Shield Module Research Articles

The ITER radial neutron camera in-port system: Nuclear analyses in support of its design development

The ITER Radial Neutron Camera (RNC) is a multichannel detection system hosted in the Equatorial Port Plug 1 (EPP 1). It is designed to measure the uncollided neutron flux from the plasma, providing information on the neutron emissivity profile and total strength. The RNC structure consists of two sub-systems based on fan-shaped arrays of cylindrical collimators: the ex-port system, covering the plasma core with 2 sets of lines of sight lying on different toroidal planes, and the in-port system, enclosed in a dedicated cassette within the EPP1 diagnostic shielding module, for the measurement of neutrons generated in the plasma edge. Due to the harsh environment in which it has to operate, the design of the in-port RNC system is particularly critical both from the measurements point of view (low signal to noise ratio induced by the high level of scattered neutrons at the detector positions) and from the structural point of view. The paper presents the results of the neutronic analyses performed with the MCNP Monte Carlo code with the aim of optimizing the in-port RNC design in order to enhance the diagnostic measurement performance and evaluating the nuclear loads that have to be withstand by its structural elements, detectors and associated components.

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.

Neutronic analysis of ITER radial x-ray camera

The radial x-ray camera (RXC) is designed to measure the poloidal profile of plasma x-ray emission with high spatial and temporal resolution. The RXC diagnostic system consists of internal camera module and external camera module that view the core region and outer region through the vertical slots of the diagnostic first wall and diagnostics shield module of the equatorial port plug. To ensure the normal performance of the silicon photodiode array detectors of the cameras in the hard neutron irradiation environment in ITER tokamak, it is necessary to calculate neutron flux, radiation damage and the nuclear heating of the silicon photodiode array detectors and simulate the radiation maps of the range of RXC. This work estimated the nuclear environment of RXC based on Monte Carlo N-particle transport code, plasma scenarios of ITER tokamak and the RXC-integrated ITER CLITE model. The neutron flux of silicon photodiode array detectors and the lifetime of the silicon photodiode detector in the camera were calculated. The neutronic analysis results show that the shielding design has achieved the effect as expected and is able to guarantee the normal work of the detector during the ITER deuterium–deuterium phase without replacement, three detectors of the external camera can be operated during the whole deuterium–tritium phase without replacement.

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.

Neutronic effects of diagnostic shield module length on radiation environment of ITER diagnostic generic upper port plug

Newly developed 3D CAD-based MCNP neutronics model of ITER Diagnostic Generic Upper Port Plug (DGUPP) is a baseline MCNP model of the ITER tokamak Upper Port Plug. This paper presents 3D neutron flux and Shut-Down Dose Rate (SDDR) maps for the DGUPP integrated into the MCNP C-lite V2 global model. The SDDR assessments have been performed with the R2Smesh approach and focused on the UPP Inter-Space Structure (ISS), where personnel access is planned. Following the ALARA principle, SDDR in the ISS should be minimized, for that reason the length of the Diagnostic Shielding Module (DSM) was extended by filling all the UP structure till the UP Closure Plate with the DSM shielding material. The corresponding MCNP model was called DGUPP “long-DSM”, in contrast to the original CAD-based “short-DSM”. The results of the DSM extension indicated that the strongest, up to 3 times reduction of fluxes & SDDR could be reached at the UP Closure Plate, while inside the maintenance corridors the SDDR reduction is 11%. Found possibility to reduce the SDDR emphasizes the importance of this work.

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.

EM Analysis of ITER Diagnostics Upper Port Plugs 14 and 11 During Plasma Disruptions

International thermonuclear experimental reactor (ITER) diagnostic port plugs perform many functions including nuclear shielding and structural support of diagnostic system, while allowing for diagnostic access to the plasma. With design advancing, the in-port diagnostic components are integrated into the port plug structure, and the diagnostic shield modules (DSMs) are customized to accommodate various in-port diagnostic components. This paper summarizes results of transient electromagnetic analysis using Opera 3-D in support of recent design activities for ITER diagnostic upper port plug 14 (UPP14) and upper port plug 11 (UPP11). A complete distribution of disruption loads on each component in the two ports is presented. Impacts of different design features to the electromagnetic loads are discussed.

Design and Analysis Progress of US ITER Diagnostic Upper Port #14

ITER is the world's largest fusion device currently under construction in the South of France with >50 diagnostic systems to be installed inside the port plugs (PPs), the interspace (IS), or the port cell (PC) region of various diagnostic ports. Four tenant diagnostic systems will be integrated into the upper port 14 (U14), the upper wide angle viewing system, the disruption mitigation system, the glow discharge cleaning (GDC) system and the plasma position reflectometry, an ex-vessel tenant installed in the U14 IS region. Multiphysics analyses were performed for an in-vessel and ex-vessel component design integration following the ITER port integration requirements and the ITER guideline for load specification. Electromagnetic (EM), inertial, and heating loads of the main U14 components under various conditions for the diagnostic shield module (DSM) and its internal shielding block attachment design have been summarized. The structural design of in-port components is based on an optimization of the EM load distribution to limit the load impact to the diagnostic first wall and the tenant diagnostics and to mitigate the effect of dynamic amplification from asymmetric plasma vertical displacement event inertial loads. The DSM needs to provide a sufficient stiffness for the protection of onboard diagnostics and structural integrity of the port integration under the design-driving EM and inertial loads. The in-port diagnostics and mounting supports, on the other hand, are designed to meet the remote handling requirement, while taking into account the thermal loads due to temperature gradient from nuclear heating, as well as matching the relative stiffness among the structural components within the port assembly. Progress on an integrated design and analysis is reported including a seismic analysis of the IS/PC region for extracting tenant interface loads. The design-driving loads for tenant systems are presented for the attachment structure design as part of the PP engineering tasks.

Prototype Design of a 700 °C In-Vacuum Blackbody Source for <italic>In Situ</italic> Calibration of the ITER ECE Diagnostic

Two blackbody sources permanently located within the international thermonuclear experimental reactor (ITER) diagnostic shield module at equatorial port 9 will operate in conjunction with two remotely retractable mirrors to generate and direct blackbody radiation to calibrate the radial and oblique views of the ITER electron cyclotron emission diagnostic. The main calibration requirements include a high-emissivity surface heated to 700 °C or higher, the ability to perform the calibration in high vacuum and in the presence of a magnetic field, and 5000-h operational lifetime over 20 years. Major constraints include a heater current limit of 40 A, no direct fluid cooling to mitigate failure risks, small size to avoid compromising neutron shielding, and adequate structural support to mitigate vibration loads. This paper describes the preliminary design of a hot calibration source supported by analysis and test results of a calibration source test prototype developed to address the calibration source requirements and its constraints.

Design and Analysis Progress of ITER Diagnostic Equatorial Port #09

ITER is the world's largest fusion device currently under construction in the South of France with >50 diagnostic systems to be installed inside the port plugs (PPs), the interspace (IS), or the port cell region of various diagnostic ports. The plasma facing diagnostic first wall (DFW) and its supporting diagnostic shield modules (DSM) are designed to protect front-end diagnostics from plasma neutron and plasma radiation, while providing apertures for diagnostic access to the plasma. The design of ITER port plug structures including the DFW and the DSM is largely driven by the electromagnetic loads included on the PP structural components during plasma major disruptions and the vertical displacement events (VDEs). Unlike DFW and DSM, the design of diagnostic system, however, is likely driven by the steady-state thermal loads from plasma volumetric and surface heating and the dynamic response of the in-port components attached to the port-specific DSM or closure plate under transient loads induced on the vacuum vessel and the port extension during asymmetric VDEs. Three tenant diagnostic systems are integrated into the equatorial port 09. The toroidal interferometer/polarimeter, or TIP system, is installed in the left drawer (DSM3, left looking from plasma) for measuring the plasma density so to control the fuel input. The electron cyclotron emission (ECE) system is installed in the middle drawer (DSM2) to provide the high spatial and temporal resolution measurements of electron temperature evolution and the electron thermal transport inferences. The visible/infrared wide angle viewing system is installed in the right drawer (DSM1) to provide visible and infrared viewing and temperature data of the first wall for its protection in support of machine operation. The port plug integration design and multiphysics analyses are performed following port integration requirements including the weight limit (45 tones total), shut down dose rate limits, the cooling/heating and structural integrity validation. Mass distribution for TIP and ECE DSMs has been optimized to minimize the total weight by a new design of the boron carbide shielding pocket. The lightened DSM maintains its front-end load distribution and the structural stiffness with minimum impact to the DFW so to better protect on-board diagnostics; while still provides sufficient front-end stiffness for its structural integrity as well as the diagnostics function requirements.

Design of portable 3-axis filament winding machine with inexpensive control system

Filament winding technology is one of the fundamental methods in composite material fields, which needs a high degree of automation. It is the process in which continuous strands or filament of fibres are wound on the mandrel, which is more suitable for making high-pressure vessels, pipes, shaft and ducts. The filament winding machines have existed in enterprises or factories, which are high costs, heavy, complex control system, simple products so far. This paper describes the development of a new 3-axis filament winding machine for the production of wound composite cylinders, which is designed to be more portable, lightweight, low costs, high efficiency and easy control system compared to previous machines. It relates to design hardware control system and the software control system. Based on three axes movement principles, a 3-axis prototype filament winding machine has been developed. The x-axis is the movement left and right of the carriage, the y-axis is the rotation of mandrel and the z-axis is the movement of the feeder. Arduino Uno and CNC (Computer Numeral Control) shield module are used as hardware of control system. Universal G-Code Sender (UGS) and Grbl (G-codes) are adopted as software control system. In conclusion, a protable 3-axis filament winding machine has been successfully designed and performed, which offers potential filament wound composite cylinders with simple control system.

The use of the long modular diagnostics shield module to mitigate shutdown dose rates in the ITER diagnostics equatorial ports

The ITER equatorial port plugs are submitted to a drained weight limit of 45 T. This limitation can conflict with their radiation shielding demands, although some weight margin is being discussed. The port interspaces are subject to a shutdown dose rate limit of 100 µSv h−1 after 106 s of cooling time. To meet it, the port plugs must show a neutron flux attenuation comparable to their neighborhood, despite considering penetrations to host systems. Most of this task relies on the drawer shield module (DSM). In this work, two DSM concepts are analyzed with this perspective: the box-based DSM and the modular DSM. Regardless the penetrations, the box-based DSM leads to unsatisfactory port plugs to meet both weight and SDDR requirements. On the contrary, the modular DSM shows a performance which allows for the adoption of such DSM concept, or equivalent, a port may comply with both requirements at the same time, provided the penetrations are well designed.

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.

Cooling circle design and testing for ITER radial X-ray camera

The ITER radial X-ray camera (RXC), which is installed in the middle diagnostics shield module (DSM) of equatorial port plug 12, is an important piece of diagnostic equipment in the tokamak system. A cooling circle is vital for RXC detectors because it can protect them from being damaged during the DSM baking phase when the temperature reaches 240 (±10)°C. Helium is used as a cooling medium in this cooling circle owing to its low activation and good heat exchange with copper. To increase the gas resistance and expand the heat transfer area, a labyrinth structure is introduced into the internal structural design of a heat exchanger. In this paper, an analysis of the heat load of the RXC and its cooling circle is presented, along with the thermal simulation and test results. Thermal simulation results indicate that the cooling circle design can meet the requirements. A test platform is built to validate the cooling circle design. The experimental results are presented, and the problems encountered during the testing are analyzed. The test results indicate that the maximum temperature of the detector is lower than 65°C with the cooling circle during the platform baking phase, which is lower than the detector operation temperature limit of 75°C.

Renewable jet fuel company Velocys shares jump 40% after British Airways deal

The ITER radial X-ray camera (RXC), which is installed in the middle diagnostics shield module (DSM) of equatorial port plug 12, is an important piece of diagnostic equipment in the tokamak system. A cooling circle is vital for RXC detectors because it can protect them from being damaged during the DSM baking phase when the temperature reaches 240 (±10)°C. Helium is used as a cooling medium in this cooling circle owing to its low activation and good heat exchange with copper. To increase the gas resistance and expand the heat transfer area, a labyrinth structure is introduced into the internal structural design of a heat exchanger. In this paper, an analysis of the heat load of the RXC and its cooling circle is presented, along with the thermal simulation and test results. Thermal simulation results indicate that the cooling circle design can meet the requirements. A test platform is built to validate the cooling circle design. The experimental results are presented, and the problems encountered during the testing are analyzed. The test results indicate that the maximum temperature of the detector is lower than 65 °C with the cooling circle during the platform baking phase, which is lower than the detector operation temperature limit of 75 °C.

Electrically Conductive Polymer Composite Dispensing Process for EMI Shielding Structure

Electro Magnetic Interference (EMI) shielding structure with low cost and high performance is proposed. Conductive polymer composite is formulated with metal based filler and silicone polymer and solvent. The liquid composite material can be easily dispensed through the low pressure injection nozzle and cured to the conformal structure with high aspect ratio. High electro conductive polymer composite dispensing process is developed to build a conformal structure for the EMI shielding of PBA(Printed Board Assembly). In order to increase electrical conductivity of composite material, Ag coated copper is adopted for conductive filler material and dendrite structure is applied to reduce content ratio of the conductive filler in the composite material. An EMI shielding structure is designed to cover the exposed surface of the chip package and modified for the shielding module containing chips and electrical elements on the PBA. The wall thickness of shielding layer is under 0.6mm and aspect ratio is lower than 2.0. The conformal structure is fabricated by dispensing process with side slot nozzle and viscoelastic behavior of composite material maintains the vertical thin wall structure. EMI shielding performance is verified by measurement of chip level EMI shielding test procedure developed in this study.

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.

EM Loads on ITER Diagnostics Equatorial Port Plug 9 and Its In-Port Components During Plasma Disruptions

International Thermonuclear Experimental Reactor (ITER) diagnostic port plugs perform many functions including nuclear shielding, structural support of diagnostic system, while allowing for diagnostic access to the plasma. With design advancing, the in-port diagnostic components are integrated into the port plug structure, and the diagnostic shield modules (DSM) are customized to accommodate the in-port diagnostic components. This technical note summarizes results of transient electro-magnetic analysis using Opera 3d in support of recent design activities for ITER diagnostic equatorial port plug (EPP). A complete distribution of disruption loads on each component in EPP9 is presented. Impacts of different design features, such as the locations of the electrical contact, to the EM loads are discussed, and solutions for improving the port structure design are proposed.

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.