Shield Module, the plasma facing component of the ITER Electron Cyclotron Heating and Current Drive Upper Launcher. The maintenance configuration considered here is the Vertical Remote Handling, meaning gravity acting along the launcher radial axis. The plant, where the maintenance under consideration is occurring, is the Hot Cell Facility Work Cell. The study here reported has been carried out within the presently ongoing EFDA Goal Oriented Training program on Remote Handling (GOT-RH), which aims to support ITER activities. This document and its contents have to be considered as part of a more vast RAMI analysis to be developed within the GOT-RH, which aims to maximize the Electron Cyclotron Heating and Current Drive system availability. The Baseline CAD model of the Electron Cyclotron Heating and Current Drive Upper Launcher is currently in its preliminary design phase and does not provide enough details for developing a fully detailed maintenance strategy. Therefore, through a System Engineering approach, a set of assumptions was conceived on the launcher structure, as a basis for development of a Remote Handling strategy. Moreover, to compare different design solutions related to the possibility of integrating a quasi-optical component into the Blanket Shield Module, a Trade-Off was made, and its contents are shown here. The outcome of this System Engineering approach has been formalized into Task Definition Forms whose contents are reported here. The Remote Handling strategy presented in this work will be tested in the near future both through Virtual Reality simulations and through prototype experiments. [Copyright &y& Elsevier]

Highlights: [•] The CODAC system is a distributed system and scattered in many buildings connected with cables. Building construction for ITER project is just started so volume reservation for different CODAC component is done based on the content of this paper. [•] The 2-D and 3-D diagrams have been prepared which are showing the location of different CODAC equipment and their connectivity. [•] The different requirements regarding raised access (false) floor, power requirements, heat load, installation, maintenance, redundancy, and segregation etc. have been considered. [Copyright &y& Elsevier]

• Indigenously development of mock-up ICRH components. • Capable to automatically match any arbitrary load (VSWR from 1.5 to 4). • Matching technique with developed algorithm based on VSWR. • Faster speed response. This paper presents the development of a mock-up Ion Cyclotron Resonance Heating (ICRH) system along with a new algorithmic approach for fast automatic impedance matching. The mock-up is developed for low power (few watts) and frequency (182.5 MHz) i.e. five times the ICRH frequency of SST-1. The size of the mock-up is considerably reduced to be fit on the test bench using it. The main objective of the work is to develop different matching approaches in order to get the optimum speed of matching. Reducing the size of the mock-up allows frequent implementation and testing of new algorithms which is generally difficult to implement in the real system of tokamak because of its size and complexity. The system consists of indigenously developed components such as 3 dB hybrid coupler, 20 dB directional coupler, tapped transmission line, rigid coaxial stub tuners, line stretcher, service stub, RF antenna, movable water tank based load. The developed system is capable of automatically match any arbitrary load in the range, (35–260) Ω , resistive + (-j25 to +j220) Ω , reactive or VSWR in the range 1.5–4.0 within 200 ms. The matching system also consists of power detectors which automatically sense the forward and reflected power using coupler and the feedback controlled moving coaxial stubs, line stretchers respond as per developed algorithms. The matching algorithm incorporates the various matching steps based on VSWR which have been interpreted for the controller program to control the devices used in the matching network. The developed system is tested using an RF antenna facing the movable saltwater tank that emulates the plasma-like loading behavior. The developed setup is also useful for various high power RF applications like RF heating, sputtering, plasma generation, RADAR, etc. in the RF frequency range. [ABSTRACT FROM AUTHOR]