Blanket Shield Module Research Articles

Research and development of copper joint for ITER In-Vessel Coils

The ITER In-Vessel Coils (IVCs) are crucial components of the Tokamak fusion reactor. They are located within the vacuum vessel (VV) behind the blanket shield modules and are responsible for compensating for fast perturbations of the plasma. There are a total of 176 joints used to connect the IVCs, feeders and feedthrough. The joint structure consists of a central copper conductor, a mineral (MgO) insulation layer, and an outer stainless-steel jacket. Given the inaccessibility of the IVCs once the blanket modules are installed, the highest level of quality is required for the joints. One key item is the welding of the copper conductor which has to be done inside the VV. To address the high thermal conductivity of copper conductor and the challenging constraints working inside the VV, a preheating method by the arc of the welding head has been developed. Macro- and micro-structure examination were performed on the welding cross sections, and tensile tests were carried out to check the strength of the copper joint, which demonstrated that arc preheating does not have a significant impact on the heat-affected zone. After more than 40 welding attempts and several design updates, the copper joint has been successfully developed and can be repeatedly and successfully welded.

Compaction technology development for ITER In-vessel coil joints

Two types of ITER In-vessel coils (IVCs) are installed inside the vacuum vessel inner wall, behind blanket shield modules (BSMs), and near the plasma, namely: edge localized mode (ELM) and vertical stabilization (VS) coils. Overall, 27 ELM coils are distributed around the vacuum vessel inner wall, while one VS coil each is mounted above and below the plasma. Moreover, 176 IVC joints are used for joining the IVC coils, feeders, and feedthroughs, while 144 and 32 joints are used for joining the 27 ELM and two VS coils, respectively.The joints, which constitute a stainless steel (SS) sleeve, MgO insulation, and copper tube, show comparable performance to that of the conductor and can thus withstand severe conditions. Furthermore, the minimum available space for joint manufacturing is limited to 60 mm, which is a major obstacle. In joint manufacturing, two adjacent copper conductors are first welded via tungsten inert gas (TIG) welding, then the MgO insulation and sleeve will be applied. To maintain the insulation performance, the MgO insulation and SS sleeve should be compacted. Since compaction is one of the most important processes for joint manufacturing, this paper elucidates the development of compaction technology for IVC joints.

Fatigue behavior testing and evaluation of ITER In-vessel mock-up coil under combined thermal and electromagnetic loads

ITER In-vessel coils (IVCs) are the components to be installed in the vacuum vessel for plasma positions compensation controlling. Every coil will be wound continuously and clamped through the welded Inconel 625 brackets. The coils are planned to be bolted to the rails which are welded to the vacuum vessel wall behind the blanket shield modules. The operation condition is severe for the IVC coils, combined with high radiation, high temperature, as well as the high thermal and electromagnetic loads. To make sure the safe operation, tests have been done on the mock-up coil to check its fatigue properties under the combined thermal and mechanical loads before the finalization of the design. The integrity of the mock-up coil has been evaluated after the fatigue cycling tests. In this paper, the details are presented on the test procedures and the measurement results are summarized.

Estimation of radiation conditions in the ITER electron cyclotron upper launcher with state-of-the-art simulation techniques

• Nuclear analyses in support of ECUL port plug design: computation of nuclear heating to the BSM during plasma operation, and mitigation of dose rate during maintenance. • Significant mitigation of dose rate achieved by shielding options located within the port plug itself; further reductions only possible by optimization of the surrounding environment. • Pioneering application of the MCNP conformal unstructured mesh capability for both geometry description and result tallying for computation of the nuclear heating in the BSM. • The UM succeeds in accelerating geometry preparation and result post-processing, and in removing unphysical results inherent to MCNP’s conventional approach. The ITER Electron Cyclotron Upper Launcher (ECUL) port plug is undergoing final design evolutions towards manufacturing, which include engineering efforts to ascertain and manage the radiation conditions during operation and maintenance. Two important aspects of these efforts are discussed here. First, design and analysis activities undertaken to ameliorate the residual radiation field (shut-down dose rate, SDDR), present during maintenance due to activation of components by plasma neutrons, are presented. Several shielding options were proposed and analysed. It was concluded that very significant improvement is achieved by those located within the port plug itself, and that further necessary reductions are only possible by considering changes in the surrounding environment. Second, the nuclear heating profile in the critical blanket shield module (BSM) component was computed. For this part of the work, a pioneering application of the novel MCNP conformal unstructured mesh capability (UM) for both geometry description and results tallying was made, in addition to the conventional approach of MCNP native geometry plus superimposed structured mesh. Different aspects of the results and functionality of both approaches were compared in a variety of conditions. It is concluded that, whilst the UM succeeds in accelerating geometry preparation and post-processing and in removing unphysical results inherent to the conventional approach, the latter still provides a more computationally efficient representation leading to uniform precision. These may be limiting factors for large analysis models such as those needed for ITER systems.

Designing for Remote Handling: the case-study of the ITER Plasma Position Reflectometry in-vessel antennas

The Plasma Position Reflectometry (PPR) system for the International Thermonuclear Experimental Reactor (ITER) consists of four reflectometers designed to measure the edge density profile at four locations, two of them with antennas installed inside the vacuum vessel in small apertures between blankets. To prevent damage, the antennas need to be removed/(re)installed by remote handling (RH) means before/after the removal/reinstallation of the blanket shield modules just above and below them. The RH operations shall be performed by the Blanket Remote Handling System (BRHS), [1], and require a tool not yet available in the ITER Remote Maintenance System. Herein, we describe a conceptual design of this tool that considers guidance and security to overcome some of the identified RH issues. The RH tool is designed to remove and (re)install the antenna in a single step using the translation and rotation capabilities of the BRHS (or a dual-arm manipulator attached to the BRHS) to bolt/unbolt the antenna to/from the support without extra operations. The RH tool is also designed to minimize the space required to operate between the blanket shield modules. However, the guidance pins of the tool are dimensioned according to the maximum positioning error allowed for the BRHS with a diagonal shape to guide the tool along the antenna. The design is implemented using pop-up fasteners and standard sizes authorised by the ITER Organization. A set of mechanical benchmarks are presented based on the CAD model of the tool.

Nearing final design of the ITER EC H&CD Upper Launcher

The ITER ECRH system consists of 24 gyrotrons with up to 24 MW installed millimeter wave heating power at 170 GHz, power supplies, control system, transmission lines, one Equatorial and the four Upper Launchers. With its high frequency and small beam focus the ECRH has the unique capability of driving locally current. While the Equatorial Launcher mainly acts for central heating and current profile shaping, the Upper Launchers aim on suppressing MHD instabilities, especially neoclassical tearing modes (NTM) triggering plasma disruptions. The Upper Launchers inject millimeter waves through a quasi-optical section consisting of three fixed and the front steering mirror set. The eight overlapping beams have focal points optimized for suppression of the q = 3/2 and q = 2/1 NTMs.Several project change requests required the redesign of the Upper Launchers and the connected ex-vessel system. This redesign includes a new boundary geometry of the launchers as well as a newly designed cooling system for the Blanket Shield Module (BSM), a modified flange of the BSM to the structural main frame and a refined optical design. Additionally shield blocks with integrated in-vessel waveguides were added and the closure plate with waveguide and supply line feedthroughs was adapted. Further changes, not all caused by PCRs, include newly designed ex-vessel waveguide components with a reduced aperture and redesigned ultra low-loss CVD diamond windows. Finally several components originally foreseen as off-the-shelf components have become part of the design scope. The new launcher design status is presented with selected results on numerical design validation.

Neutronics Analysis of the In-Vessel Components of the ITER Plasma-Position Reflectometry System on the High-Field Side

The ITER plasma position reflectometry system will be used to estimate the distance between the position of the magnetic separatrix and the first wall at four predefined locations, complementing the magnetic diagnostics system. The antennas of the system studied in this paper are to be installed between the blanket shield modules of rows 3 and 4, on the high-field side of the tokamak. As the antennas and part of the corresponding waveguides will be directly exposed to the plasma, they will be subjected to high radiation doses from neutrons and gammas, which may cause irradiation-induced changes in the material properties and compromise the integrity of the components. In this paper, the Monte Carlo simulation code MCNP6 and the most up-to-date ITER reference neutronics model were used to estimate the thermal loads in the system. The results, complemented with a finite-element analysis performed with ANSYS Mechanical, show that the nuclear heat loads, combined with the thermal radiation coming from the plasma, translate into operation temperatures that reach 480 °C at the tips of the antennas, slightly above the maximum recommended temperature of operation of stainless steel for ITER in-vessel components under irradiation (450 °C). Further, thermal analyses and design iterations are therefore required to improve the thermal and mechanical behavior of the system.

Status of the final design of the EC UPP launcher

Four EC H&CD (Electron Cyclotron Heating and Current Drive) launchers will be installed into the upper ports #12, #13, #15 and #16 in ITER. Beside plasma heating their main purpose is to counteract plasma instabilities by injecting up to 20 MW of microwave power at a frequency of 170 GHz at dedicated positions into the plasma. The microwave power is generated in 24 Gyrotrons outside the Tokamak and transmitted into the dedicated ports by 8 waveguide lines per launcher. After passing a CVD (Chemical Vapor Deposition) diamond window and another section of circular waveguides, the microwaves are quasi-optically guided into the plasma by an adjusted set of mirrors. These mirrors and also the foremost segments of the waveguides are mounted into the Upper Port Plug, which is a massive steel structure capable to dissipate up to 800 kW heat and to withstand heavy mechanical loads induced mainly from plasma disruptions. This paper presents the most recent status of the design of the EC H&CD Upper Port Plug (UPP), featuring the construction of the supporting structure and the BSM (Blanket Shield Module) with its plasma facing First Wall, the PHTS (Primary Heat Transfer System) water cooling layout, the shielding components engineering, the microwave system integration and comprehensive manufacturing strategies. Also correlated computational analyses to prove the final design of the launcheŕs structural system are given.

Investigation of material homogenisation on the nuclear heating estimates of the electron cyclotron-heating upper launcher blanket shield module

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.

Electro-Magnetic Analysis of the ITER Upper Visible Infrared Wide Angle Viewing System

The Upper Visible Infrared Wide Angle Viewing System (UWAVS) is a diagnostic used in five upper ports of ITER. Each UWAVS provides visible and infrared views of various sections of the divertor. A single UWAVS is designed in three main sections: in-vessel, interspace and port cell assemblies. Each assembly utilizes multiple steering and relay mirrors to direct the in-vessel light out of the tokamak to the port cell camera sensors.For the in-vessel components, the transient electro-magnetic (EM) environment resulting from the ITER magnet operation and plasma events induces design driving Lorentz forces. As such, all in-vessel systems require detailed electro-magnetic finite element analysis (FEA) to derive the resulting time dependent Lorentz loads.ANSYS Maxwell software was used to perform transient electro-magnetic simulations of the UWAVS in ITER upper port 14. A 20 degree sector, cyclic symmetric model was employed and included, inner and outer vacuum vessel, blanket shield modules, diagnostic fist wall (DFW) and shield module (DSM), upper port plug structure, DSM shield blocks, and a detailed model of the UWAVS in-vessel assembly.The resulting data includes eddy current density and vector plots along with force and moment summation for various UWAVS components. Front end optical components are specifically reported as these components have significant EM loads.

Modeling of ITER TF cooling system through 2D thermal analyses and enthalpy balance

The winding pack of the ITER Toroidal Field (TF) coils is composed of 134 turns of Nb3Sn Cable in Conduit Conductor (CICCs) wound in 7 double pancakes and cooled by supercritical helium (He) at cryogenic temperature. The cooling of the Stainless Steel (SS) case supporting the winding pack is guaranteed by He circulation in 74 parallel channels.A 2D approach to compute the temperature distribution in the ITER TF winding pack is here proposed. The TF is divided in 32 poloidal segments, for each segment the corresponding 2D model is built and a thermal analysis is performed applying the corresponding nuclear heating computed with MCNP code considering the latest design updates, such as thickness increase of the blanket shield module. The Heat Transfer coefficient (HTC) of the He flowing in the CICC and in the cooling channels of the SS case is computed with Dittus Boelter correlation at the nominal inlet pressure of 6bar. The He is assumed to enter the coil at 4.5K in the lower terminal junction, and then the bulk temperature in all the CICCs in each of the 32 segments is calculated by means of enthalpy balance between segments, considering the actual direction of He circulation, i.e. clockwise or counter-clockwise in neighboring pancakes. The He properties needed to compute the HTC, such as viscosity, specific heat and thermal conductivity, are also varied using the same strategy. With these assumptions, He temperatures close to 5.7K are computed, due to the high values of nuclear heating (which is estimated as high as 21.58kW in the 18TF). In the paper, the methodology is presented and the results are discussed in detail. Further parametric analyses are also presented to show the impact of the inlet temperature and of the nuclear heating on the temperature distribution.

Progress on the design development and prototype manufacturing of the ITER In-vessel coils

Abstracrt ITER is incorporating two types of In-Vessel Coils (IVCs): ELM Coils to mitigate Edge Localized Modes and VS Coils to provide a reliable Vertical Stabilization of the plasma. Strong coupling with the plasma is required in order that the ELM and VS Coils can meet their performance requirements. Accordingly, the IVCs are mounted on the Vacuum Vessel (VV) inner wall, in close proximity to the plasma, just behind the Blanket Shield Modules (BSM). Due to high radiation environment, mineral insulated copper conductors enclosed in a stainless steel jacket have been selected. The reference design and prototype work provided a good basis for the development of radiation resistant conductor capable of operating within the harsh conditions in ITER vacuum chamber. However, this effort identified shortcomings in achieving satisfactory manufacturing solution, and most significantly, difficulties in brazing the brackets onto the ELM coil conductor. Since this process has not proven successful, alternative designs are under development and prototyping. Prototype manufacturing on the alternative designs has been completed at ICAS, Italy and ASIPP, China. The aim was to eliminate the need for internal coil joints, to prove the principle of longer conductor length manufacturing, and to perform bending and welding trials on two different conductor cross-sections: circular and square. The procurement of the IVCs and their conductors will be done via direct call for tender from the ITER Organization. This paper will give an overview of the alternative design and prototype manufacturing of the ITER In-Vessel coils.

Neutronics analysis of the in-vessel components of the ITER plasma-position reflectometry system on the low-field side

The ITER Plasma Position Reflectometry (PPR) system will be used to estimate the distance between the position of the magnetic separatrix and the first-wall at four pre-defined locations. For gap 4, the antennas are to be installed in-vessel between two blanket shield modules. The antennas and adjacent waveguides are directly exposed to the plasma through cut-outs in the blanket modules and are therefore subjected to high radiation doses from neutrons and photons, which may cause irradiation-induced changes in the material properties and compromise the integrity of these components. Here, we report on the neutronics analysis of the PPR system of gap 4 performed with the Monte Carlo simulation code MCNP6, with the objective of estimating the thermal loads in the system. The nuclear loads obtained in the neutronics analysis will be used as input in the thermal analysis of the in-vessel components of the ITER PPR system.

Shutdown dose rate analysis of European test blanket modules shields in ITER Equatorial Port #16

‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). An essential element of the Conceptual Design Review (CDR) of these TBSs is the demonstration of capability of Test Blanket Modules (TBM) and their shields to fulfil their function and comply with the design requirements. One of the TBM shields highly relevant design aspects is the project target for shutdown dose rates (SDDR) in the interspace. We investigated two functions of the TBMs and TBM shields—the neutron flux attenuation along the shields, and the reduction of the activation of the components contributing to SDDR. It is shown that TBMs and TBM shields reduce significantly the neutron flux in the port plug (PP). In terms of neutron flux attenuation, the TBM shield provides sufficient neutron flux reduction, being responsible for 5×106n/cm2s at port interspace, while the EPP gaps and BSM gaps are responsible for 5×107n/cm2s each. When considering closed upper, lower and lateral neighbour equatorial ports (thus, excluding the cross-talk between ports), a SDDR of 121μSv/h averaged near the port closure flange was obtained, out of which, only 4μSv/h are due to the activation of TBMs and TBM shields. Maximum SDDR in the range of 300–350μSv/h were observed in the interspace. Thus, although the SDDR are found above the project target the good performance of the TBM shields design was demonstrated. Two main reasons for high SDDR were identified in this work—the neutron streaming through the equatorial PP gaps and the neutron penetration through the blanket shield modules and vacuum vessel.

Design modification of ITER equatorial EC launcher for electron cyclotron heating and current drive optimization

The design modification of the ITER equatorial electron cyclotron (EC) launcher to enhance off-axis current drive performance, changing the steering capability of injection beams from toroidal to poloidal, is described. The modification requirement impacts the design of all 3millimeter wave beam paths transmitted through the port plug and blanket shield modules (BSMs) installed in front. The optimized configuration of the mm wave design consisting of both steering and fixed mirrors, beam ducts between the waveguide outlet and BSM openings are successfully developed. Assuming a 100% pure HE11 mode propagated at the waveguide output, a 99% transmission efficiency from the waveguide outlet to the BSM output is obtained. This modification ensures that mm wave beams from both middle and bottom beam row pass through the same BSM opening and will lead to the increase of neutron shielding potential. The configurations of BSMs are also modified maintaining the structural design concept of each module. The preliminary result of electromagnetic analysis shows that induced force and momentum torque on the modified BSMs are below the mechanical criteria of the BSMs’ support structure.

Development of ITER Equatorial EC Launcher Components Toward the Final Design

Development of an electron cyclotron equatorial launcher has proceeded with a series of prototype tests and design enhancements intended to improve reliability and functionality of the launcher. The design enhancements include adaptation of the launcher steering angles such that one of three beam rows of the launcher is necessarily flipped to perform counter current drive to conform to a new ITER physics requirement. Also, the top and bottom steering rows have been tilted at an angle of 5 deg so that the top and bottom beam rows can be accessed from on-axis to near midradius. Furthermore, the position of the fixed focusing mirror that forms a quasi-optical in-vessel millimeter-wave (mm-wave) transmission line is modified to increase the nuclear shielding capability. A high-power experiment of the mm-wave launching system mock-up fabricated based on the design confirmed a successful steering capability of 20 to 40 deg. It was measured that some stray radio-frequency power propagated in the beam duct and generated some heat on the duct. Prototype tests also include the fabrication of the blanket shield module and partial port plug mock-up and have shown no serious technological issue regarding the fabrication and cooling functionality.

F4E studies for the electromagnetic analysis of ITER components

Fusion for Energy (F4E) is involved in a relevant number of activities in the area of electromagnetic analysis in support of ITER general design and EU in-kind procurement. In particular several ITER components (vacuum vessel, blanket shield modules and first wall panels, test blanket modules, ICRH antenna) are being analyzed from the electromagnetic point of view. In this paper we give an updated description of our main activities, highlighting the main assumptions, objectives, results and conclusions. The plasma instabilities we consider, typically disruptions and VDEs, can be both toroidally symmetric and asymmetric. This implies that, depending on the specific component and loading conditions, FE models we use span from a sector of 10 up to 360° of the ITER machine. The techniques for simulating the electromagnetic phenomena involved in a disruption and the postprocessing of the results to obtain the loads acting on the structures are described. Finally we summarize the typical loads applied to different components and give a critical view of the results.

Mechanical Performance of the ITER ELM Coils' Bracket Material of Inconel 625

The effects of edge localized modes (ELMs) in ITER Tokamak device are being designed to be addressed by the ELM coils. The designed ELM coils include nine toroidal sectors of three (upper, midplane, and lower) six-turn stainless steel jacketed mineral insulated coils. They will be mounted to the vacuum vessel wall behind the blanket shield modules, and fixed by a series of brackets. Mechanical performance of the brackets should achieve the design requirements to resist with the special challenges, such as the nuclear radiation (~ 3000 MGy), high temperature (~ 200°C), as well as bearing the additional loads from ELM coils. Inconel 625 is designed to manufacture the ITER ELM coils' brackets. This paper focuses on the mechanical performance of Inconel 625 materials after forging and heat treatment, especially the fatigue performances. The tensile properties and the fatigue crack growth rate of Inconel 625 after forging and heat treatment have been tested at room temperature. The test results show that the mechanical properties of the Inconel 625 material fit the design requirement.

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 3MW/m3 at the First Wall Panel’ surface, decaying down to a tenth in a distance of 0.5m behind of it will affect the BSM and the FWP. Additional heating of maximum 0.5MW/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.

Multiaxial fatigue analysis for IMIC of ITER upper ELM coil

Inconel Jacketed Mineral Insulated Conductor (IMIC) is a very important component of International Thermonuclear Experimental Reactor (ITER) Edge Localized Modes (ELM) coils, which are located between the vacuum vessel (VV) and blanket shield modules and subject to high radiation levels, high temperature and high magnetic field. These coils will experience thermal pulsed, cyclic electromagnetic (EM) load during operation. They are designed to sustain at 1.5e8 total stress cycles and shall have sufficient strength and excellent fatigue to transport and bear the high cyclic load. For IMIC, multiaxial fatigue analysis is used to evaluate failure. Two methods based on the alternating stress and mean stress in American Society of Mechanical Engineers (ASME) code provide the design codes for multiaxial fatigue evaluation: constant principal stress direction and variation of principal stress direction. Results show that using the two methods obtains basically the same equivalent alternating stress. Both of them can be recommended for the ELM coils and IMIC can meet the fatigue criteria.