Vacuum Vessel Port Research Articles

Concept of operation of the Agile Robot Transporter for ITER in-vessel maintenance

ITER machine availability is crucial for fusion reactors in order to be the main source of electricity, which is affected by many aspects. This paper focuses on maintenance aspects of the machine. As the plasma facing components inside the fusion reactors become highly activated caused by the neutrons generated by fusion reactions, maintenance of the fusion reactors needs to be done by remote means. In order to increase machine availability, it is required to have efficient maintenance scenario in terms of time and effort, which can be achieved through effective in-vessel remote handling (RH) operations, simplified maintenance of the RH system itself, well-coordinated logistics, and trained personnel, etc. All these aspects need to be investigated during the design phases, which requires a well-defined concept of operation of the RH systems. This paper describes the concept of operation of the Agile Robot Transporter (ART), an ITER in-vessel RH system, that is under development to carry out neutron and optical diagnostic calibrations, and removal/replacement of the blanket first wall samples. Based on the functional analysis of the required in-vessel tasks, system requirements are defined. Among them, key requirements to improve RH operation efficiency are identified and the ART is designed accordingly. Efficient deployment and removal operations of the ART are crucial so that it requires less vacuum vessel (VV) port opening and less cask transportation, which reduces needs for hands-on operations resulting in less human radiation exposure. Various and flexible tool exchange and storage concepts are proposed that allow parallel operations between the in-vessel and ex-vessel RH systems. Modular and re-configurable design is proposed for easy RH system maintenance, and effective rescue operations to rescue the RH system when it fails.

EU-DEMO vacuum vessel port closure plate fastening developments

The EU-DEMO project, which has been in the Concept Design phase since 2021, is already focusing on the integration of the main system components but is also analysing variants to simplify this process. One of these is the fastening and sealing of vacuum vessel closures, known as port closure plates, in order to simplify the execution of the work with the aid of remote maintenance. Recognising the key role of industry, the fusion programme roadmap foresees intensive industry involvement in specific tasks. To provide a starting point for such collaboration, the Fusion Technology Department (FTD) has developed an initial design for a mechanical fastening solution based on the design of ship bulkhead doors. This could be combined with a welded seal, also developed to an early stage in FTD, which could cope with all load cases without additional measures. This system of easy to use fastening and reliable welded seal could be a significant simplification compared to other established technical solutions to this function. This paper presents the above issues and describes the current state of development. It also outlines initial plans to involve experienced industrial partners in these tasks.

The effect of blanket system and port plugs on the harmonic components of error field and optimizing approach investigation

The fusion plasma is sensitive to the penetration of the toroidal mode numbers harmonic components of the error fields (EFs), and the toroidal mode components of the EFs are closely tied to the toroidal distribution of the magnetic field source. The primary source of EFs within the magnetic field sources is the current source, such as the magnet system, while the magnetic field produced by magnetized ferromagnetic materials may also generate EFs. The ferromagnetic field source, such the reduced-activation ferritic-martensitic steel, represent a smaller portion of the EFs source. However, the use of ferromagnetic materials is expected to increase significantly as tokamak research transitions from experimental devices to demonstration and commercial reactors, particularly for the structural materials of in-vessel components and port plugs. Predictably, the EFs induced by ferromagnetic materials will increase and may even surpass the design specifications. In this study, the EFs introduced by the water-cooling ceramic breeder blanket system and the vacuum vessel port plugs of China Fusion Engineering Test Reactor are comprehensively analyzed by an analytical approach and finite element numerical approach, which is based on the Fourier decomposition. The results reveal a clear linear relationship between the EFs formed by the saturated magnetization module and its rigid displacement deviating from periodic symmetry, as well as the same behavior in EFs and overall magnetization intensity. Compared to the assembly error of the blanket system, the EFs that may be introduced by the equatorial ports and plugs are substantial, including larger toroidal mode harmonic components that can lead to lock mode and disruption. Moreover, using the correlation between harmonic components and symmetrical periods, specific toroidal mode components can be selectively shielded.

European DEMO Fusion Reactor: Design and Integration of the Breeding Blanket Feeding Pipes

This article describes the design and configuration of the DEMO Breeding Blanket (BB) feeding pipes inside the upper port. As large BB segments require periodic replacement via the upper vertical ports, the space inside the upper port needs to be maximized. At the same time, the size of the upper port is constrained by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. The BB feeding pipes inside the vertical port need to be removed prior to BB maintenance, as they obstruct the removal kinematics. Since they are connected to the BB segments on the top and far from their vertical support on the bottom, the pipes need to be sufficiently flexible to allow for the thermal expansion of the BB segments and the pipes themselves. The optimization and verification of these BB pipes inside the upper port design are critical aspects in the development of DEMO. This article presents the chosen pipe configuration for both BB concepts considered for DEMO (helium- and water-cooled) and their structural verification for some of the most relevant thermal conditions. A 3D model of the pipes forest, both for the Helium-Cooled Pebble Bed (HCPB) and Water-Cooled Lithium Lead (WCLL) concepts, has been developed and integrated inside the DEMO Upper Port (UP), Upper Port Ring Channel, and Upper Port Annex (UPA). A preliminary structural analysis of the pipeline was carried out to check the structural integrity of the pipes, their flexibility against the thermal load, their internal pressure, and the deflection induced by the thermal expansion of the BB segments. The results showed that the secondary stress on the hot leg of the HCPB pipeline was above the limit, suggesting future improvements in its shape to increase the flexibility. Moreover, the WCLL concept did not have a critical point in terms of the secondary stress on the pipeline, since the thicknesses and the diameters of these pipes were smaller than the HCPB ones.

Progress in the development of the in-vessel transporter and the upper port cask for the remote replacement of the DEMO breeding blanket

The breeding blanket (BB) segments are by far the largest in-vessel components of DEMO. For their remote replacement through the upper vertical ports of the vacuum vessel (VV) recently a new concept has been developed, [1]. The concept minimizes the spread of contamination as all in-vessel operations are carried out from within a cask that is sealed to the VV and located within a sealed room providing a second confinement barrier inside the nuclear building. The removal of the BB segments from the VV is carried out by a BB transporter that is operated on the elevator system of the >20m higher cask. The limited available space makes the compact design solutions that have been developed critical to the overall concept. The BB transporter is designed according to nuclear design codes and for high payloads since the BB segments may weigh up to 180 tons. Due to the eccentric engagement points on the backside of the BB segments and due to seismic accelerations, that need to be considered, too, the BB transporter resists also to bending moments. It can carry out translational as well as tilting movements as required to disengage the BB segments from their supports and to remove them through the upper VV port. The main requirements regarding integration, BB manipulation and structural integrity have been verified. Next development steps need to include further design improvements, integration of in-vessel position survey, definition and control of motion actuations, supply cable routing, the development of rescue and recovery scenarios as well as the validation in relevant test facilities. This article describes the design of the BB lifting tools including several modifications following a set of analyses that were recently performed.

Early design validation on the Vacuum Vessel ports sealing interface installation and removal with Virtual Reality in ITER TBM port cells

Port Cell (PC) rooms in the ITER tokamak building host the devices required for the Port Plugs operation and provide connection to services. During Plasma Operation State, the Vacuum Vessel ports sealing interface provides vacuum tightness and first confinement barrier as essential functions. When a Port Plug needs to be replaced or maintained, this significant sealing component must be removed and reinstalled. The general strategy for this operation has been reviewed resulting in modification proposals to address the requirements for static and dynamic confinement and management of radiation considering the ALARA (As Low As Reasonably Achievable) approach. As Dose Reduction Measures (DRM), additional provisions have been proposed to ensure a robust confinement control during the machine shutdown operation. They mostly consist in an airlock that allows the extension of the first confinement barrier equipped with specific handling equipment to perform human-assisted operations on the sealing flanges. In particular, this paper presents the study carried out in TBM dedicated PC during the conceptual design phase of the maintenance cabin airlock and the handling tools by means of Virtual Reality (VR). VR techniques allow working with virtual prototypes at full scale from the early design stage to ease the design process and the concepts validation with important results concerning the assessments of accessibility issues. The study describes also the outcomes of the VR studies and explains how VR can be an effective instrument to conduct integration studies with complex kinematics and human access operations in the execution of the engineering validation process.

DEMO vacuum vessel port closure plate sealing and fixation activities

DEMO should be as close as possible to a future fusion power plant, produce electricity and have a high level of availability. In the current Concept Design phase, the design is to be further elaborated and industry is to be increasingly involved as a partner in the development. An essential component will be the vacuum vessel (VV), which has to fulfil many tasks. It serves as a first confinement barrier and houses in-vessel components (IVCs) that are necessary for operation. The VV has numerous openings - the ports. These are closed during DEMO operation by so-called port closure plates (PCPs). All handling of the PCPs and IVCs, as well as inspection throughout the operational phase, must be done through remote maintenance tools and without the presence of personnel. Within the framework of qualified partners from industry the fastening and sealing of the PCPs is studied. Many requirements are associated with the technical implementation. As the development of DEMO is still in an early phase, the present goal is not yet the development to a high degree of maturity. Instead the review of established industrial solutions regarding their suitability to DEMO as well as the identification of adapted solutions with more attractive features for the DEMO application are aimed for. In this paper, the background as well as the status of this task is explained.

Prediction and Validation for Welding Deformation of the Upper Port Stub

The vacuum vessel (VV) as an ultrahigh vacuum pressure vessel is one of the essential components of the Chinese Fusion Engineering Test Reactor (CFETR) device. VV has strict position tolerance. The VV port stub is a typical large-scaled welded structure manufactured by electron beam welding (EBW) technique. Welding has the most significant influence on deformation. Therefore, it is essential to analyze welding deformation. A systematic welding simulation method aimed at welding deformation controlling for a large EB welded structure is proposed in this article. First, the dynamic process of EBW of 50-mm-thick ultralow carbon austenitic stainless steel was simulated based on the thermal elastoplastic theory. The weld forming mechanism of high energy density welding of the thick plate was revealed. Then the inherent strain extracted from the results of the simulation of local welding was introduced to the simulation of tailor welding of the port stub. Finally, optimal welding sequence and clamping conditions were obtained, and the reliability of the simulation was verified by the profile measurement results of the inner shell of the upper port stub.

HEAVY ION BEAM PROBING CONCEPTUAL DESIGN FOR THE GLOBUS-M2 TOKAMAK

The paper discusses the application of the heavy ion beam probe (HIBP) diagnostic to the Globus-M2 spherical tokamak. Probing beam trajectory calculations were conducted to find the optimal position for HIBP primary and secondary beam-linesin the realistic machine geometry. Three configurations of the vacuum vessel ports of Globus-M2 were considered for the regime with toroidal magnetic field Btor=0.7 T and plasma current Ipl=0.5 MA. The optimal probing scheme with the widest area of the plasma cross-section covered by the detector grid was selected. For this scheme, the secondary beam-line was proposed.

Advanced ASDEX Upgrade pellet guiding system design.

Cryogenic pellet injection will be the prime candidate to fuel future fusion power plants. In order to harvest optimum fueling performance, it is essential to inject pellets from the magnetic high field side of the tokamak. The pellet launching system of the tokamak ASDEX Upgrade injects cryogenic hydrogen pellets with a speed of up to 1000 m/s from the magnetic high field side via curved guiding tubes. Pellets passing the guiding tube are sliding on a gas cushion, generated by the Leidenfrost effect. The actual track has a rectangular cross section and is composed of a series of ellipses in order to generate the required 270° looping type turn; the path length is 17 m. The last part of this track is marked by strong geometrical constraints from the vacuum vessel port. The previous design was composed of a sequence of three sections of ellipses too, tangentially constant but discontinuous with regard to the curvature. It had been in operation for almost 20 years. Its steps in the curvature are supposed to limit the system performance. A novel and advanced geometry concept, adopting a method well-known from civil engineering (e.g., for the railroad track design), has been applied to develop an improved design. It relies on clothoid shape sections keeping the track curvatures continuous and, thus, provides a smooth transition between all the elements. The new design presented improves the pellet launching system performance on ASDEX Upgrade and provides knowledge for an advanced design of pellet guiding tubes in future fusion devices.

Crack propagation analysis of ITER Vacuum Vessel port stub with Radial Basis Functions mesh morphing

The ITER Vacuum Vessel (VV) is one of the most important components of the tokamak machine. The severe operating conditions impose a design to withstand strong dynamic loads. A special focus is put on defects embedded in the component that, due to the not total accessibility of the VV to non-destructive examination (NDE), but also to identify their minimum safe dimension, must be assessed through Fracture Mechanics (FM) analyses. Nuclear codes give useful guidelines for design verification using simplified models that, for practical sake, do not give a detailed description of the physical phenomena. Nowadays, Finite Element Method (FEM) is a common tool for the assessment of structural components, and detailed methods for Crack Propagation analyses have been developed. A novel procedure for the crack shape evolution during cyclic loadings will be presented in this article, relying on Finite Elements Analysis (FEA) in conjunction with Radial Basis Function morphing technique, allowing a fast arrangement of the existing mesh to a new configuration. The results of the proposed method are compared with those available in literature references and in nuclear codes.

Phased Array Detection Parameters for Electron Beam Welds in CFETR Vacuum Vessel Port Stubs

The China Fusion Engineering Testing Reactor (CFETR) vacuum vessel (VV) port stub was formed by electron beam welding (EBW). In this article, the phased array ultrasonic testing (PAUT) method was introduced for testing the quality of the weld. First, the acoustic field simulation and analysis of the EBW of 316L austenitic stainless steel were carried out with different frequency probes, then the microstructure of electron beam welds were observed by microscopic metallographic to analyze the influence of the type and influence of probe, finally fabricating the reference test blocks according to the NB/T47013-2015 standard to conducting the PAUT experiment. The test results show that the type of probe has less influence on the signal-to-noise ratio (SNR) and localization of the detection of large-thickness austenitic stainless steel electron beam welds under low-frequency conditions. The SNR of EBW detection was gradually decreased as the frequency of the probe increases under the same type of probe. In the direction of depth of the EBW, with the increase in the depth, the weld microstructure size becomes finer, so the sector scan image of $\phi ~1$ -mm side drilled hole (SDH) near the upper weld was severely distorted, while the sector scan image of $\phi ~1$ -mm SDH near the lower EBW was relatively good.

High fidelity numerical fracture mechanics assisted by RBF mesh morphing

The study and design of cyclically loaded structures cannot neglect the evaluation of their fatigue behavior. Today numerical prediction tools allow adopting, in various industrial fields, refined and consolidated procedures for the assessment of cracked parts through analyses based on fracture mechanics. An high level of detail can be obtained through the use of well consolidated FEM methods, allowing an accurate and reliable calculation of the flaw Stress Intensity Factor (SIF) and its resulting prediction in terms of crack propagation. A challenging step for this computational workflow remains, however, the generation and update of the computational grid during crack evolution. It is in this context that radial basis functions (RBF) mesh morphing is emerging as a viable solution to replace the complex and time-consuming remeshing operation. The flaw front is updated, according to its propagation, by automatically deforming the numerical grid obtaining an evolutionary workflow suitable to be used for industriallysized numerical meshes (many millions of nodes). A review of applications, obtained by exploiting FEA (Ansys Mechanical) and mesh morphing (RBF Morph) state of-the-art tools, is presented in this work. At first the proposed workflow is applied on a circular notched bar with a defect controlled by a two-parameters evolution. The same approach is then refined and demonstrated for a Multi Degree of Freedom (MDoF) case on the same geometry and on the vacuum vessel port stub from the fusion nuclear reactor Iter.

Initial port integration concept for EC and NB systems in EU DEMO tokamak

The integration of the heating and current drive (HCD) systems in the EU DEMO tokamak must address a number of issues, namely space constraints in the tokamak building, remote handling requirements, breeding blanket penetration, neutron and photon radiation shielding, compliance of penetrations of the primary vacuum with safety and vacuum criteria, and a large number of loading conditions, in particular heat, electromagnetic (EM), and pressure loads in normal and off-normal conditions. A number of pre-conceptual design options for the vacuum vessel (VV) port and the port-plug are under assessment and need to be verified against all requirements and related criteria. The identification of the functional (or physics) requirements of the HCD systems remains an on-going process during the pre-conceptual design phase, hence some initial assumptions had to be made as a basis for development of the design of the VV ports and the HCD port plugs.The paper will provide an overview of present margins in the functional/physics requirements and the rationale behind the assumptions made in order to facilitate development of the pre-conceptual design options. Furthermore it will introduce the initial design concepts of the electron cyclotron (EC) Launchers and the neutral beam (NB) injectors integrated in equatorial ports. The NB duct design in DEMO and related issues such as transmission and re-ionization losses will be also addressed.

Phased array ultrasonic test of vertical defect on butt-joint weld of CFETR vacuum vessel port stub

The CFETR vacuum vessel (VV) main-body and ports are connected via port stub. For the benefits of deep penetration and small deformation, electron beam welding (EBW) was selected for port stub butt-joint welding. Unacceptable defects such as vertical crack, non-fusion and non-penetration might appear in this kind of large depth-to-width ratio weld.In this paper, the phased array ultrasonic testing (PAUT) method was introduced. Reference blocks with artificial defects of different lengths and heights were fabricated, and then tested by 2.5 MHz 64el dual element matrix array probe and 2.25 MHz 64el single linear array probe. Detection sensitivity of 0.2 mm in width was achieved. In the end, the above probes were successfully applied on the detection of the EBW testing blocks with natural defects. The testing results of reference and testing blocks implied the full capability of the selected probes.

Vacuum vessel Upper Port design assessment of the European DEMO

The present work focuses on the design assessment of the DEMO Upper Port. The size of the upper port is defined by the available space in between the toroidal field coils and the required space to integrate a thermal shield between the vacuum vessel (VV) port and the coils. Since the large breeding blanket (BB) segments will require periodic replacement via the upper vertical ports the space inside the upper port needs to be maximized. For this reason the optimization and verification of the upper port design is a critical aspect in the development of DEMO project. The work here presented investigates the possibility to have an upper port with single walled sidewalls to increase the space available inside the port for the integration of pipe work and to allow the handling of the BB segments. The work carried out evaluates the feasibility of the design solution from the structural and thermal point of view verifying the upper port structure against nuclear heating, in-vessel pressure, and electromagnetic loads due to a toroidal field coil fast discharge and plasma disruptions according to nuclear codes.

Calibration and applications of visible imaging cameras on the C-2U advanced beam-driven field-reversed configuration device.

Two filtered, fast-imaging instruments, with radial and axial views, respectively, were used on the C-2U device to visualize line emission from impurities and hydrogenic neutrals. Novel calibration techniques needed to be developed for these instruments because the accelerated pace of C-2U operations precluded access to the interior of the vacuum vessel, targets used in typical calibration methods were not available, and in order to account for effects which have not been sufficiently addressed in the literature. Spatial calibration involved optimizing parameters in a generic camera model: ex situ using a checkerboard target and in situ using the vacuum vessel port geometry. Radiometric calibration was performed ex situ in three stages. First, the camera relative response function was mapped using an algorithm developed for high dynamic-range imaging. Second, the non-uniformity of the optical system was measured using a large LCD monitor with a characterized angular emission pattern. Finally, the absolute photon efficiency of each interference filter was determined using a calibrated uniform radiance source while also accounting for reduction in the filter transmission for off-normal rays. Periodically during the run campaign, line emission from neutral beams fired into a gas target was used as an in situ reference to check for degradation of viewport transmission. One application using calibrated camera data was tomographic reconstruction of passive impurity emission, which provided a sanity check to the excluded-flux radius inferred from wall-mounted magnetic sensors.

ITER transfer cask: Preliminary assessment of dose rate due to dust remained in the cask

The Remote Handling tasks scheduled during the ITER maintenance shut-down require transportation of in-vessel components and remote-handling tools from the Vacuum Vessel (VV) ports to the Hot Cell Building (HCB). These components and tools will be moved using the Cask and Plug Remote Handling System (CPRHS).During plasma operations, plasma facing components will be highly activated by neutrons and/or contaminated with tritium. After plasma operations, activated dust will be removed from the VV but some amounts will remain. Therefore, the CPRHS may be contaminated by residual activated dust due to the transportation of these components between the VV and the HCB. As the CPRHS is not shielded, residual activated dust may lead to a residual dose rate around the CPRHS.To assess the risk of external exposition in case of human intervention for maintenance purpose inside or close to the CPRHS, dose rate estimations were performed around and inside the CPRHS for several initial dust configurations with the normalized value of 1 g of residual activated dust. The results of this study constitute a dosimetric data base and may support ITER Organization in the definition of a decontamination level and maintenance plan.

High resolution Thomson scattering on the COMPASS tokamak—extending edge plasma view and increasing repetition rate

The Thomson scattering (TS) diagnostic system on the COMPASS tokamak consists of separate collection optics for core and edge plasmas. Till now, the adverse orientation of the viewport available for the edge TS limited the investigated plasma scenarios to those, where H-mode plasma pedestal was possible to be observed. The vacuum vessel port has been recently modified and directed more to the plasma edge. To bridge the time for designing and manufacturing a new lens, the so far used edge TS collection lens was tilted and the fibre bundles rearranged to comply with the new geometry. However, the full potential of the new port is utilized only by the newly designed and manufactured lens. Comparison of performance of the collection lenses before and after the viewport modification is presented. Apart from the edge collection system upgrade, the laser system was extended by two lasers with parameters similar to the original two, thus increasing the repetition rate of the TS system to 120 Hz or allowing to obtain four profiles with tight timing in order to measure fast transient events in the plasma.

Fusion energy development status in Korea

This overview discusses the status of fusion energy development in the Republic of Korea. Korea studies fusion energy through research and development in 2 ways: one by performing the Korea Superconducting Tokamak Advanced Research (KSTAR) for a domestic approach and the other by participating in the international collaboration project ITER. Korea had remarkable progresses not only in KSTAR but also in the ITER project in 2016. In KSTAR, a world record of more than 70 seconds in high-performance plasma (H-mode) operation in the superconducting tokamak has been achieved. For the ITER project, toroidal field conductors have been successfully accomplished and the other major equipment such as vacuum vessel sectors, ports, and assembly tooling seem to be reaching an end. This paper deals with current research and development results on both fields of KSTAR study and ITER project progress.