ITER Organization Research Articles

Supervised Retinal Vessel Segmentation Based Average Filter and Iterative Self Organizing Data Analysis Technique

Retinal fundus is the inner surface of the eye associated with the lens. The identification of disease needs some parts of retinal fundus, such as blood vessel. Blood vessels are part of circulation system which functions to supply blood to retina area. This research proposed a method for segmentation of blood vessel in retinal image with Average Filter and Iterative Self-Organizing Data Analysis (ISODATA) Technique. The first step with the input image changed to Gamma Correction, increasing contrast with Contrast Limited Adaptive Histogram Equalization (CLAHE), the filtering process with Average Filter. The segmentation is used for ISODATA. Region of Interest was applied to take the center of a vessel object and remove the background. In the final stage, the process of noise reduction and removal of small pixel values with Median Filter and Closing Morphology. Datasets used in this research were DRIVE and STARE. The average result was obtained for STARE dataset with an accuracy of 94.41%, Sensitivity of 55.57%, Specification of 98.31%, F1 Score of 64.81% while for the DRIVE dataset with accuracy of 94.78%, Sensitivity of 43.46%, Specification of 99.81%, and F1 Score of 59.39%.

Large-Scale Convergence of Receptor Cell Arrays Onto Afferent Terminal Arbors in the Lorenzinian Electroreceptors of Polyodon.

Certain sensory receptors contain many transducers, converging onto few afferents. Convergence creates star-topology neural networks, of iterative parallel organization, that may yield special functional properties. We quantitated large-scale convergence in electroreceptors on the rostrum of preadult paddlefish, Polyodon spathula (Acipenseriforme vertebrates), and analyzed the afferent terminal branching underlying the convergence. From neurophysiological mapping, a recorded afferent innervated 23.3 ± 9.1 (range 6–45) ampullary organs, and innervated every ampullary organ within the receptive field’s sharp boundary. Ampullary organs each contained ∼665 Lorenzinian receptor cells, from imaging and modeling. We imaged three serial types of afferent branching at electroreceptors, after immunofluorescent labeling for neurite filaments, glial sheaths, or nodal ion channels, or by DiI tracing. (i) Myelinated tree: Each of 3.08 ± 0.51 (2–4) parallel afferents from a cranial nerve (ALLn) entered a receptive field from deeper tissue, then branched into a laminar tree of large myelinated dendrites, parallel to the skin, that branched radially until ∼9 extremities with heminodes, which were candidate sites of spike encoders. (ii) Inline transition: Each myelinated extremity led distally into local unmyelinated arbors originating at inline branching structures covered by terminal (satellite) glia. The unmyelinated transition zones included globular afferent modules, 4–6 microns wide, from which erupted fine fascicles of parallel submicron neurites, a possibly novel type of neuronal branching. The neurite fascicles formed loose bundles projecting ∼105 microns distally to innervate local groups of ∼3 adjacent ampullary organs. (iii) Radial arbors: Receptor cells in an electrosensory neuroepithelium covering the basal pole of each ampullary organ were innervated by bouton endings of radial neurites, unmyelinated and submicron, forming a thin curviplanar lamina distal to the lectin+ basal lamina. The profuse radial neurites diverged from thicker (∼2 micron) basolateral trunks. Overall, an average Polyodon electroreceptor formed a star topology array of ∼9 sensor groups. Total convergence ratios were 15,495 ± 6,052 parallel receptor cells per afferent per mean receptive field, assuming 100% innervation. Large-scale convergence likely increases the signal-to-noise ratio (SNR) of stimulus encoding into spiking afferent output, increasing receiver sensitivity. Unmyelinated arbors may also regenerate and repair the afferent innervation of ampullary organs.LSID: urn:lsid:zoobank.org:act:09BCF04C-3C3C-4B6C-9DC9-A2BF43087369

Summary report of the 3rd IAEA technical meeting on fusion data processing validation and analysis (FDPVA)

The Third IAEA Technical Meeting on Fusion Data Processing, Validation and Analysis took place at the IAEA Headquarters in Vienna, Austria, from 28–31 May and brought together more than 60 scientists and engineers from 19 Member States, the European Commission and ITER Organisation working on data analysis and machine learning (ML) methods for the processing of fusion data, collected either from experimental diagnostics or from plasma simulations.‘Accurate data processing leads to a better understanding of the physics related to nuclear fusion research. It is essential for the careful estimate of the error bars of the raw measurements and processed data,’ said D. Mazon (Chair, CEA, France) in his introductory talk and progress has been shown in this direction during the meeting. In particular the meeting discussed new developments in fusion R&D applications in the following areas: inversion techniques, such as tomography; magnetic topology reconstruction, such as equilibrium reconstructions; system identification; scaling laws determination and their accuracy for extrapolation from current machines to fusion reactors; model-based algorithms for control applications; identification of spurious and undesired events, such as disruptive phenomena or hot spots in infra-red images using sophisticated mathematical techniques like neural network and support vector machines.Discussions also focused on the potential use of these techniques for ITER, in particular, on their relevance to the first ITER plasma. The use of Integrated Modelling & Analysis Suite (IMAS) infrastructure—a convenient platform that allows implementation of different simulation codes in the same format—and synthetic diagnostics that could be coupled with this structure, in order to test the different mathematical approaches developed by the meeting participants, were explored in detail. Recent results and progress in the development of new tools will be discussed during the next meeting on the topic, scheduled in 2021.

Preparation of the ITER Central Solenoid Assembly

The Central Solenoid (CS) is a key element of the ITER Magnet system, including six identical coils, called modules, assembled together to form a 4 m outer diameter, 13 m high solenoid. It is a superconducting magnet, using a 45 kA Nb3Sn conductor internally cooled by circulation of supercritical helium at 4.5 K with a peak field up to 13 T. It is enclosed inside a structure providing vertical pre-compression and mechanical support. Procurement of the components and the special assembly tooling of the ITER CS is the responsibility of US ITER, the ITER Domestic Agency of the USA, while the ITER Organization (IO) will carry out the assembly of these components. US ITER has awarded several contracts since 2011 to supply seven modules, including a spare, structure components, and the special tooling required for the CS pre-assembly. All deliveries are scheduled with the objective to start the CS assembly at IO site early 2021. IO is now actively preparing this new phase. This paper describes the general CS pre-assembly activities from modules stacking to the pre-compression at 210 MN. The special assembly processes and related tooling are detailed. A focus is given on the module lifting operation, the extensions assembly and the pre-compression. For these last two processes, IO has investigated alternative assembly options that will be presented.

Post-processing for ITER Scrape-Off Layer Plasma Simulations (SOLPS-ITER) in IMAS Framework

With the ongoing effort to establish the Integrated Modelling and Analysis Suite (IMAS) and the Interface Data Structures (IDSs) as the standard relevant software infrastructure for ITER Organization and its parties for all ITER related activities, we report on steps towards the integration of the B2.5-Eirene plasma fluid solver, one of the codes in the SOLPS-ITER code package, within IMAS workflows, by means of a set of modules which allow post-processing of solution data into the IMAS model. This way, after a case is successfully completed, the solver simulation output results can be stored to the IDSs. To provide the means for visualization of the edge plasma state from the IDSs, a plugin for ParaView, an open-source scientific data analysis and visualization application, named ReadUALEdge was developed. With the data displayed within the ParaView application, all tools and utilities provided by ParaView can be used on the data for convenient and efficient analysis. In this article, we give an overview of the IMAS framework and SOLPS-ITER status and present the main aspects and features of the latest ReadUALEdge plugin for the ParaView application.

Study of decontamination and maintenance for vehicle manipulator of ITER blanket remote handling system

Abstract The ITER blanket remote handling system (BRHS) will be operated in the ITER vacuum vessel (VV) after plasma operation to exchange the blanket first walls (FWs) and shield blocks (SBs), which can weigh up to 4.5 t and be larger than 1.5 m. The most important operation for the BRHS is the two-year maintenance campaign in which all 440 FWs must be replaced. The in-vessel components will have been activated by neutrons produced by the D-T reaction, with the total dose estimated to be 5 MGy. Moreover, activated dust will have accumulated in the VV and will contaminate the BRHS. The BRHS must first be transported to the red zone of the hot cell facility where it will be decontaminated remotely. The BRHS will then be transported to the hands-on area to be decontaminated manually and undergo maintenance. The ITER Organization has defined the dose rate criteria in the hands-on area as 100 μSv/h, thus, decontamination and maintenance activities must be performed in compliance with these criteria. In this paper, the parts that will require replacement in the two-year maintenance campaign are estimated based on the results of past irradiation experiments studies; areas of the BRHS that could not be decontaminated remotely are specified; and the dose rate to workers standing near the BRHS are calculated by using Monte Carlo N–Particle Transport Code analysis to establish the BRHS maintenance plan.

Design of Test Platform of High-Payload Transporter Joint for a Tokamak

A general purpose remote-handling system with high-payload handing capability is necessary for Tokamak fusion reactors. Such a system helps maintain in-vessel components that could be damaged by high-temperature plasma. A high-payload transporter has been designed and a key rotary joint has been selected by the ITER Organization. To verify the feasibility of the high-payload transporter in engineering applications, its rotary joints that will be exposed to severe dynamic loads needs to be designed, manufactured, and tested. This system includes a test platform, a control system, and measuring instruments. This article discusses the design, analyses, and principle of the load simulation for the test platform. The mechanical jigs of the test platform are designed, and they consist of the bracket, Oldham coupling, and load shafts. The strength and deformation analyses of the loading shafts are conducted. The equivalent stress of the loading shafts is approximately 285.42 MPa, and the total deflection of joint #4 and the shafts is approximately 10.81 mm. The results show that the mechanical structure of the test platform conforms to the Chinese national standard. Hydraulic cylinders are used to generate the shear force, bending moment, and torque as well as to simulate the payload. The loading principle and method of the hydraulic system are introduced to illustrate that the hydraulic loading scheme can meet the loading requirements. The design scheme satisfies the test requirements, and is confirmed and accepted by the ITER Organization.

Fabrication and factory acceptance test of ITER UPSE installation tool

Abstract The ITER upper port stub extension (UPSE) which is welded on the upper position of the vacuum vessel (VV) is classified into 2 categories; one is called as the central (even) UPSE which is welded at upper central position of the VV sector at factory. The other is called as the lateral (odd) UPSE which is welded at upper lateral position of the VV sector. The central UPSE is welded at factory before delivery, but the lateral UPSE is welded after whole nine VVs welded in Tokamak pit. The final design including analysis considering design factor and mock-up test for the lateral UPSE assembly have been developed by the Korean domestic agency (KODA). For the final welding between UPSE and VV, alignment system to adjust the UPSE to its final position is also required to meet the functional requirements requested by IO-CT (ITER Organization Central Team). For the verification of the final design to check alignment system and traveling mechanism, a full scaled mock-up was fabricated and tested according to the technical requirement of the UPSE installation tool [ 8 , 9 ]. In this paper, the structural analysis to verify the structural integrity according to Machinery Directive and factory acceptance test (FAT), full load test to check the maximum bending deflection in the worst case, will be introduced.

Optimal Design Method to Improve the Magnetic Field Distribution of Multiple Square Coil Systems

Power electronic equipment regulated by the International Thermonuclear Experimental Reactor (ITER) organization must be tested in a relevant steady-state magnetic field immunity test, and the test area is generally required to be a cube shape. For this reason, a multiple square coil system structure is used. In this article, a Taylor expansion and an optimization program are used to determine the optimal parameters of a coil system with different groups. For a three-coil system, the effects of the conductor cross-section, conductor spacing, helical structure, and conductor connection on the magnetic field uniformity are analyzed. A clear understanding of the influence degree of the error helps design an actual coil system. ITER organization is currently conducting research on magnetic field generator for electronic devices in large scientific devices. The design method proposed in this article can reduce the design difficulty and provide theoretical support.

ITER Framework for RAMI Engineering

The systems engineering process starts with the discovery of the real issues that need to be resolved and the identification of failures that are the most probable or/and have the highest negative impact during the life cycle of a project. Systems engineering involves finding mitigations to these most critical problems. This logic is fully followed in reliability, availability, maintainability, inspectability (RAMI) engineering. Although this area is at its beginning in fusion technologies, a few years ago the ITER Organization developed an approach to assess the RAMI requirement of systems. As an example of what a RAMI analysis can bring to the maintainability and thus operational availability of a nuclear fusion facility like ITER, the availability of the cask and plug remote handling system in charge of handling of port plugs and their moving between the port cells to the hot cell facility is addressed in the case of diagnostic equatorial port plugs.

ITER Tokamak Cooling Water System Design Status

The tokamak cooling water system (TCWS) is the primary cooling system of the ITER tokamak machine, providing cooling water to the vacuum vessel and in-vacuum vessel components. In addition, it provides water and gas baking to its clients as well as drying them prior to maintenance activities. It is a Safety Important System that is subject to the French Order on Nuclear Pressure Equipment. The TCWS design has been modified significantly since the preliminary design. Such changes required the approval of the French Nuclear Safety Authority (ASN). The first main modification was to relocate the main equipment of the cooling loop for in-vessel components from level L4 to level L3 of the Tokamak Building in order to improve the overall building shielding capacities, and the second dealt with the ability of the TCWS to significantly reduce the mass and energy released in case of a loss-of-coolant accident (LOCA). This second improvement was required to allow a strong improvement of the vacuum vessel pressure suppression system (VVPSS), i.e., the system that prevents overpressure in the vacuum vessel in case of a LOCA. The green light for the two aforementioned modifications of the TCWS (together with the new VVPSS configuration) has been recently obtained from the ASN. Most recently, the three major subsystems of the TCWS needed for first plasma operations (to start the end of 2025) have passed the final design review process. The final design has been approved and procurement/manufacturing has started. In addition, the final design of all piping not functionally required for first plasma operation but nevertheless needed due to installation constraints has been reviewed and approved as well. A few kilometers of nuclear-grade piping have been delivered to the ITER Organization (IO) together with the main drain tanks of the system. All the other components are expected to be received on site starting May 2021. This paper gives a detailed description and status of the main TCWS subsystems needed for first plasma either because they are functionally required or because of installation constraints. The major modifications highlighted above as well as all the improvements accomplished in the final design are also detailed, together with a status on procurement and construction activities.

Manufacture and Test of the ITER TF Type HTS Current Lead Prototypes

High temperature superconducting current leads (HTS-CL) are designed to supply the current to the large superconducting ITER magnets for the operation with reduced heat load to the cryogenic system. The Toroidal field (TF) current leads are the largest with a current capacity of 68 kA each. The Institute of Plasma Physics of the Chinese Academy of Sciences (ASIPP) is responsible for the supply of the current leads based on a design jointly developed with the ITER organization. Before the supply of the TF HTS-CL series, a pair of prototypes was manufactured in 2014 by ASIPP and associated manufacturers according to previously qualified manufacturing procedures. Rigorous quality control measures were developed and applied in preparation for series manufacturing. To verify compliance of the prototypes with the ITER specification, thorough testing was conducted in 2015. The test items under particular scrutiny were: the pressure drop in the counter-flow heat exchanger, the loss of flow accident test after steady state operation at 68 kA current, the so-called overheating time of the HTS module following an induced quench, the electrical resistances of the soldered joints inside the lead assembly (i.e., the low temperature superconducting (LTS) to busbar and the HTS to LTS joints), and the conduction heat load per lead to the 4.5 K end. In this paper, the main manufacturing steps are discussed, and test results are presented and discussed.

Study of qualification process for sub-components of terminal service box

The terminal service box (TSB) is an important component of the ITER correction coil (CC) that has a complex structure and can be manufactured by several processes. The qualification process of the TSB can be divided into sub-component qualification and overall assembly qualification. The qualification process and testing are difficult because each process exhibits non-standard properties. The qualification process of the sub-components includes the welding process qualification (resistance spot welding and vacuum brazing), insulation process qualification, and electrical test qualification. In response to the ITER Organization requirements, a mock-up was designed to meet the process qualification requirements, defining the specific methods and requirements for the implementation of a single qualification process for welding, insulation, and electricity. Based on full-size vacuum pressure impregnation of the bottom and top correction coil (BTCC) insulation sample, the overall assembly scheme of the BTCC TSB was designed. In particular, an electrical test was conducted on the entire ground insulation of the BTCC TSB prototype, and the test item, content, and acceptance criteria were determined.

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.

Development of the first ITER gyrotron in QST**The views and opinions expressed herein do not necessarily reflect those of the ITER Organization.

This paper presents a summary of recent progress pertaining to the manufacturing and inspection of ITER gyrotrons and the operation system in QST. Major achievements are as follows. (i) The final design of the ITER gyrotron was accomplished and manufacturing of two ITER gyrotrons was completed. Then operation test in QST prior the shipment to ITER has started with ITER relevant high voltage power supply configuration. The 1st ITER gyrotron has achieved 1.05 MW operation for 300 s with 51% efficiency. Measured cooling channel waveform of 300 s pulse demonstrated thermally stable condition representing sufficient cooling performance for 1 MW CW operation; (ii) 5 kHz modulation operation was demonstrated up to 200 s at with >0.8 MW at flat top of pulses; (iii) 300 s/1 MW operation was repeated for 20 shots with successful 19 shots which demonstrating >95% of reliability requirement. These results lead to success of ITER EC H&CD system construction and commissioning toward first plasma.

Current design and R&D progress of the Chinese helium cooled ceramic breeder test blanket system

As the testing mockup of tritium breeding blanket for DEMO, the Chinese helium cooled ceramic breeder test blanket system (HCCB TBS) is under development by China and will be tested in ITER to verify key tritium breeding blanket technologies. After the approval of the conceptual design of the HCCB TBS by the ITER Organization in 2015, the design optimization and more research and development activities for the HCCB TBS have begun for the preliminary design. This paper gives an overview of the recent progress of the HCCB TBS development in China, including the design optimizations, material development, mockup fabrication and test facility construction.

Receiving inspection and tests of the ITER poloidal field #4 cryostat feed-through

The ITER Magnets are in the procurement and assembly phases. During these phases, critical components need to be tested and assembly procedures need to be developed and qualified on mockups. To this end, ITER Organization (IO) and the Commissariat à l’Energie Atomique (CEA) have built up a support structure with space, expertise and equipment: MIFI – Magnet Infrastructure Facilities for ITER. In this framework, IO and CEA perform Receiving Inspection and Tests (RIT) on components delivered to IO by the Domestic Agencies (DA). RIT is a critical stage in the life process of components. These tests are meant to ensure the good quality of components and occur between the Factory Acceptance Test (performed before delivery to IO) and the Site Acceptance Tests that are performed after installation in the tokamak pit. The paper explains the RIT sequence managed on the Poloidal Field #4 Cryostat Feedthrough (PF4 CFT), the first ITER Magnets component to be delivered by a Domestic Agency to ITER Organization, and details leak tests, metrology tests and High Voltage tests performed.

Experimental validation of enhanced heat flux first wall panel mechanical attachment system

JSC NIKIET is a main supplier of ITER in-vessel components and its responsibility includes the manufacture of the FW beam, finger bodies, and the mechanical attachment system of the Enhanced Heat Flux (EHF) First Wall (FW) Panels in the framework of Procurement Arrangement 1.6.P1А.RF.01 dated 14.02.2014. The mechanical attachment system comprises the central bolt, threaded barrel and system washers. This FW mechanical attachment system allows the mounting of the FW onto the installed Shield Block (SB) while accommodating all external forces acting during plasma disruptions and normal operation.The design of EHF FW mechanical attachment system has been developed and optimized through a collaboration between the ITER Organization Central Team (IO-CT) and Russian specialists in order to provide the possibility to use it for all the FW panels. Full scale mockups of the FW attachment system components have been manufactured in JSC NIKIET in 2015–2016 and mechanical tests have been performed to demonstrate operation of these mock-ups under a 1 MN cyclic force.This paper summarizes the design description and experimental results of the FW attachment system.

SPIDER gas injection and vacuum system: From design to commissioning

The ITER project requires at least two Heating Neutral Beam Injectors (NBIs), each accelerating up to 1 MV a 40 A beam of negative H-/D-.ions, to deliver to the plasma a total power of about 33 MW for one hour.Since these requirements have never been experimentally met, it was recognized necessary to build-up a test facility, named PRIMA including both a full-size negative ion source (SPIDER – Source for Production of Ion of Deuterium Extracted from Rf plasma) and a full prototype of the ITER injector (MITICA – Megavolt ITER Injector & Concept Advancement). This realization is made with the main contribution of the European Union, through the Joint Undertaking for ITER (F4E), the ITER Organization and Consorzio RFX (CRFX) that hosts the Test Facility in Padova, Italy. SPIDER is a Radio Frequency ion source that has the same characteristics foreseen for the ITER NBI but with beam energy limited to 100 keV. The mission of SPIDER is to increase the understanding of the source operation and to optimize the source performance in terms of extracted current density, uniformity and pulse duration.The paper describes the Gas injection and Vacuum System (GVS), from the analysis of requirements to the system commissioning. In particular, it presents the rationale behind the main design choices and specific manufacturing details of the gas injection plant feeding the RF source. Furthermore, the paper describes the interfaces with the SPIDER central interlock and safety systems. Finally, the main results of the GVS commissioning with the control and interlock systems are presented.

Supporting analysis of the ITER TBM Frame and Dummy TBM designs

The validation and testing of tritium breeding blankets concepts, which are relevant for a future commercial reactor, is one of the goals of the ITER project. To achieve these objectives, mock-ups of breeding blankets, called Test Blanket Modules (TBMs), are tested in three ITER equatorial ports. Each TBM and its associated shield form a TBM-Set that is mechanically attached to a steel Frame. A Frame and two TBM-Sets form a TBM port plug (TBM PP). In case a TBM-Set is not available, it can be replaced by a Dummy TBM, steel made only. Both the design and manufacture of TBM Frames and Dummy TBMs are fully under the responsibility of the ITER organization. This paper describes the summary of recent analysis activities to support the design of the TBM Frame and Dummy TBM, which is presently in the preliminary design phase.