Design status of the ITER ECRH upper launcher mm-wave system

The purpose of the ITER electron cyclotron resonance heating (ECRH) upper launcher (UL), or antennae will be to provide localised current drive by accurately directing mm-wave beams up to 2MW, out of the four allocated upper port plugs, at chosen rational magnetic flux surfaces in order to stabilise neoclassical tearing modes (NTMs). This paper will present an overview of the UL, with emphasis on the mm-wave components. The mm-wave layout includes corrugated waveguide sections and a quasi-optical path with both focusing mirrors and plane steering mirrors. One of the essential components of the UL is the Steering Mechanism Assembly (SMA), providing variable poloidal injection angles fulfilling high deposition accuracy requirements at the plasma location. The Actuator principle and rotor bearings are frictionless and backlash free, avoiding tribological difficulties such as stickslip and seizure. The underlying working principle is the use of mechanically compliant structures. Validation and proof testing of the steering principle is achieved with an uncooled first prototype demonstrator. A second prototype is currently being manufactured, comprising the functionalities needed for the ITER compatible system such as water cooling and high power mm-wave compatibility. In order to perform the fatigue tests of the actuator bellows, a test facility has been built, under ITER-like vacuum and temperature working conditions. Results of the cyclic fatigue tests are compared to the various manufacturer standards and codes, combining stress and strain controlled material fatigue properties.     

A remotely steered millimetre wave launcher for electron cyclotron heating and current drive on ITER

High-power millimetre wave beams employed on ITER for heating and current drive at the 170 GHz electron cyclotron resonance frequency require agile steering and tight focusing of the beams to suppress neoclassical tearing modes. This paper presents experimental validation of the remote steering (RS) concept of the ITER upper port millimetre wave beam launcher. Remote steering at the entrance of the upper port launcher rather than at the plasma side offers advantages in reliability and maintenance of the mechanically vulnerable steering system. A one-to-one scale mock-up consisting of a transmission line, mitre bends, remote steering unit, vacuum window, square corrugated waveguide and front mirror simulates the ITER launcher design configuration. Validation is based on low-power heterodyne measurements of the complex amplitude and phase distribution of the steered Gaussian beam. High-power (400 kW) short pulse (10 ms) operation under vacuum, diagnosed by calorimetry and thermography of the near- and far-field beam patterns, confirms high-power operation, but shows increased power loss attributed to deteriorating input beam quality compared with low-power operation. Polarization measurements show little variation with steering, which is important for effective current drive requiring elliptical polarization for O-mode excitation. Results show that a RS range of up to −12° to +12° can be achieved with acceptable beam quality. These measurements confirm the back-up design of the ITER ECRH&CD launcher with future application for DEMO.     

Shut-Down Dose Rate analysis for ITER Diagnostic Equatorial and Upper Ports

The Shut-Down Dose Rate (SDDR) is an important criterion of radiation safety for the personnel access for maintenance operations in ITER ports after the cessation of the D-T 14 MeV neutron fusion source. Therefore, the problem of the SDDR calculations attracts the attention of fusion neutronics community because SDDR in such a large and geometrically complicated fusion device as the ITER tokamak is challenging to compute. This challenge has been faced and overcome by applying dedicated methodological approaches explained in this paper. The results of the SDDR analysis allowed us to propose several design solutions for improvement of the radiation shielding of the ITER Generic Diagnostic Equatorial and Upper Port Plugs (EPP and UPP). The SDDR analysis was focused on the interspace area located between the ITER bioshield and port plugs where the personnel access is envisaged at ∼12 days after the ITER shut-down. By this analysis the radiation streaming pathways and dominant sources of decay radiation were revealed and the methods to mitigate the streaming and subsequent activation were found. The optimization of the port shielding was targeted on minimization of the SDDR in the interspace area following the ALARA principle and taking into account the feasibility to implement proposed shielding options with the actual hardware. Among them, wrapping the EPP walls with the B₄C tiles improves the EPP shielding performance. While void around the ELM/in-vessel coils and blanket manifolds leads to the performance reduction. The SDDR inside the Generic UPP interspace depends mainly on the environment (blanket, manifolds, and gaps).     

Gamma-ray dose analysis for ITER JA WCCB-TBM

To evaluate the nuclear properties of the International Thermonuclear Experimental Reactor (ITER) JA Water-Cooled Ceramic Breeder Test Blanket Module (WCCB-TBM) and to ensure its design conforms to nuclear licensing regulations, nuclear analyses have been performed for the WCCB-TBM's components, including its frame, shield, flange, port extension, pipe forest, bio-shield and Ancillary Equipment Unit (AEU). Utilising Monte Carlo code MCNP5.14, activation code ACT-4 and the Fusion Evaluated Nuclear Data Library FENDL-2.1, this paper focusses on the shutdown dose rate calculation for the WCCB-TBM. Monte Carlo N-Particle Transport Code (MCNP) geometry input data for the TBM are created from computer-aided design (CAD) data using the CAD/MCNP automatic conversion code GEOMIT, and other geometry input data are created manually. The ‘Direct 1-Step Monte Carlo’ method is adopted for the decay gamma-ray dose rate calculation. Behind the bio-shield, the effective dose rates 1 day after shutdown are about 0.2 μSv h⁻¹, which are much lower than 10 μSv h⁻¹, the upper limit for human access. Behind the flange, the effective dose rates 10⁶ s after shutdown are 50–80 μSv h⁻¹, which are lower than 100 μSv h⁻¹, the upper limit for human hands-on access for workers performing maintenance.     

Benchmarking the MCR2S system for high-resolution activation dose analysis in ITER

The MCR2S system is a tool for the analysis of activation dose in ITER and fusion devices based on the rigorous-2-step formalism, MCNP mesh-tallies and the FISPACT nuclear inventory code. It couples transport and activation calculations in an integrated manner, allowing for the necessarily fine energy and spatial resolutions required for this kind of analysis and overcoming limitations of previous similar tools. The system is routinely used at UKAEA to assist the design of several ITER components such as the ICRH heating system and LIDAR diagnostic. Here we describe its functionality and a first benchmark exercise performed using the fusion-relevant Frascati Neutron Generator shutdown dose rate experiment results. Neutron spectra, reaction rates, photon spectra and activation decay dose rates at several positions and times, obtained using FENDL-2.1/MC and EAF-2007 data, are presented and compared to those of earlier software and libraries, showing good agreement. C/E values close to unity are obtained in all cases, illustrating the robustness and reliability of this novel software.     

Gripping tool for the ITER upper port plug RH extraction/insertion

The conceptual design of several gripping tools and their mechanical interfaces is being carried out for the ITER ECH UPP within the WP10-GOTRH programme. EFDA finances the GOT RH (Goal Oriented Training Programme for Remote Handling). The purpose of this paper is to introduce new concepts of gripping tools for the plug extraction/insertion in the upper port of ITER. All these gripping tools are designed according to IO input data and geometrical constraints. The gripping tools have to be able to extract/insert the plug in the scenario of maximum misalignment between the plug and the tractor. The paper also defines the functional requirements the gripping tools need to comply with. The requirements and input data are verified and validated through 3D simulation with Catia mock-ups of the gripping tools. The strengths and weaknesses of each gripping tool model are compared.     

ITER上窗口屏蔽中子学分析研究

利用CAD/MCNP自动建模程序MCAM建立ITER新上窗口中子学计算模型,使用中子/光子耦合输运程序MCNP/4CI、AEA聚变核数据库FENDL1.0和集成上窗口模型的ITER基本中子学模型计算并分析上窗口新的工程设计的屏蔽能力以检验设计的合理性。结果表明,与以前的上窗口设计相比,新设计的上窗口的周围剂量控制点的快中子注量率、停堆剂量率以及线圈核热等都增大了好几倍,建议进一步改进上窗口设计。     

Shutdown dose rate assessment with the Advanced D1S method: Development,applications and validation

The present paper addresses the recent developments and applications of Advanced-D1S to the calculations of shutdown dose rate in tokamak devices. Results of benchmarking with measurements and Rigorous 2-Step (R2S) calculations are summarized and discussed as well as limitations and further developments. The outcomes confirm the essential role of the Advanced-D1S methodology and the evidence for its complementary use with the R2Smesh approach for the reliable assessment of shutdown dose rates and related statistical uncertainties in present and future fusion devices.     

Optimized mass flow rate distribution analysis for cooling the ITER Blanket System

This paper presents the rationale to the optimization of water distribution in ITER blanket modules, meeting both Blanket System requirements and interface compliance requirements.The key challenging constraints include to: be compatible with the overall water allocation (3140 kg/s for 440 wall mounted BMs); meet the critical heat flux margin of 1.4 in the plasma facing units; meet a maximum temperature increase of 70 °C at the outlet of each single BM; and ensure that water velocity is less than 7 m/s in all manifolds, and that the pressure drops of all BMs can be equilibrated. The methodology and the successful result are presented.     

Conceptual design of the ECH upper launcher system for ITER

The challenge of developing the conceptual design of the ECH Upper Launcher system for MHD control in the ITER plasmas has been tackled by team of European Associations together with the European Domestic Agency (“F4E”). The launcher system has to meet the following requirements: (a) a mm-wave system extending from the interface to the transmission line up to the target absorption zone in the plasma and performing as an intelligent antenna; (b) a structural system integrating the mm-wave system and ensuring sufficient thermal and nuclear shielding; (c) port plug remote handling and testing capability ensuring high port plug system availability. The paper describes the reference launcher design. The mm-wave system is composed of waveguide and quasi-optical sections with a front steering system. An automated feedback control system is developed as a concept based on an assimilation procedure between predicted and diagnosed absorption location. The structural system consists of the blanket shield module, the port plug frame, and the internal shield for appropriate neutron shielding towards the launcher back-end. The specific advantages of a double walled structure are discussed with respect to adequate baking, to rigidity towards launcher deflection under plasma-generated loads and to removal of thermal loads, including nuclear ones. Basic studies of remote handling (RH) to validate design development are initiated using a virtual reality simulation backed by experimental validation, for which a launcher handling test facility (LHT) is set up as a full scale experimental site allowing furthermore thermohydraulic studies with ITER blanket water parameters.     

Prototyping of the Blanket Shield Module for the ITER EC H&CD Upper launcher

The design of the ITER Electron Cyclotron Heating and Current Drive (ECH&CD) Upper launcher is recently in the first of two final design phases. The first phase deals with the finalization of all FCS (First Confinement System) components as well as with specific design progress for the remaining In-vessel components.The most outstanding structural In-vessel component of an ECH&CD Upper launcher is the Blanket Shield Module (BSM) with the First Wall Panel (FWP). Both of them form the plasma facing part of the launcher, which has to meet strong demands on dissipation of nuclear heat loads and mechanical rigidity. Nuclear heat loads from 3 MW/m³ at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5 m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5 MW/m² due to plasma radiation must be dissipated from the FWP.To guarantee save and homogenous removal of such extensive heat loads, the BSM is designed as a welded steel-case with specific cooling channels inside its wall structure. Attached to its face side is the FWP with a high-power cooling structure.Based on computational analysis the optimum cooling channel geometry has been investigated. Specific pre-prototype tests have been made and associated assembly parameters have been determined in order to identify optimum manufacturing processes and joining techniques, which guarantee a robust design with maximum geometrical accuracy.This paper describes the design, manufacturing and testing of a full-size mock-up of the BSM. The study was carried out in an industrial cooperation with MAN Diesel and Turbo SE.     

A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas. The shield module is made of stainless steel plates and water channels. It is identified as an important component for radiation protection because of its radiation exposure control functionality. The radiation protection classification leads to more assurance of the component design. In order to validate and verify the design of the shield module, a neutronic laboratory-scale experiment is designed and executed. The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 10 10 ns ⁻¹. The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials. The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene), respectively. The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 cross-section data. The paper covers neutronic design, irradiation and the outcome of the experiment in detail.     

Design of a high power millimeter wave launcher for EC H&CD system on ITER

An optimization procedure of the quasi-optical system for a millimeter wave launcher is developed for the ITER electron cyclotron heating and current drive (EC H&CD) launcher. In the launcher, the radiated RF beams from eight corrugated waveguides are reflected sequentially by two mirrors and injected into a plasma through a small aperture in the blanket shield module on the top of the launcher. Using a steepest decent method, the heat load on the mirrors is successfully reduced to the acceptable level by flattening the RF power profile on the mirrors keeping the scattering of the RF power to a minimum from the mirrors. It is found that 20 MW injection will be acceptable even when the resistivity of 2.64 × 10⁻⁸ Ωm for the surface of the mirror (dispersion strengthened copper, 151 °C assumed) is increased by a factor of ∼10 with a contamination.     

Maintenance implications of critical components in ITER CXRS upper port plug design

Already in the early phase of a design for ITER, the maintenance aspects should be taken into account, since they might have serious implications. This paper presents the arguments in support of the case for the maintainability of the design, notably if this maintenance is to be performed by advanced remote methods. This structure is compliant to the evolving maintenance strategy of ITER. Initial results of a Failure Mode Effects and Criticality Analysis (FMECA) and a development risk analysis for the ITER upper port plug #3, housing the Charge Exchange Recombination Spectroscopy (CXRS) diagnostic, are employed for the definition of the maintenance strategy.The CXRS upper port plug is essentially an optical system which transfers visible light from the plasma into a fiber bundle. The most critical component in this path is the first mirror (M1) whose reflectivity degrades during operation due to deposition and/or erosion dominated effects. Amongst other measures to mitigate these effects, the strategy is to allow for a replacement of this mirror. Therefore it is mounted on a retractable central tube. The main purpose of this tube is to make frequent replacements possible without hindering operation. The maintenance method in terms of time, geometry and spare part policy has a large impact on cost of the system and time usage in the hot cell.Replacement of the tube under vacuum and magnetic field seems infeasible due to the operational risk involved. The preferred solution is to have a spare tube available which is replaced in parallel with other maintenance operations on the vessel, as to avoid any interference in the hot cell with the shutdown scheduling. This avoids having to refurbish a full port plug and also allows for a more frequent replacement of M1, as we can replace the mirror anytime the vacuum vessel is vented, estimated to be once a year.     

Structural analysis and optimization for ITER upper ELM coil

ITER ELM coils are used to mitigate or suppress Edge Localized Modes (ELM), which are located between the vacuum vessel (VV) and shielding blanket modules and subject to high radiation levels, high temperature and high magnetic field. These coils shall have high heat transfer performance to avoid high thermal stress, sufficient strength and excellent fatigue to transport and bear the alternating electromagnetic force due to the combination of the high magnetic field and the AC current in the coil. Therefore these coils should be designed and analyzed to confirm the temperature distribution, strength and fatigue performance in the case of conservative assumption. To verify the design structural feasibility of the upper ELM coil under EM and thermal loads, thermal, static and fatigue structural analysis have been performed in detail using ANSYS. In addition, design optimization has been done to enhance the structural performance of the upper ELM coil.     

Design optimisation of the ITER divertor magnetic probes using FEM analyses

This paper is focused on the design, simulation and optimisation of the ITER divertor magnetic tangential coils. The most critical issue for the divertor coils is to minimise RITES [G. Vayakis, et al., Radiation-induced thermoelectric sensitivity (RITES) in ITER prototype magnetic sensors, Rev. Sci. Instrum. 75 (10) (2004) 4324-4327] and TIEMF [R. Vila, E.R. Hodson, Thermally induced EMF in unirradiated MI cables, J. Nucl. Mater. 367-370 (Part 2) (2007) 1044-1047] by combining a proper choice of conductor with low temperature variation in the coil. Instead of mineral insulated cable (MIC), which was foreseen as the preferred winding, a winding made of ceramic-coated steel wire was recently proposed [G. Chitarin, L. Grando, S. Peruzzo, C. Tacconet, Design developments for the ITER in-vessel equilibrium Halo current sensors, 24th SOFT Conference, Warsaw, Poland, September 2006, Fusion Eng. Design, in press]. It is thought that, for this wire, maintaining a temperature variation in the wiring below 10 K will be sufficient to allow long-pulse operation. Variations of the divertor coil design have been simulated with the help of ANSYS. The aim was to keep the temperature variation in the winding pack within this limit. The optimisation of the coil based only on a cooling by conduction was not sufficient to meet the 10 K target. Therefore, an actively water-cooled coil was designed which finally met these requirements.     

Structural analyses for equatorial and lower ports of the ITER vacuum vessel

To investigate the structural integrity of the ITER vacuum vessel (VV) and ports, the structural analyses of the regular equatorial and the lower remote handling (RH) ports have been performed. The advanced design of the equatorial regular port adopting a pure friction type flange has been recommended as a reference design by the ITER International Organization. The structural integrity of the equatorial port flange, sealing unit, and connecting duct has been reviewed by conducting nonlinear finite element analyses. The advanced design of the regular equatorial port flange with proper pretension is acceptable in the structural design point of view.From the local analyses for a connecting duct and a sealing unit, it has been found that the stresses are less than the allowable values.The structural analyses of the lower RH port have been also performed to verify the capability for supporting the VV. Since high local stress occurs at the gusset and supporting block, the case study for the lower port has been conducted to mitigate the stress concentration and to modify the component design. The strength of the lower RH port structures can be improved by the design modification of poloidal and toroidal gusset.     

Thermal hydraulic feasibility analysis of the IBED PHTS for ITER

One of the main challenges of the ITER fusion reactor is to effectively remove large amount of heat deposited to the surface of the plasma facing components. The tokamak cooling water system (TCWS) will accomplish the objective of removing about 1 GW of peak heat load from in-vessel components while maintaining pressures and temperatures of the coolant within acceptable and safe limits during different operational scenarios. A study of feasibility has been launched for the IBED PHTS (Integrated Blanket, Edge localized mode coils (ELMs) and Divertor Primary Heat Transfer System; it consists of five independent cooling trains (four operational and one in stand-by), one steam pressurizer, supply and return headers, ring manifolds and connections to the all in-vessel components (i.e. First Wall Blanket, Divertor, ELM, Diagnostics and other Ports clients).The dynamic behaviour of the IBED PHTS has been investigated by means of RELAP5^® code to simulate the response of the system during plasma pulse and baking operations. Due to the plasma heat deposition on the surfaces of the in-vessel components and subsequent increase in hot leg temperature, a large amount of water volume is transferred from the hot legs of the circuit to the surge-line of the pressurizer during each burn cycle. This causes rapid increase of pressure and temperature of the system and the following actions are proposed to counteract these variations: spray injection in the upper dome of the pressurizer from the Chemical and Volume Control System (CVCS) to reduce the pressure and active control of flow rates through heat exchangers and their bypass loops to regulate the heat transfer from the primary system to the environment via secondary and tertiary loops.This paper focuses on the prediction of the thermal hydraulic behaviour of the IBED PHTS during plasma pulses and baking scenarios, describing the various activity of the analysis, the geometrical assessment of the circuit and the modelling with RELAP5^® code. The results have been compared with design and operational requirement. Possible strategies to enhance the system performances have been formulated.     

Manufacturing studies of double wall components for the ITER EC H&CD upper launcher

To counteract plasma instabilities, Electron Cyclotron Launchers will be installed in four of the ITER Upper Ports. The structural system of an EC Upper Launcher accommodates the MM-wave-components and has to meet strong demands on alignment, removal of nuclear heat loads, mechanical strength and nuclear shielding. The EC Upper Launcher has successfully undergone the Preliminary Design Review in 2009 and is now in the final design phase. Nuclear heat loads from 0.1 W/cm³ up to 0.8 W/cm³ will affect the front area of the launcher main frame. To guarantee save and homogenous removal of those heat loads, the front part of the launcher main frame is designed as a double wall steel-casing with cooling channels inside the shell structure. To finalize the design of this double wall component, the main emphasis is now to define the cooling channels geometry and to identify the optimum manufacturing route to assure adequate flow of coolant and sufficient mechanical strength in compliance with required dimension tolerances and quality of the welds. Several manufacturing options have been investigated and were evaluated by computational analysis and fabrication of pre-prototypes. To come to a final design, the most promising route will be chosen to manufacture a full-size mock-up of the double wall main frame. It will be tested at the KIT Launcher Handling Test facility to check the compliance with the design goals related to geometrical accuracy and thermo-hydraulic characteristics. This paper describes the design and the manufacturing routes of the prototypic double wall main frame.     

Implementation and application of PyNE sub-voxel R2S for shutdown dose rate analysis

PyNE R2S is a mesh-based R2S implementation with the capability of performing shutdown dose rate (SDR) analysis directly on CAD geometry with Cartesian or tetrahedral meshes. It supports advanced variance reduction for fusion energy systems. However, the assumption of homogenized materials of PyNE R2S with a Cartesian mesh throughout a mesh voxel introduces an approximation in the case where a voxel covers multiple non-void cells. This work implements a sub-voxel method to add fidelity to PyNE R2S with a Cartesian mesh during the process of activation and photon source sampling by performing independent inventory calculations for each cell within a mesh voxel and using the results of those independent calculations to sample the photon source more precisely. PyNE sub-voxel R2S has been verified with the Frascati Neutron Generator (FNG)-ITER and ITER computational shutdown dose rate benchmark problems. The results for sub-voxel R2S show satisfactory agreement with the experimental values or reference results. PyNE sub-voxel R2S has been applied to the shutdown dose rate calculation of the Chinese Fusion Engineering Testing Reactor (CFETR). In conclusion, sub-voxel R2S is a reliable tool for SDR calculation and obtains more accurate results with the same voxel size than voxel R2S.