In-vessel Environment Research Articles

Design of a multichannel vacuum ultraviolet spectroscopy system for SPARC.

The design of a vacuum ultraviolet spectroscopy system has been performed to monitor and provide feedback for impurity control in SPARC. The spectrometer, covering a wavelength range of 10-2000Å through a flat-field configuration with diffraction gratings, incorporates five survey lines of sight. This allows for comprehensive impurity analysis across the core and four divertor regions (inner/outer and upper/lower). Its compact modular design facilitates vertical stacking of each spectrometer unit, significantly reducing space in the tokamak hall, where a dedicated radiation shielding bunker will be built. Safety features include a secondary helium enclosure to mitigate tritium permeation risks during deuterium-tritium (D-T) operations and shielding within the beamlines for enhanced radiation protection. The silicon carbide mirror design for divertor observation ensures its survivability in the in-vessel environment of SPARC, validated by thermal and electromagnetic analysis. Signal modeling and data acquisition testing results show that an exposure time of a few milliseconds is appropriate considering photon flux reaching the detector, demonstrating the system's capability for discharge control that includes disruption avoidance.

Design updates of ITER Blanket Remote Handling System to accommodate in-vessel environment

This paper presents the recent advancement of the ITER Blanket Remote handling system (BRHS), aimed at addressing the challenging in-vessel environment of ITER. Due to the combination of gamma radiation and recently identified in-vessel humidity, original design of the BRHS with plated carbon steels must be replaced with stainless steels design. The BRHS design was modified based on evaluation of the corrosion resistance of various metals in a humid environment under gamma ray irradiation, considering strength and manufacturability of stainless steels. Furthermore, authors have also found that maraging steel treated by a black chrome plating can withstand a cumulative dose of 1 MGy without corroding, enabling the use of maraging steel wrenches, which are essential for FW central bolt torquing. The feasibility of the remote maintenance equipment design, which partially utilizes ultra-high-strength steel while employing stainless steels for the entire system, was confirmed.

Final Design of ITER In-Vessel Coils and Manufacturing of In-Vessel Coil Conductor

Following a decision made at the ITER Council in November 2013, two types of in-vessel coils (IVCs), namely, edge-localized mode (ELM) coils to mitigate ELMs and vertical stability (VS) coils to provide vertical stabilization of the plasma, have been incorporated in the ITER baseline design. The in-vessel environment is severe, characterized by large transient electromagnetic fields, high radiation flux, and high temperature. To withstand this environment and provide the required functionality, a “mineral insulated conductor” (MIC) technology has been selected for the IVC conductor. It consists of an axially water-cooled copper conductor surrounded by magnesium oxide insulation and a stainless steel jacket. A major advantage of the coil design is the choice of the same conductor material and dimensions for both VS and ELM coils, long conductor length which eliminates the need for any internal joints, and Cu welding for the joints between coils and feeders. In situ winding of the VS coils is asking for the development of a creative solution for the unspooling, straightening, precise winding tools, bending, forming, and metrology processes in a tight and congested environment. Mock-ups for design verification and manufacturing feasibility have been produced, in particular, for the bracket manufacture and to perform thermal and mechanical fatigue testing on an ELM control coil winding pack assembly. Installation trials of a half-size ELM control coil have been performed to verify the feasibility of the assembly procedure. This article will give an overview of the IVC design status and the progress of ITER IVC conductor manufacturing.

Large-scale collecting mirrors for ITER optical diagnostic

The challenges of a large-scale mirror design for the in-vessel collection system of ITER divertor Thomson scattering (DTS) are under consideration. These are the mirrors located out-of-line vision of fusion plasma (so called ‘second mirrors’) with a high-reflective Ag-based coating. The paper outlines: approaches providing optical surface shape and angular position stability; Ag coating applicability for ITER in-vessel environment, including resistance to accidental steam ingress. The proposed solutions may also appear to be of some use in other ITER diagnostics and in diagnostics of other fusion devices.

A Geometric Approach to Inverse Kinematics of Hyper-Redundant Manipulators for tokamaks maintenance

This work proposes a geometric approach to inverse kinematics of hyper-redundant manipulators used for remote maintenance of nuclear fusion reactors. The approach is particularly suited to be adopted in real-time human-in-the-loop control strategies involving high-frequency control feedback and requiring safe interaction between the manipulator and the in-vessel environment. The capability of the inverse kinematic method to find a solution for a set of different robot end-effector poses, inside a toroidal environment, was tested on the HyRMan kinematics, i.e. the Hyper Redundant Manipulator developed in the framework of the Divertor Tokamak Test (DTT) project. The simulation tests were aimed at assessing performance of the proposed method in terms of accuracy in the end-effector positioning, computational burden, distance from obstacle, distance from joint angles and torque limits and success rate of the task execution. The achieved results were compared to the ones obtained through an iterative method proposed in literature, i.e. the one based on the computation of the Jacobian pseudo-inverse, demonstrating overall higher performance of the proposed approach and comparable ability to safely avoid obstacles and joint limits.

Design and experimental tests of an active cooling system for a kind of in-vessel inspection manipulator

PurposeThis paper aims to present the design and experimental tests of an active circulating cooling system for the Experimental Advanced Superconducting Tokamak in-vessel inspection manipulator, which will help the current manipulator prototype to achieve a full-scale in-vessel high temperature environment compatibility.Design/methodology/approachThe high-temperature effects and heat transfer conditions of the manipulator under in-vessel environment were analyzed. An active circulating cooling system was designed and implemented on the manipulator prototype. A simulative in-vessel inspection task in a high temperature environment of 100°C was carried out to evaluate the performance of the active circulating cooling system.FindingsThe proposed active circulating cooling system was proved effective in helping the manipulator prototype to achieve its basic in-vessel inspection capability in a high temperature environment. The active circulating cooling system performance can be further improved considering the cooling structure coefficient differences in different manipulator parts.Originality/valueFor the first time, the active circulating cooling system was implemented and tested on a full-scale of the in-vessel inspection manipulator. The experimental data of the temperature distribution inside the manipulator and the operating status of the circulating system were helpful to evaluate the current active circulating cooling system design and provided effective guidance for improving the overall system performance.

A New Facility for Combined-Load Testing of Fusion Reactor In-Vessel Components

Meeting the challenge of realizing fusion power production will require considerable and increasing investment in facilities for testing and development of fusion technology. Particularly important will be testing of components destined for the harsh in-vessel environment of the reactor. To help address this need, the U.K. Government is investing in major new fusion technology facilities, which will offer integrated laboratories covering the complete development life cycle from materials to manufacturing processes and load testing of components. A major part of these facilities shall be a test device named CHIMERA (combined heating and magnetic research apparatus), offering testing under fusion-relevant loads for meter-scale in-vessel component mock-ups. Among the major challenges addressed are electromagnetic loads, high heat flux (HHF), and proving complex and high-risk manufacturing. The ability to test technology in magnetic fields will be unparalleled and could prove vital for breeding blanket designs featuring a ferromagnetic structural material or a liquid metal breeder. The CHIMERA magnet system uses a split-pair NbTi superconducting magnet, combined with a vertical-axis pulsed resistive solenoid to simulate plasma disruptions. Furthermore, in order to provide semi-integrated testing including possible synergistic effects, CHIMERA will enable tests of resilience against magnetic and thermal loads in combination. The heating systems will deliver at least 0.5 MW/m 2 at the module surface, HHF in localized areas, and power for volumetric heating of a module. This article introduces the CHIMERA device, reports the motivation and technical basis, describes the system specification, and outlines the future plan.

Technical challenges on material and design criteria development for fusion in-vessel components

Preparation of irradiation database is essential for the design activity of fusion reactor in-vessel components since it is difficult to suppress the irradiation-induced degradation of material properties within the negligible level. On the other hand, the preparation of irradiation database corresponding to the real fusion in-vessel environment in the level that the existing design codes require for new material is not realistic. This limitation is due to the limited capability of irradiation experiments, and the fusion neutron irradiation environment is not available for the next ten years. The probability-based design method could be a potential alternative approach to evaluating the possibility of components’ failure, where the material property is defined with a probability density function. The statistical analyses of data are required to provide the function, and a case study on as prepared and post-irradiation tensile properties of Japanese reduced activation ferritic/martensitic steel F82H was conducted. It was indicated that the most data show a normal distribution, but post-irradiation elongation data shows a possibility to follow a Weibull distribution.

Research on Nondestructive Examination of Bracket Welds of ITER In-Vessel Coils (IVC)

The ITER in-vessel coils are comprised of edge-localized mode (ELM) and vertical stabilization (VS) coils. There are 27 ELM coils and 2 VS coils. All the ELM and VS coils are supported by the clamped brackets bolted to the rails which are welded to the vacuum vessel wall. The in-vessel environment is severe, characterized by large transient electromagnetic fields, high radiation flux, and high temperature. Therefore, an appropriate nondestructive examination approach should be established and qualified. The difficulty in the development of the bracket weld nondestructive examination is the complex structure which brings space limit and various reflect echoes. Phased array ultrasonic test (PAUT) was studied and developed for nondestructive examination. In order to obtain the appropriate test parameters, CIVA simulation software was utilized to visualize the acoustic field distribution and coverage. The sensitivity of the PAUT method was also demonstrated by using machined defect provided by ITER Organization. Experiment results shows that all of reference defects can be detected with peak amplitude higher than 80% of FSH for both first leg and second leg.

Design and qualification tests of a robotic joint module for tokamak in-vessel manipulator use

PurposeThis paper aims to present the design and a prototype experiment of a robotic joint module for tokamak in-vessel manipulator-related research; the results will promote the adaptation of current in-vessel inspection manipulator to achieve full tokamak in-vessel environment compatibility.Design/methodology/approachA flexible metallic bellow-enclosed working chamber is used to protect the main servo drive components, the active cooling method for high temperature protection and the servo control structure simplification for high radiation endurance. A joint module prototype is manufactured and tested under a similar in-vessel environmental condition for extreme condition protection validation and basic servo control ability evaluation.FindingsThe joint module prototype successfully survived the similar in-vessel environment tests and proved good mobility via closed-loop servo control. A conceptual design of a serial linkage manipulator with joint module structure is proposed for future in-vessel inspection manipulator development.Originality/valueThe proposed joint module uses common industrial servo components to achieve its full extreme in-vessel environment compatibility. Different from traditional metallic bellow application in a vacuum environment to produce a linear movement result, the proposed joint module aims to achieve rotating movement directly from the metallic bellow structure, thereby reducing the joint structure space requirement, simplifying the vacuum environment movement transmission structure and increasing the vacuum environment compatibility degree.

A new trial of tokamak in-vessel inspection manipulator

In this paper, we discuss the design and partial implementation of an in-vessel inspection manipulator in detail, which is considered to serve for China's Experimental Advanced Superconducting Tokamak (EAST). Besides the ordinary kinematic/dynamic constraints and specifications for a multiple degrees of freedom (DOF) manipulator suitable for EAST in-vessel inspection, there is extra necessity in design for the extreme in-vessel environment, e.g., high temperature and high vacuum. Based on our recent developed active cooling system, a specific proposal is explored, which employs ordinary commercial mechanical/electrical components only, as if the manipulator works in normal temperature environment. This paper also emphasizes some challenging technical issues toward an implementation, such as an optimization of thermal gradient/cooling path in the manipulator, a trade-off between large reachable space and large rotation angle of each joint, a special designed revolute joint structure for cooling tube arrangement and so on. We use an EtherCAT based real time control platform connecting drivers and sensors, which achieves a robust closed-loop system and a clean cable aspect simultaneously. In the later part of the paper, basic mechanical tests and inspection process are described. Evaluation on recent progress and future work toward a whole-scale test is stated and expected.

Structural analysis and optimization of the break-out & feeder of the ITER lower vertical stabilization coil

Break-out & feeder is an essential component connecting the ITER lower vertical stabilization (VS) coil to the outside power source. It plays the role of interface between the in-vessel and out-vessel devices and has large influence on the coil performance. Due to the special location of the ITER lower VS coil, the break-out & feeder has to endure severe in-vessel environment such as high temperature, strong magnetic field and neutron irradiation. High temperature and restricted cooling paths can easily make the break-out under high thermal stress. While square crossing with the Tokamak toroidal magnetic field will result in large Lorentz forces in the feeders. Structural analysis of the break-out & feeder shows overlarge thermal stress concentrating in the coil spine where the conductors lead out. The primary stresses in the feeders are also extremely high. Moreover, there is difference in loading in comparison with the coil main body, which is designed to be mainly in compression and with relaxed crack growth issues, the break-out & feeder is not so compressive and will endure large tensile stress in work. Therefore, relieving the high stresses is important for its design. According to the results of further analysis, structure optimizations such as using block structures in the break-out and modifying the feeder supports are proved with good effect.

Recent In-Vessel Integration Concepts for a European DEMO Reactor

A part of the recent scoping studies for a European DEMO reactor deals with the design of the in-vessel components and their integration inside the reactor. The main in-vessel components are the Breeding Blankets (Helium Cooled Lithium Lead and Helium Cooled Pebble Bed), the helium supply units called Manifolds (MF) and the Neutron Shields. Alternative concepts for the integration of these components have been developed in parallel by different Europe an associations (FZK, CEA, and EFET). Nevertheless these concepts are all based on the vertical segmentation concept called “Multi Module Segment” (MMS). The big advantage of the MMS concept dwells in the fact that blankets and MF constitute a vertical non-permanent segment to be installed and dismantled by Remote Handling (RH) tools through the upper ports of the reactor. The dimensions, geometry and materials are strictly dependent on the harsh conditions of the in-vessel environment: high temperatures, high neutron fluxes, and high thermo-mechanical loads during normal operations and disruptive events. In addition, suitable systems of attachment able to withstand thermal expansion and the expected loads have been developed to provide reliability and easy in-vessel maintenance. The general aspects of the MMS system and the common RH procedures foreseen are presented and the different specific options for solving attachment and in-vessel assembly issues are discussed.

Magnetic measurements using array of integrated Hall sensors on the CASTOR tokamak

We have performed the first tests of "integrated" Hall sensors (IHSs) in a tokamak in-vessel environment. IHS combines the sensing element together with the complex electronic circuitry on a single small chip. The on-chip integrated circuits provide stabilization of the supply voltage, output amplification, noise suppression, and elimination of temperature dependencies. Eight IHSs of A1322LUA type produced by Allegro MicroSystems, Inc. were mounted on a stainless steel ring symmetrically encircling the CASTOR plasmas in poloidal direction 10 mm outside the limiter radius. IHSs were oriented such that they measure the horizontal and vertical magnetic fields. We found out that these sensors qualify for in-vessel use of small to middle sized fusion devices where the radiation is not an issue and the temperature below 150 degrees C can be guaranteed. The main advantages over the traditional pickup coils are the smaller size and more straightforward interpretation of output without the need of rather cumbersome integration and drift removal procedure associated with the use of inductive loops. We successfully exploited the sensors for determination of vertical plasma displacement on CASTOR. This new diagnostic helped us to shed more light into long term observed discrepancy on CASTOR between vertical plasma displacement as deduced by standard magnetic and by nonmagnetic diagnostics (Langmuir probes, bolometers).

Applicability of β-ray-induced X-ray spectrometry to in situ measurements of tritium retention in plasma-facing materials in ITER

Applicability of β-ray-induced X-ray spectrometry (BIXS) for in situ measurements of tritium retention by plasma-facing materials in ITER was examined using two γ-emitters and a divertor tile with metallic supports employed during D–T fusion experiments in JET. Measurements of the tile with a γ-emitter showed that the presence of the γ-field has no influence in the shape of the β-ray-induced X-ray spectra, although the whole intensity in the observed energy region was dependent on radioactivity of the γ-emitters. To simulate more accurately the in-vessel environment of PFMs, one of the divertor tiles with metallic supports was subjected to BIXS measurements. Although the metallic supports were gamma active, no significant effect was observed on the recorded X-ray spectrum. These results clearly indicate that BIXS can be applied to in situ measurements of tritium retention by PFMs with a simple shielding of the X-ray detector.

Remote metrology, mapping, and motion sensing of plasma facing components using FM coherent laser radar

Metrology inside a D/T burning fusion reactor must necessarily be conducted remotely since the in-vessel environment would be highly radioactive due to neutron activation of the torus walls. A technique based on frequency modulated coherent laser radar (FM CLR) for such remote metrology is described. Since the FM CLR relies on frequency shift to measure distances, the results are largely insensitive to surface reflectance characteristics. Results of measurements in TFTR and NSTX fusion devices using a prototype FM CLR unit, capable of remotely measuring distances (range) up to 22 m with better than 0.1-mm precision, are provided. These results illustrate that the FM CLR can be used for precision remote metrology as well as viewing. It is also shown that by conducting Doppler corrected range measurements using the CLR, the motion of objects can be tracked. Thus, the FM CLR has the potential to remotely measure the motion of plasma facing components (PFCs) during plasma disruptions.