Search

Abstract: In this paper, the low cycle fatigue properties of JLF-1 steel were studied from room temperature (RT) to 873K in a vacuum condition using engineering size round bar specimens. The fatigue life at elevated temperature was almost the same as that at RT when the life was plotted against the total strain range. But when the life was plotted against the plastic strain range, the temperature effect on fatigue life exceeded the 95% confidence band. Based on TEM analysis, this tendency was caused by the interaction of dislocation changing with temperature, which will affect the possibility of crack initiation. Cyclic softening was observed in the fatigue test at elevated temperature. The cyclic stress–strain curve was obtained from the fatigue stress–strain hysteresis curves at around half life. The cyclic yield point was lower than the static one, especially at 873K. Thus, cyclic softening is a design issue at elevated temperature, which reduces the design margin significantly. [Copyright &y& Elsevier]

• Cable-in-conduit conductor are less affected by external magnetic fields. • Its residual magnetic field relaxation time is even shorter than the other one. • Constructing a 3D model and solving it using the T-A formulation. • The structure's internal conductors rotated 60° in sequence is the best arrangement. With its high critical current density and excellent mechanical properties, second generation high temperature superconducting tapes (2G HTS) are attractive as high current conductors in power transmission and high magnetic field applications. Due to the high width to thickness ratio of the conductor, the screening current and screening current induced field (SCIF) caused by the alternating magnetic field during stacking cannot be ignored, and the instability of the central magnetic field due to the relaxation of the SCIF is also one of the problems studied in this paper. Therefore, the study of SCIF is important for superconducting magnets and power transmission applications. In this paper, the SCIF in quasi-isotropic stranded Cable-in-conduit conductor (QIS-CICC) is investigated by simulation. The effect of the amplitude and direction of the alternating magnetic field on the SCIF, is obtained by applying the alternating magnetic field to the different directions of the QIS-CICC. The simulation results are also compared with Twisted Stacked tape cable based CICC (TSTC-CICC), which are necessary for the design of the next generation of superconducting magnets. [ABSTRACT FROM AUTHOR]

Fusion for Energy (F4E), the European Domestic Agency for ITER, is responsible for the procurement of the divertor inner vertical target (IVT) and the cassette body (CB). Both IVT and CB have been supported by extensive Research and Development (R&D) work programmes, implemented over more than 15 years in Europe, to validate design, develop various fabrication technologies, assess thermo-mechanical and thermohydraulics performances and RH installation and maintainability. This paper presents the work performed and the technical challenges associated to the fabrication of the CB prototypes, the results of the prototype qualification as well as the lessons learned. [ABSTRACT FROM AUTHOR]

• The paper confirms the suitability of the design of the ITER Cryo-distribution (CD)components. • The paper suggests the use of the liquid helium baths present in the ITER CD system as one of the heat load mitigation schemes to the ITER cryoplant. • The paper suggests the minimum opening percentage of the liquid helium baths to minimize the helium inventory in the ITER CD system. • The paper presents the dynamics of the heat load, mass flow, and pressure on the ITER cryoplant interface with the ITER CD system. The ITER cryogenic system consists of the Liquid Helium (LHe) plant, the Cryo-Distribution (CD) system, and the cryo-lines. The Auxiliary Cold Boxes (ACBs) dedicated to cooling the superconducting (SC) magnet system and the Cryoplant Termination Cold Box (CTCB) of the ITER CD system are in the factory acceptance phase. The internal components of ACBs, e.g. , cryogenic valves, a cold compressor (CCp), heat exchangers, and a cold circulator (CCr), have been sized and assembled, ensuring their functionality. The interdependency of the functional parameters of one component over the others needs to be assessed, as their integrated performance under the dynamic heat load deposition from the SC magnets may impact the overall operation of the ITER cryogenic system. The ACBs are equipped with two helium baths having ∼ 1200 kg of He inventory and situated inside the Tokamak building. These baths act as a thermal buffer for the LHe plant, situated in the cryoplant building, allowing it to operate at a quasi-steady state despite heat load variation from the applications. Such a large helium inventory can challenge the secondary confinement system of ITER due to helium ingress accidental events and thus needs to be optimized. The integrated system-level simulation is therefore necessary for the safe and reliable operation of the cryogenic system under such demanding requirements. The present study summarizes the results obtained for ACBs dedicated to the magnet system, including CTCB for the enhanced ITER operation modes, and confirms the integrated performance of the system. The results show that the LHe baths inside the ACBs can be used as a thermal buffer with the proposed limit of initial filling and by keeping a constant opening of the respective J-T valves upstream of the LHe baths. The study outcome and the proposed recommendations would be beneficial to mitigate the pulsed heat load to the LHe plant while minimizing the helium inventory. [ABSTRACT FROM AUTHOR]

• In this paper, a novel fusion power topology is proposed. • A simple fusion power supply cost model was constructed. • Demonstrated performance and cost advantages of novel fusion power supply system. Large capacity fusion devices power supply poses a significant challenge to the stability of power grid, as it can lead to power outages and jeopardize the safety of fusion devices. And traditional distribution methods result in a significant waste of resources. This paper proposes novel topologies with integrated energy storage. In these topologies, high-amplitude pulsed power is supplied by the energy storage devices, while low-amplitude stable power is obtained from the grid. This decouples the pulsed power from power grid, and significantly reducing its impact. Moreover, it can reduce the design capacity of distribution equipment and lowers investment costs. To optimize the deployment of the energy storage device, a hybrid topology is proposed, which further reducing the cost of the novel power supply. Additionally, a cost model for the fusion power supply is developed and validated using simulation data from ITER. Through the case study shows that the HEPS topology saves more than 10 % of the investment cost and about 60 % of the annual operating cost compared to the traditional converter topology. [ABSTRACT FROM AUTHOR]

• The reliability of high-power switches in Fast Discharge Unit (FDU) is very important for the safety operation of superconducting magnets. In this paper, the reliability and test number of FDU switches are discussed. • The FDU configuration characteristics of different types of superconducting magnets are introduced, and some explanation and calculation on reliability are provided. • The reliability interval estimation after a certain test number is discussed by using the binomial distribution theory. • The reliability conformance test based on the claimed reliability is discussed to derive the test number. High-power switches, as the important active part of Fast Discharge Unit (FDU) in Tokamak, are designed to interrupt the circuit current and transfer the quench energy for protecting expensive superconducting magnets. Hence, their reliability has always been highly underlined. In this paper, the reliability and test number of FDU switches are discussed, mainly including three issues. First, the FDU configuration characteristics of different types of superconducting magnets are introduced, and explanations and calculations on reliability are provided. Second, the reliability interval estimation after a certain test number is discussed by using the binomial distribution theory. Three, the compliance test based on the claimed reliability is studied, and two types of test plans are discussed and compared to derive the test number. The discussion in this paper focuses on the FDU switches and provides valuable guidance for their reliability test and assessment. [ABSTRACT FROM AUTHOR]

• The safety and interlock protection system constitute integral components of HHFTF's overall control system, providing the capability to manage various subsystems, many of these systems are operated reliably close to their performance limits in order to achieve the desired goals. • Demonstration of an enhanced interlock protection system, encompassing both software and critical hardwired interlocks, within the high heat flux test facility. • Three different types of architecture have been developed: (1) Slow Architecture based on PLCs, for functions where response time of longer than 20 ms is adequate; (2) Fast Architecture based on FPGAs, for functions requiring fast response time beyond the capabilities of the PLC; and (3) Hardwired Architecture for critical functions. • Activation times of interlocks are demonstrated within a range from 100 microseconds to 116 milliseconds. This paper introduces the safety and protection measures implemented in the High Heat Flux Test Facility (HHFTF). The interlock system has been enhanced and integrated into HHFTF to ensure the protection of the facility and the safety of personnel, thereby mitigating risks during operations. The primary purpose of these interlocks is to secure the entire facility under well-defined conditions to prevent accidents. The safety and interlock protection system constitute integral components of HHFTF's overall control system, providing the capability to manage various subsystems, many of these systems are operated reliably close to their performance limits in order to achieve the desired goals. The overall safety, interlock protection and the control system encompasses both hardware components and software, employing Programmable Logic Controllers (PLCs) and Field Programmable Gate Arrays (FPGAs), along with wired logic based on relays and special logic cards. Three different types of architecture have been developed: (1) Slow Architecture based on PLCs, for functions where response time of longer than 20 ms is adequate; (2) Fast Architecture based on FPGAs, for functions requiring fast response time beyond the capabilities of the PLC; and (3) Hardwired Architecture for critical functions. The paper showcases the successful testing and outcomes of an enhanced interlock protection system, encompassing both software and critical hardwired interlocks, within the HHFTF. Key parameters monitored include the maximum allowable job threshold temperature, flow rates (for coolant loss detection), and chamber pressure. Activation times of interlocks were observed within a range from 100 microseconds to 116 milliseconds. [ABSTRACT FROM AUTHOR]

• We have devised a composite gear transmission system, which comprises a RV reducer, two parallel shaft transmissions, and a planetary gear transmission. • A rescue module has been integrated into the driving system to ensure the necessary rescue operations for the manipulator. • The structural configuration of the relevant shells has been developed, with comprehensive explanation on dimensions, manufacturing methodology and strength assessment. • A detailed exposition on the internal architecture and rotational capabilities of these joints has been provided. • An effective internal cable routing scheme for the root joints has been proposed. This study concentrates on the engineering design and analysis of the root joints for the CFETR Multi-Purpose Overload Robot (CMOR), a crucial component in the China Fusion Engineering Test Reactor (CFETR). The CFETR tokamak is an engineering test reactor in China for magnet confinement fusion research. CMOR is a 7-DOF manipulator characterized by heavy-load capacity, high precision, and compact structures, designed for internal maintenance inside the CFETR. This presents significant challenges in its design. The design of the root joints is significant for the CMOR. This paper provides an in-depth exploration of the engineering design and analysis of the root joints. It offers a thorough explanation of the driving systems' design, including an integrated rescue module to address potential manipulator faults. Additionally, this paper delves into the conceptual design and manufacturing schemes of the corresponding shells, which are validated under various loading conditions—including regular and earthquake scenarios—through finite element analysis. The assembled root joints' internal structure and cable routing scheme are also discussed in-depth. This research provides valuable insights for future joint designs of the CMOR heavy load manipulator, advancing maintenance operations for fusion reactors. [ABSTRACT FROM AUTHOR]

• Context within EUROfusion: the work is conducted within the EUROfusion framework, emphasizing its significance within the fusion research community. • Addressing Fundamental Issues: it aims to tackle fundamental issues crucial for achieving energy extraction from fusion reactors, indicating its importance in advancing fusion technology. • Plasma Aspect Ratio and PF system size: choosing a significantly high aspect ratio for the plasma results in a configuration where the PF coil system is too distant and subsequently too large to control the plasma in a conventional manner. • Machine Assembly Configuration: it proposes an alternative machine assembly configuration, highlighting the importance of proximity between poloidal magnets and the plasma. The PF system is now placed inside the TF system. Both systems are based on HTS magnets. • Feasibility Analysis: the paper includes a feasibility analysis of the proposed solution, providing details on a proposed overall fabrication and assembly sequence for the tokamak. The paper introduces a new approach to assemble tokamak machines applied to the plasma-based Volumetric Neutron Source (VNS) device, addressing challenges related to machine topology. The VNS, designed as a compact tokamak, serves as a qualification testbed for in-vessel fusion components, especially the breeding blanket. The project aims to bridge the technology gap between ITER and DEMO. The machine size has been minimized to maximize the Neutron Wall Load (NWL). This results in a small, dense plasma and, due to the massive neutron shielding structure, a comparably large distance between the plasma and the equilibrium magnets, challenging the primary stabilization and shaping of the plasma. To overcome this, a magnet system architecture is proposed for VNS in which the poloidal field (PF) coil system is placed inside the TF coils. An assembly procedure is proposed, where the superconducting PF coils are first completed and positioned around the vacuum vessel, followed by the in-situ winding and assembly of the toroidal field coil system. The main magnetic cage components of the machine are based on High-Temperature Superconductors (HTS) to avoid the need for heat treatment. The paper presents the advantages of the magnet system architecture, and the preliminary feasibility study of the machine assembly process. [ABSTRACT FROM AUTHOR]

• This paper focuses on the defect of high driving power caused by the low density of helium coolant in the tritium breeding blanket of CFETR. Based on extensive research, a novel approach is proposed to mix supercritical CO2 into the blanket to optimize the driving power and improve the energy conversion efficiency. The innovations are summarized as follows. • The change of coolant property as a function of temperature and the constant property parameters show that the change of property with temperature cannot be ignored.Therefore, the physical parameters of coolants with different mixing ratios are fitted, and a UDF program is developed to calculate them. • The necessary fan driving power for different gas mixtures is calculated based on the calculation results of the first wall under different coolants and the data of Helium coolant in the whole blanket, and the energy proportion of the fan driving power in the total thermal power of the reactor is determined. • Combined with the energy conversion system of the secondary loop, the thermal-electric conversion efficiency of the whole reactor is calculated. It is found that the mixture with the lowest fan driving power may not necessarily have the highest thermal-electric conversion efficiency, which provides some value for further research. The first wall of the tritium breeding blanket is an important high-heat flux component in fusion reactors, and its thermal analysis is critical. Helium is chosen as the coolant for the CFETR tritium breeding blanket due to its good compatibility with structural materials and excellent thermal properties. However, its low density requires a significant amount of drive power. After extensive feasibility research, using He-CO 2 mixture as coolant is one of the main optimization directions. In this paper, a new method is proposed to consider the variation of thermal properties parameters of the He-CO 2 mixture with temperature. A fitting formula for the He-CO 2 mixture at different ratios under 12MPa is provided, and the mixture is simulated in the commercial CFD software Fluent by compiling a UDF program. The results show that the optimum He mole fraction in the He-CO 2 mixture is 60%, which reduces the required drive power by 89.36% compared with using pure He as coolant and by 81.57% compared with using CO 2.With further programming calculations and considering the effect of the secondary circuit, it is calculated that the highest energy conversion efficiency is 29.43%.This study provides valuable insights into the thermal-hydraulic performance of the tritium breeding blanket with a binary gas mixture coolant in fusion reactors. [ABSTRACT FROM AUTHOR]

• MAXFEA 2D axisymmetric plasma disruption modelling for SPARC tokamak. • Self-consistent FEM modelling approach used to characterize SPARC plasma evolution. • Development of predictive simulation database as disruption synthetic experiments. • Preliminary description of a vacuum vessel synthetic displacements diagnostic. The evaluation of electromagnetic (EM) behaviour of tokamak components under disruption loads and its correlation with vertical instabilities is critical and characterizing the design considerations. This is particularly crucial for high-field and burning plasma experiments such as the SPARC tokamak, the compact prototype currently under construction by Commonwealth Fusion Systems (CFS). The combination of high toroidal magnetic field (12.2 T) and high plasma current (8.7 MA) makes SPARC less tolerant to disruptions than currently operating research devices. In this context, recent efforts have been made to characterize the disruption loads in SPARC by conducting a series of analyses to estimate the EM response of the SPARC vacuum vessel (VV) during a plasma vertical displacement event (VDE). In this paper, the aforementioned analyses are used as starting point for a comprehensive characterization of VDE scenarios. These analyses are utilized to create a database of predictive disruption simulations, serving as synthetic experiments to generate and collect pertinent data for upcoming research studies. In this context, the paper discusses a preliminary description of an innovative VV synthetic displacements diagnostic, designed starting from disruption database, presented here, and by exploiting the MAXFEA-ANSYS combined use methodology as further step. The disruptive plasma scenarios database will be functional in estimating the displacement and strain fields of SPARC components using a limited set of VV indirect displacement measurements and integral plasma observations, such as plasma current and effective plasma position. [ABSTRACT FROM AUTHOR]

• This paper presents the design and computational analysis of the bolometer camera for the MT-I spherical tokamak. The estimation of the line of sight (LOS) and field of view (FOV) of the AXUV system. • Optimization of the aperture by signal-to-noise ratio method and estimation of geometrical etendue. The computation of plasma emissivity distribution using impurity densities and their cooling rates. • The computation of power incident on the photodiode array and signal-to-noise ratio (SNR). The MT-I is a spherical tokamak at the Pakistan Tokamak Plasma Research Institute, and has a major radius of 15 cm and a minor radius of 9 cm. This paper presents the design and computational analysis of the Absolute eXtreme Ultra-Violet (AXUV) photodiode-based bolometer camera, which will be used for the first time in the MT-I tokamak. The bolometer camera contains a single array of 16 AXUV photodiodes placed behind an aperture to compute the line of sight (LOS) and field of view (FOV) of the system. This array is located at the midplane of the vessel, which provides a radial resolution of about 16 mm. The individual FOV of each photodiode and the collective FOV of the array are 4.1° and ±34.5°, respectively. The impurity densities and their cooling rates for elements like carbon, oxygen, and iron are used to compute the plasma emissivity distribution, the power incident on the photodiode array, and signal-to-noise ratio (SNR) analysis. The analysis reveals that the radiated power loss is about 10 % of the ohmic heating input power. The geometrical etendue falls within the range of 2.2 × 10−5 to 3.2 × 10−5 Sr cm2. The bolometer exhibits a frequency response bandwidth of 10 kHz, and its temporal resolution is around 0.1 milliseconds. [ABSTRACT FROM AUTHOR]

• Saturated-wave theory for snubber modeling is modified to fix three conceptual errors. • The equivalent resistance and inductance of the snubber are recalculated. • Calculating the thickness of eddy-current region no longer relies on external circuit. • The current rise of the model is nearly the same as the experiment with about 10 % over-estimated peak shorting current. Core snubber is a common device used for breakdown protection in auxiliary heating systems. It absorbs energy from the high-voltage power supply to protect the load. The existing snubber models are based on saturation-wave theory, but the theory still has several conceptual problems. In this paper, an improved saturated-wave theory is presented. We consider that the eddy current region of the tape needs to be layered and that there is an unsaturated region in the tape which enables the snubber to be equated to an inductor. Then we divided the working process of snubber into 4 stages and rederived the circuit model of the snubber. The comparison with the existing simulation model shows that the new model in this paper is closer to the experimental data, especially in the rising phase of short-circuit current. It can more accurately analyze the conditions within a few microseconds after breakdown, and give a useful reference for the design of breakdown protection and power supply systems. [ABSTRACT FROM AUTHOR]

In the design of fusion reactors, extensive shielding calculations are necessary to verify the feasibility and safety of the design. The issue of deep penetration is a common and significant problem in fusion reactor shielding calculations. When the thickness of the shielding layer exceeds a certain value, the statistical error in Monte Carlo method calculations significantly increases. Therefore, it is imperative to adopt variance reduction techniques to enhance the computational accuracy for deep penetration problems. cosRMC is a reactor Monte Carlo code, and its accuracy in calculating neutron-deep-penetration shielding problems should be verified. For this purpose, three deep penetration benchmarks from Shielding Integral Benchmark Archive Database (SINBAD) were simulated by cosRMC and neutron cross section data library ENDF/B-VII.0. These benchmarks describe the penetration behaviors of neutrons in models with different geometric shapes and materials. This paper shows detailed cosRMC models based on benchmark experimental models, used to calculate neutron flux spectra at different penetration depths, which are recorded by detectors in the related benchmark experiments. In this way, a direct comparison between cosRMC results and experimental measurements is provided. In order to improve the accuracy of cosRMC in calculating deep penetration problems, the variance reduction technique of cell importance was used. The neutron flux spectrum calculated by cosRMC was compared with the neutron flux spectrum results calculated by MCNP5 and the experimental values of the benchmarks. The comparison results show that the calculations of cosRMC are in compliance with the MCNP5 reference values and experimental values, and the relative deviations between calculations of cosRMC and MCNP5 is within three time of standard errors. Therefore, this paper has verified that cosRMC can applies to neutron deep penetration problems of fusion reactors. [ABSTRACT FROM AUTHOR]

• A 10° of freedom manipulator driven by electro-hydraulic hybrid is introduced. • Geometric and numerical solutions are combined to analyze the inverse kinematics of redundant manipulator. • The dynamics of the redundant manipulator is analyzed by the inward-iteration method based on newton-euler equation. Remote handling (RH), as one of the key technologies for fusion reactors in the future, has been the focus of research by researchers. This paper presents the design and analysis of a redundant manipulator with 10° of freedom (DOF) intended for use in vacuum vessel of the Tokamak. With a span of 3.15 m and a payload of 10 kg, the manipulator is powered by a combination of hydraulics and motors. The manipulator is mainly composed of the macro-operating arm, the micro-operating arm, and an end-effector, each serving its specific functions. This paper introduces and discusses the roles and capabilities of these components in the manipulation process. The finite element analysis (FEA) of the manipulator's structure reveals that the maximum stress of the structure is significantly lower than the permissible stress of the material Aluminium alloy 2A14. By utilizing the modified Denavit-Hartenberg (DH) method, the kinematic model was created, the forward kinematic matrix was established, and the workspace was analysed accordingly. The inverse kinematics solution was elaborated by the combination of geometric and numerical solutions. Given velocity and acceleration of the joints and the mass of the links, the force and torque of each joint were computed by the inward iterations method based on the Newton-Euler equation, and the rationality of the results was verified by ADAMS simulation. The results indicate that the manipulator, with sufficient strength and appropriate kinematic and dynamic properties, can fulfill the design objective and functional requirements. [ABSTRACT FROM AUTHOR]

• ITER TBM Program plans to test 4 DEMO-relevant tritium breeding blanket systems. • There are two water-cooled and two helium-cooled Test Blanket Systems (TBS). • Each TBS is formed by various sub-systems with different functions and technologies. • The paper describes the on-going associated R&D and the main testing objectives. The ITER TBM Program foresees that two ITER equatorial ports are dedicated to the simultaneous operation and testing of mock-ups of four different concepts of tritium breeding blankets (TBB) ensuring tritium breeding self-sufficiency and extraction of high-grade heat for a demonstration fusion reactor (DEMO). The paper starts with the description of the four Test Blanket Systems (TBSs) for the initial configuration and the main on-going R&D carried out by the ITER Members (IMs) in support of the TBS designs. The second part addresses the main objectives of the ITER TBM program to be obtained during the testing campaigns in ITER in order to demonstrate the feasibility of the corresponding DEMO TBBs. It also addresses the main technological data/information that are expected to be obtained during the TBSs design and manufacturing phases and relevant for the DEMO breeding blanket designs. [ABSTRACT FROM AUTHOR]

• Challenging optical design with 120° field of view at high spatial resolution. • Tight position tolerances of optics achieved using laser tracker and adjustment parts. • High spatial resolution demonstrated using MTF test over entire field of view. • Successful operation of endoscopes in W7-X, enabling real-time divertor protection algorithm. • IR images biased by plasma radiation at 4.06 µm wave length. The modular stellarator Wendelstein 7-X (W7-X) in Greifswald (Germany) started operation in 2015 with short pulse limiter plasmas and continued with pulsed divertor plasmas in 2017–2018. In 2022, operation phase (OP) 2.1 was run after installation of ten water-cooled divertors, allowing for steady state operation. Each water-cooled divertor needs to be monitored by an endoscope to interrupt plasma operation once overload is detected. The endoscopes transmit infrared (IR) radiation form the plasma-facing surface and plasma edge radiation in the visible (VIS) range from a pinhole in a water-cooled plasma facing head to the rear side outside the vacuum, where the light is split and captured by an IR and VIS camera. The challenge is to reach a high-resolution image while capturing the large field of view (FOV) of ∼120° of the entire target. In OP2.1, experience was collected with the operation of the first two endoscopes, which were installed in the so-called AEA ports. In a previous paper (Fellinger et al., 2020) the design of these endoscopes was presented. The current paper describes their manufacturing, installation, commissioning and operation. Lessons learnt are highlighted which are being considered for the other endoscopes in the AEF ports. Those will be installed in 2025. [ABSTRACT FROM AUTHOR]

• The magnitude of current and electromagnetic force was calculated through simulation analysis combined with theoretical formulas. • The target plate structure was calculated through finite element analysis under the influence of eddy currents and halo current electromagnetic forces. • Transient dynamic analysis was conducted using modal superposition method. • Based on the safety performance evaluation criterion 3Sm for internal components of fusion devices, the safety performance assessment of the structure was conducted. China Fusion Engineering Test Reactor (CFETR) is a fully superconducting tokamak device independently designed and developed by China. The divertor of CFETR is one of the core components, which will withstand huge and complex electromagnetic loads. Electromagnetic loads are often of large magnitude and short duration, which will cause impact damage to internal components.During the plasma disruption event, the plasma current drops rapidly in milliseconds, which creates large eddy currents on the divertor, and eddy current will generate electromagnetic force and torque under the action of toroidal and poloidal magnetic field, which will cause the divertor to be pulled, compressed or twisted; during the plasma vertical displacement event, halo current is generated on the plasma surface, then a large electromagnetic force will be generated on the divertor under the action of the toroidal and the poloidal magnetic field.In this paper, the electromagnetic analysis under the action of halo current and eddy current is carried out for the hypervapotron structure, and transient dynamic analysis of the real stress state during the electromagnetic load impact process is simulated.Strength analysis based on the results of transient analysis shows that the structure has a high reliability. The work in this paper is based on the evaluation of reliability of the hypervapotron structure under the maximum electromagnetic load. [ABSTRACT FROM AUTHOR]

In this paper we report on the development and demonstration of density feedback control for MAST-U. Sinusoidal perturbations are used to measure the frequency response from a deuterium gas valve (actuator) to line-integrated core electron density measured by the interferometer (sensor). In the frequency range relevant for control design, only two system-identification experiments were needed to regress a first-order dynamic model. This control-oriented model informs the offline design of a proportional integral controller with the established loop-shaping controller design method. After offline verification of the controller implementation, control is demonstrated by experimentally tracking a staircase reference for the line-integrated electron density. This paper demonstrates the efficiency of controller design using system-identification and loop-shaping, providing reliable density control for MAST-U. [Display omitted] • Gas valve calibrations on the Upgraded Mega Ampére Spherical Tokamak (MAST-U). • System identification in the frequency domain to precisely measure dynamics. • Feedback controller design for core plasma density on MAST-U using two discharges. • Dynamics from gas injection to integrated density behave linearly from 7 Hz to 40 Hz. [ABSTRACT FROM AUTHOR]

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. • Concept of operation of a remote handling (RH) system for ITER vacuum vessel. • Remote handling system working in high radiation environment. • Multi degree of freedom heavy load robotic arm. • Efficient in-vessel component and tool exchange. • Rescue of the failed RH system. [ABSTRACT FROM AUTHOR]

The International Thermonuclear Experimental Reactor (ITER) magnetics diagnostic system will exploit high performance signal processing, network streaming and telemetry capabilities. Under this perspective, the adoption of the Multithreaded Application Real-Time Executor (MARTe2) control framework appeared as the most performing solution. To this aim, MARTe2 must be deployed over specific System-on-Module (SoM) platforms. In this paper, we focus on the porting of MARTe2 for the ARM® processor architecture, since it allows MARTe2 to run on a vast range of embedded platform. A series of further integration steps with respect to the state-of-the-art must be undertaken to deploy and run the complete solution. The paper will present the implementation and integration steps followed, giving also context about the application and motivations on the choice. As the target platform uses a quite standardized reference design, the described procedure can be adapted to a large number of System-on-Module based embedded boards. Specific platform adaptations, achieved using custom-developed components, offer another foundation for the further integration of the framework. Moreover, the ITER magnetics diagnostic system is required to run in real-time, therefore the implementation leverages the Asymmetric Multi-Processing (AMP) approach. AMP, in contrast with Symmetrical Multi-Processing (SMP), enables the central processing unit to run each core isolated from the others, without any operating system or orchestration layer in-between. Considering the strict quality assurance and continuous integration constraints required by the MARTe2 framework development process, the additional build system will be fully described. The described approach is based on the generic MARTe2 framework, ported on widespread ARM® architecture, and it could be easily repurposed for all scenarios where similar real-time signal acquisition, processing and streaming are required, thus enabling the rapid deployment of real-time control platforms with off-the-shelf components. • MARTe2 represents a standard in realtime control of complex plants, such as nuclear fusion plants. • The paper describes the steps needed to integrate MARTe2 with a real world embedded platform. • The discussion covers the relevant aspects including AMP scheme, shared memory inter-core communication and boot sequencer. • The paper aims at representing a complete porting guide of MARTe2 on ARM platforms. [ABSTRACT FROM AUTHOR]

The In-Vessel Inspection System (IVIS) is remote handling in-service system to perform visual inspections of the SST-1 scale tokamak under vacuum and high temperature in between the plasma shots. The IVIS system is approximately 4 m in length and has 5- degrees of freedom (DOF), consisting of four rotary joints and one linear motion for deployment inside the tokamak. The IVIS system is designed to support a cantilevered payload of approximately 1 kilogram while maintaining a positional accuracy of <2 mm. Initially, the IVIS system is stored in a vacuum chamber that is approximately 5 m long and maintained under Ultra-High Vacuum (UHV) conditions. This storage vacuum chamber (SVC) is separated from the Vacuum Vessel (VV) by a UHV gate valve. When performing visual inspections, the gate valve is opened, allowing the IVIS to be deployed within the VV. It then carries out the inspection procedure and subsequently returns to its initial position outside the VV. IVIS structural components are five links of arms made of aluminium alloy 6061-T6 and base link, lugs, stiffeners and shafts made of SS 304. Each module can provide rotation (±90° in the horizontal plane). With varying lengths, simulations were performed in virtual reality model to arrive at the present configuration. Based on the VR simulation, it was found that IVIS with a total of 5 links, 1 fixed link & maximum link lengths of 0.7 m (joint to joint) is sufficient to encompass the 180° sector inside SST-1 like machine. This paper offers an overview of the conceptual design, development, and testing phases of the IVIS and SVC to perform visual inspections on the SST-1 scale tokamak while it is under vacuum conditions in the intervals between plasma shots. The paper presents theoretical calculations, a comprehensive kinematic assessments, and structural integrity analyses. Additionally, it discusses strategies for optimizing the design based on these results. A feasibility assessment was conducted to evaluate the appropriateness of technologies for operating the IVIS under conditions resembling those of SST-1, including vacuum and temperature. The IVIS components are designed to withstand temperatures of approximately 100 °C during VV inspection and vacuum conditions (with a pressure of less than 1 × 10−8 mbar) for conditioning before entering the VV. The restrictions on out-gassing within the VV impose significant limitations on the design. The COTS components selection and availability (actuators, bearings, encoders, vacuum feed-throughs & cable tray) to match the performances compatible with high temperature, UHV requirements and to overcome pollution issues of the tokamak environment are discussed in this paper. [ABSTRACT FROM AUTHOR]

The purpose of this work was to establish an approach to drive the development, at pace, of a commercially relevant Spherical Tokamak Concept. Building on the table of parameters and idealised geometry generated from the work of S. Muldrew et al. [1] , applying core principles, tools, and approaches. The generation and utilisation of a reflective 3D Spatial Model to represent the current concept, now forms a strong basis for identifying, evaluating, and delivering future integrated decisions, developed within a 16 Month timeframe. STEP is an ambitious programme, tasked with delivering a prototype plant by 2040 as a pathway to commercial fusion energy. Tackling the integration of a system that itself is based on emergent technologies, significant assumptions (e.g. Plasma Physics) and uncertainties is a significant challenge whilst maintaining acceptable levels of commercial relevance and feasibility. This paper shares the developments and approaches to collaboration and communication taken, including the management of a graphical representation (CAD or Computer Aided Design) to overcome this challenge. The work conducted has resulted in a single point CAD (utilising CATIA software) model representing the current integration status including an appropriate approach to configuration management. Clear channels and methods of communication nurtured a successful collaborative and cross-functional integration approach. Transitioning swiftly from an externally driven approach to integration, to an inherent, self-orchestrating one across all system areas. This paper presents results as a single point Concept Design for the Integrated Tokamak, a strategic management approach, technological integration challenges (e.g. Services, Maintenance, Neutronics) faced, and considerations for the future (post concept design). Introducing a centralised design process, appropriate control and clear communication utilising CAD has been the key to maturing the STEP Tokamak Concept. Initial focus on facilitating collaboration and utilising this environment as a 'sandbox' enabled early establishment of a 'space-claim' concept design in which to iterate. Staging the integration approach and CAD model fidelity developments through 'Identification, Realisation and Accommodation' built a resilient foundation to progress through the full design process. This approach has enabled early identification of not only where the integration challenges lay but also their significance, resulting in strategic deployment of valuable resource. [ABSTRACT FROM AUTHOR]

• Plasma neutralisers can significantly improve the global efficiency of a negative ion based neutral beam injector. • Operation of a neutral beam injector on a fusion reactor imposes many constraints on the design of the injector and the plasma neutraliser, and materials compatible with the expected radiation environment, gas flow into the injector, and space constraints are very important for the design and operation of the plasma neutraliser. • The gas flowing into an injector is stored locally on the cryopumps. The total amount of D2 gas stored in the injector is strictly limited by the possible explosion between D 2 and O 2 if O 2 is present in the injector because of a leak from the surrounding atmosphere. The gas storage imposes periodic regeneration of the cryopumps. Considerations of the regeneration cycle imposes limits on the gas flow into the plasma neutraliser. • A beam driven plasma neutraliser has many advantages over a neutraliser using alternative methods to create the plasma in the neutraliser. If, due to the recombination of electrons and molecular ions, an acceptable plasma density not be achieved with D 2 as the target gas in the neutraliser, the use of argon as the target gas should be considered. The global efficiency of neutral beam injectors to be used on a fusion reactor needs to be of the order of, or greater than, 60 %, which is considerably higher than the efficiency of presently operating NBIs. The required efficiency can only be achieved by increasing the neutralisation efficiency from the ≈56 % obtained with the gas neutralisers currently used, which can be done using a plasma or photon neutraliser. This paper examines possible designs of a plasma neutraliser taking into account the many constraints imposed by operation in the environment around a fusion reactor, and a schematic design is proposed that meets those constraints. The design chosen uses a beam driven plasma neutraliser, which requires no additional power systems, diagnostics or interlocks. Using a plasma neutraliser is shown to improve the global efficiency to ≈55 %. Although that is below the desired value, reasons why that might be acceptable are discussed. This paper also discusses the possible use and benefits of such a neutraliser as an upgrade of the ITER heating neutral beam system. [ABSTRACT FROM AUTHOR]

ITER needs a Disruption Mitigation System (DMS) for protection from the consequences of plasma disruptions during high-power operation. The current DMS for ITER is based on the Shattered Pellet Injection (SPI) technology. This works on the basic principle of a cryogenic pipe gun with around a half-meter-long acceleration barrel. It is important to avoid any contact between the pellet and the internal surfaces of the flight line to minimize the risk of premature pellet breakage. Therefore, the internal dimensions of the flight tube of the SPI are relatively large compared to the pellet hence providing a bypass for propellant gas resulting in a significant quantity arriving in the plasma ahead of the pellet. This will cause an undesirable instability in the plasma and compromise the effectiveness of disruption mitigation. The described events appear on a scale of a few milliseconds. One possible solution for this issue could be using a fast-acting shutter in the flight line, which will close after the pellet has passed through to hold back the propellant gas. Due to the environmental requirements, the shutter needs to close an approximately 40 mm aperture within a few ms, the shutter has limited space and limited access and acts under a magnetic field, therefore it should endure a high number of cycles and should perform well in a very demanding situation. This paper describes the optioneering study of the shutter and possible solutions for the demanding criteria. The laboratory testing of the Electromechanical Test Bench and the Prototype design are also detailed in separate papers. [ABSTRACT FROM AUTHOR]

• We have devised a composite gear transmission system for the rotary joints' driving systems of the CMOR. • The engineering design of the joint driving systems for the CMOR has been successfully completed. • We have proposed fast testing methods for the transmission error and backlash error of the driving system. • The manufacturing and testing of joint J6's driving system for the CMOR have been successfully completed. This study focuses on the engineering design and testing of the rotary joints' driving systems for the China Fusion Engineering Test Reactor (CFETR) Multi-Purpose Overload Robot (CMOR). CMOR is a critical component in the CFETR. The CFETR tokamak serves as an engineering test reactor in China for magnet confinement fusion research. CMOR, a 7-DOF manipulator, is specifically designed for internal maintenance within the CFETR. This poses significant challenges in the design of the joint driving systems. The joint types of CMOR are categorized into three, comprising one translation joint, two swing joints, and four rotary joints. Using rotary joint J6 of CMOR as an example, this paper provides a detailed exploration of the engineering design and testing of the rotary joints' driving systems, offering a comprehensive explanation of their design specifications and principles. Specification and strength assessments are conducted on the engineering design results, meeting all requirements. Furthermore, the paper delves into the testing methods of the driving systems, including the assessment of transmission error and backlash error, two crucial indicators for the driving system. This research contributes valuable insights for future joint designs and testing of the CMOR heavy-load manipulator. [ABSTRACT FROM AUTHOR]

The ITER diagnostic Equatorial Modular Port Plugs (EPPs) are located in the equatorial vacuum vessel ports. Port plugs consist of structural components, shielding elements, services and tenant systems and they are designed to operate in a harsh nuclear environment. Main functions of the port plugs are to provide nuclear shielding and to host tenants systems. Shielding performance optimization and calculation of heat release across the port plugs and on their systems are both crucial for technically sound integration layout of a port plug. Also, assessment of radiological waste is necessary for the decommissioning procedure. Neutronic analyses presented in this paper are focused on the standard shielding and integration solutions on a representative Equatorial Modular Port Plug of port number 8. This Port Plug hosts the Disruption Mitigation System (DMS) and other five diagnostic systems distributed in the three Diagnostic Shielding Modules (DSM): Lost Alpha Monitor, Tangential Neutron Spectrometer, Visible Spectroscopy system, Density Interferometer Polarimeter and the Flow Monitor. Highly detailed MCNP model of the EPP#8 has been developed and integrated in the ITER 40° MCNP C-Model. The model has been used to calculate the relevant nuclear quantities: neutron and gamma fluxes distribution, nuclear heating in the EPP#8 components, tritium and helium production at the ITER end of life and the displacement damage will be presented in the paper for the In-Vessel region. [ABSTRACT FROM AUTHOR]

The ITER WAVS (visible and infrared Wide Angle Viewing System) provides measurements of surface temperature for the plasma facing components by infrared thermography. One of its most important roles of WAVS is to protect the plasma facing components from damage. It also takes images of the plasma visible spectral range emission. The WAVS diagnostic consists of 15 views distributed in four equatorial ports (3, 9, 12 and 17), and it will contribute to Machine protection, Basic control and Physics analysis. The final design of the WAVS system in the equatorial port 12 (EP12) includes three lines of sight (LoS) that have to installed for ITER first plasma. This paper is focussed on the final design for the opto-mechanical ex-vessel components in equatorial port 12 (EP12). The most important function of the ex-vessel optics is to relay the image produced by the Port Plug optics at the vacuum window, up to the detectors placed at the backend of the optical chain in the Port Cell, while keeping the required performance. The image relays through the Interspace, Bioshield and Port Cell covers around 10 m in length before reaching the cameras. The ex-vessel optics is arranged in several optical modules, assembled on the EP12 support structures, i.e. Interspace Support Structure (ISS) and Port Cell Support Structure (PCSS). A detailed description of the opto-mechanical design that supports the optical modules, including the alignment capabilities, will be presented in the paper. Moreover, the assembly and alignment strategies planned for the ex-vessel subsystems, taking into account all constraints imposed on the site, are also described in the paper. [ABSTRACT FROM AUTHOR]

This paper presents the experimental results of the pressure drop through the cooling pipes of the ITER Thermal Shield (TS) panels. High pressure and room temperature nitrogen gas is selected for the experiment, which is equivalent operating condition with the 80 K helium gas from fluid mechanics point of view. Flow rate is set by a mass flow controller and pressure difference between the inlet and the outlet of the cooling pipe is measured by a differential pressure gauge. The experimental set-up is open loop type, which is connected to high pressure gas bottle. Flexible corrugated pipes are used to connect the TS cooling pipes with the measurement system line. The effect of the corrugated pipe is evaluated separately and excluded from the measurement results of the TS pipes. Test results of several TS panels are shown in this paper and are compared with the calculation using existing friction factor correlation and local loss coefficient of the pipe bends. The limitation of the calculation is discussed considering the hydrodynamic developing flow in the pipe routing. [ABSTRACT FROM AUTHOR]

• The control system for the quench protection unit is crucial for ensuring the safety of superconducting magnet coils. • The control system for the quench protection unit is used to control high-power DC switch action to protect the safety of magnet coils. This paper introduces a control system design method for the quench protection unit with soft and hard separation. • The software design and hardware selection of the control system for the quench protection unit are analyzed in detail. • A cross-core scheduling algorithm is proposed for fast task prioritization to meet the hundred-microsecond action requirement of high-power DC switches. Quench Protection Unit (QPU) is required to immediately control the high-power DC switch when the quench event occurs in the superconducting magnet. The energy in the magnet coil is quickly transferred to the Fast Discharge Resistor (FDR), which dissipates a large amount of stored energy and protects the magnet coil from damage. To ensure reliable protection of magnet coils, this paper proposes a design scheme of a soft-hard separation control system based on The Windows Control and Automation Technology (TwinCAT3). According to the high-power DC switch requirement for hundred-microsecond operation in QPU, this paper proposes a fast task priority cross-core scheduling algorithm. Tasks with short cycle times are prioritized and allocated to different cores. The experiment results demonstrate that the control system has excellent accuracy, speed, and reliability. Therefore, it can meet the high-power DC switch action requirements in QPU. [ABSTRACT FROM AUTHOR]

• Development of the diagnostic rack and the locking mechanism for the ITER vacuum vessel lower port. • Addressing manufacturability issues and adapting to the ITER environment. • Accommodating a wide range of reaction forces and improving positioning repeatability. • Feasibility of the design considering manufacturability and uncertainties, thanks to its robustness. The development of the diagnostic rack and the locking mechanism, to be placed in the ITER vacuum vessel's lower port, are in progress. This paper presents the details of these development activities. This paper addresses solving manufacturability issues encountered in previous designs on the diagnostic rack and confirming the validity of the locking mechanism in the ITER environment. The locking mechanism must generate reaction forces to lock the diagnostic rack in the lower port by bending a beam structure, considering robustness, is determined using thermal-structural analysis. The range of the reaction forces, that the locking mechanism can accommodate considering variation due to loads, interface structure, and uncertainties such as frictions are broad, approximately 150 kN–325 kN via 3–7 mm bending displacement of the beam structure. Subsequently, a mock-up of the locking mechanism was manufactured and tested to assess reaction force and positioning repeatability, which is necessary for ensuring diagnostic measurements. The reaction forces were balanced with assumptions of uncertainties, such as friction force and positioning misalignment. The positioning repeatability was improved while increasing the reaction force, the positioning repeatability is 0.49 mm as the two standard deviations (2σ) while satisfying a requirement of +/- 1 mm. We conclude that the design is validated, even when considering realistic manufacturability and uncertainties such as friction and dimensional measurement accuracy, due to the locking mechanism's robust design. [ABSTRACT FROM AUTHOR]

One of the two most promising Breeding Blanket (BB) concepts to be chosen as driver design of the EU-DEMO fusion reactor is the Water-Cooled Lead Lithium (WCLL) BB. A crucial point of this key component is the cooling of the BB structural elements and of the Lithium-Lead alloy, used as neutronic multiplier and Tritium breeder. Indeed, the high neutronic flux needed for breeding implies high volumetric power affecting the Breeding Zone (BZ) materials. The current BB layout employs water-cooled Double Wall Tubes (DWTs) with a C-shaped configuration for the BZ cooling. Despite the WCLL BB have reached a mature design in the last years, some open issues remain to be solved to increase the reliability and performance of this technology. The present paper describes a promising upgrade of the BZ cooling layout adopting helical-shaped DWTs. This solution has the potential to increase the BB reliability, cooling performance, Tritium Breeding Ratio and cooling water flow-path simplicity. The new proposed layout is described in the paper, along with results of CFD analyses carried out to investigate the new cooling performances in BZ. Advantages and drawbacks with respect to the current BB layout have been also highlighted in the paper. [ABSTRACT FROM AUTHOR]

In this paper we describe and show the results of a novel activity on the validation of numerical codes for the low frequency electromagnetic analysis of ITER components. The aim of the activity is to propose a set of problems focused on induced currents in nonlinear materials and on hysteresis behaviours. For this purpose, already existing benchmark problems are considered and new ones are developed for specific investigations. The TEAM (Testing Electromagnetic Analysis Methods) problems 10 and 32 from the COMPUMAG society have been tested, together with an ad hoc problem proposed by one of the authors. Four computer codes (EleFAnT3D, ANSYS EMAG, ANSYS Maxwell, COMSOL Multiphysics) have been used and results obtained have been compared. Globally a good agreement among results obtained with the numerical codes and also with experimental data has been found, but some small deviation and peculiarities have been observed which are discussed in the paper. [ABSTRACT FROM AUTHOR]

• The breakdown model of NBI system is improved, which contains more accurate process. • Some model parameters are corrected based on the design schemes of ITER and CFETR. • The need and required parameters for protect against breakdown are evaluated. Breakdown commonly occurs in the acceleration grids of the NBI system. It will cause destructive effects on the accelerating grids, and the generated transient overvoltage and current will threaten the normal operation of the acceleration grid power supply (AGPS). Therefore, it is very necessary to model the circuit to analyze the effects of breakdown on the NBI system. However, the current models are not accurate enough and are likely to have insufficient knowledge of the consequences of breakdown. In this paper, a series of improvements are made to the model by analyzing in detail the change process of AGPS during breakdown and correcting the model parameters in conjunction with the existing AGPS design scheme. Subsequently, the key indicators of the NBI system under breakdown are listed. The impact of breakdown and the need for some protection are assessed based on the results of these indicators. The results of this paper are instructive for the design and breakdown protection of future NBI systems. [ABSTRACT FROM AUTHOR]

The design of future fusion reactors involves the production of tritium inside the breeder blanket. The most promising material for solid breeder blankets is a two-phase lithium ceramic containing orthosilicate Li 4 SiO 4 (LOS) and metatitanate Li 2 TiO 3 (LMT) of lithium in various proportions. Tritium is formed in lithium under neutron irradiation by the reaction 6Li(n,α)T. Further, this tritium is extracted from the blanket with a purge gas and returned to the fusion zone, realizing the concept of a closed fusion cycle. Irradiation under fission reactor conditions is still one of the few available methods for estimating the parameters of tritium generation and release from lithium-containing materials in the "in-situ" mode. This paper presents the results of experiments on neutron irradiation of two-phase lithium ceramics of various ratios (LOS + 35 mol. % LMT (pebble size 250–1250 μm), LOS + 35 mol. % LMT (pebble size 500–710 μm) and LOS + 25 mol. % LMT (pebble size 500–710 μm) at the WWR-K research reactor. Irradiation of each batch of samples lasted from 5 to 22 days. The experiments were carried out by the vacuum extraction method. This paper describes the main methodological aspects of the studies, namely the technical features of four irradiation campaigns, the sequence and scope of the studies. A comparison is also made of the initial sections of reactor experiments for all campaigns, where the reactor was sequentially brought to power, according to which the parameters of the Arrhenius dependence of the effective tritium diffusion coefficients were estimated. [ABSTRACT FROM AUTHOR]

• The Beam Line Vessel and its components are modeled by Solidworks, the vessel pressure distribution, and adsorption uniformity of the crysorption pump are analyzed by simulation using the Molflow software. • The performance of the south cryopump is tested experimentally using in-situ measurement method. • With the continuous injection of gas, the adsorption probability of the cryopump decreases, so the pumping speed of the cryopump will decrease. • Due to the limited capacity of cryopump adsorption, it is necessary to regenerate the cryopump near the maximum. The cryogenic vacuum system is an important subsystem of Comprehensive Research Facility for Fusion Technology Negative Neutral Beam Injector, which provides vacuum environment support for various experiments related to the beam generation and transport process. In the paper, a simulation and experimental study of the Comprehensive Research Facility for Fusion Reactor Negative Neutral Beam Injector (CRAFT NNBI) cryopump performance has been carried out. The Beam Line Vessel (BLV) and its components are modeled by Solidworks, the pressure distribution, gas molecular velocity distribution and adsorption uniformity of the crysorption pump in the BLV are analyzed by simulation using the Molflow software, at the same time, the performance of the south cryopump is tested experimentally using in-situ measurement method. The simulation results show that the average pressure near the Neutralizer is 8.5*10-3 Pa, the average pressure near the Electrical Residual Ion Dump (ERID) is 8*10-3 Pa, and the average pressure near the Calorimeter is 9*10-3 Pa. The adsorption percentage fluctuates around 12%, which indicates that single structure plays the function of adsorption efficiently, the theoretical pumping speed of the south cryopump on H2 was calculated to be 7.34*105 L/s, total theoretical pumping speed of two cryopumps can reach 3.67*106 L/s. The experimental results showed that when the south cryopump was operation, the pressures near the Neutralizer were 2.8*10-2 Pa, the pressures near the ERID were 2.7*10-2 Pa, and the pressures near the Calorimeter were 7*10-3 Pa, the actual pumping speed of the south cryopump on H2 was 6.37*105 L/s, total actual pumping speed of two cryopumps can reach 3.1*106 L/s. [ABSTRACT FROM AUTHOR]

This paper presents the results of experimental tests on samples made of copper coated with alumina layer, performed to assess the reliability of its dielectric properties for applications in the low temperature plasma at the edge of a fusion device. The cue of the study was related to the plasma facing components of the RFX-mod2 fusion device (Marrelli et al., 2019, Peruzzo et al., 2023, Peruzzo et al., 2019), devoted to the experimental study of the magnetic confinement of fusion plasmas in a variety of configurations, including the reversed-field pinch and the tokamak. In RFX-mod2 an in-vacuum copper shell for the passive stabilization of MHD modes will surround the plasma. To avoid potentially harmful electrical discharges, which could be induced by rapid transients of the plasma current, this structure must be covered with an electrically insulating layer. For RFX-mod2 an alumina coating was chosen, whose dielectric properties have been tested both in air and in the presence of weakly ionized plasma. Electrical tests, conducted on copper samples with alumina deposits of about 100 μ m thickness, revealed that the ceramic layer has a high electrical resistance value in air (> 1 G Ω), but electrical discharges can occur in presence of a weakly ionized plasma, depending on compactness and porosity of the alumina layer, causing local melting of the alumina and expulsion of copper droplets from the substrate. Scanning Electron Microscope (SEM) analyses revealed that in the failed samples the ceramic layer was irregular and rough, with interconnected cavities and cracks, which could reduce its effective thickness and explain the dielectric breakdown at relatively low voltages (< 400 V). The analyses also showed that samples with a more compact layer present a higher dielectric strength in the presence of the plasma, highlighting that compactness and porosity play crucial roles in ensuring good insulation for materials in a plasma. This study led the definition of the requirements for the insulating coating of the plasma facing components of the RFX-mod2 fusion machine, however the results can be useful for other fusion and non-fusion plasma applications requiring electrical insulation, which can span from industrial devices to spacecrafts. • The RFX-mod2 design requires insulating the copper shell to prevent harmful discharges • Alumina deposit on copper has shown high resistance in air but breakdowns in plasma • SEM analysis shows alumina's microstructure impacts electrical insulation in plasma [ABSTRACT FROM AUTHOR]

• New high-field side magnetic probe array system is developed on EAST tokamak. • The 3D magnetic probe provides additional toroidal magnetic fluctuation measurement. • New system can measure fluctuation with higher frequency and toroidal mode numbers. • Main subsystems of the high field-side magnetic probe array system are introduced. • The calibrations and preliminary applications of the new system are presented. To better study the magnetic fluctuations in the high-field side of EAST tokamak, a new high-field side magnetic probe array (HFS-MPA) has been developed on EAST recently. The HFS-MPA consists of 12 identical three-dimensional (3D) magnetic probes, which are mounted on the HFS wall with carefully designed arrangement. The HFS-MPA magnetic probes bring additional toroidal magnetic fluctuation measurements compared with the HFS regular magnetic probes which can only provide poloidal and radial magnetic fluctuation measurements. The upper limits of frequency and toroidal mode number (n) measurements of the magnetic fluctuations have been improved by comparing HFS-MPA with the HFS regular magnetic probes, i.e., 650 kHz vs 100 kHz and n = 23 vs n = 1. In developing the HFS-MPA diagnostic system, many practical challenges have been overcome and many special designs have been developed. These will be mentioned in the main subsystem description of the HFS-MPA diagnostic in this paper, which might be useful for developing new magnetic probe diagnostics in the future on EAST or other magnetically confined fusion devices. The calibration of the effective area and frequency response of the HFS-MPA is also described. The preliminary application in studying the frequency and propagation characteristics of the magnetic fluctuations with HFS-MPA compared with EAST regular magnetic probes shows that the HFS-MPA is well developed for plasma physics studies. [ABSTRACT FROM AUTHOR]

• The main requirements of Configuration Management platform of the DTT project are defined and the suitable Product Lifecycle Management platform is then selected. • A novel platform that ensures the Technical Integration of the DTT project is proposed and a SWOT analysis is carried out. • The Product Breakdown Structure of the DTT facility is implemented evaluating benefits and limits of the chosen collaborative cloud-based platform. Effective project management methods, tools and working practices shall be applied to facilitate the communication and collaboration among the different institutions involved in the Divertor Tokamak Test (DTT) project. This paper deals with the definition of the configuration management platform architecture enabling technical integration of the DTT system. The first step consists in the identification of main requirements the platform should satisfy, considering the multidisciplinary domains and the geographically dispersed working teams characterizing the nuclear fusion sector and the maturity level of the specific project. Main characteristics of the most advanced Product Lifecycle Management (PLM) tools are identified, their limits and benefits are evaluated, and the suitable PLM platform is selected, satisfying the DTT project needs. The definition of the architecture of the configuration management platform for the DTT project aims at implementing the DTT assembly model in a unique environment able to exchange models and data even developed outside the platform ensuring the congruence of the design, the traceability of design changes and the adoption of a proper Systems Engineering approach. [ABSTRACT FROM AUTHOR]

• Vacuum pressure in the HL-3 vacuum vessel reached to 1.8 × 10-6 pa. • The temperature of the VV is baked by nitrogen to 176 °C. • A solution based on control variable method is used to adjust the temperature difference between the divertor and the VV. • A 100 % zero oil-free, high speed centrifugal fan is developed for the HL-3 baking system. HL-3 is a tokamak device constructed in 2020 and it has achieved remarkable results in a short period of time. Baking is an important way to obtain a high vacuum required for plasma discharge in tokamak devices. A nitrogen gas closed-circuit reheat baking system is utilized to concurrently bake the vacuum vessel (VV) and the lower divertor of HL-3. So far, three rounds of baking experiments have been carried out, with vacuum pressure up to 1.8 × 10–6 Pa and the results from the mass spectrometer indicated that the amount of compounds were greatly reduced in the residual gas. Besides the design description and experimental results of HL-3 baking system, this paper also introduces a simulation-based solution for addressing the issue of the lower divertor temperature being significantly lower than that of the VV after baking to the high temperature stage. [ABSTRACT FROM AUTHOR]

• The paper overviews the LLMHD experimental facility at IPR: among the few integrated PbLi MHD experimental facilities in the scientific fusion community. • We present the results of first MHD experiment performed in the LLMHD experimental facility with a simple circular U-bend test mock up in presence of ∼1.06T magnetic field. • The results of circumferential variations of wall electric potential at selected locations, and pressure drop have been presented. • The impact of MHD effects have been observed in the experimental measurements e.g. flow rate decreases when increasing the magnetic field, wall electric potential and pressure drop increases when magnetic field increases. • The distribution of wall electric potential difference has been found to be sinusoidal in nature along the circumferential direction at the given locations of the test mock up. The Liquid Lead lithium Magneto Hydro Dynamics (LLMHD) experimental facility has been constructed at Institute for Plasma Research (IPR), Gujarat, India to perform various R & D MHD experiments associated with the flow of electrically conducting liquid metal under strong transverse magnetic field. The electromagnet having C-shaped soft iron core has been designed and developed, to provide a uniform magnetic field of up to 1.4T within its polar volume 1000 mm (H) ×400 mm (W) ×370 mm (L). The magnetic field lines are aligned along the length (L). A relatively large polar volume inside the electromagnet to place the test mock up for MHD experiments is its particularity. It enables the study of MHD flows with complex flow geometries and having longer flow length perpendicular to the magnetic field. We have started running the LLMHD loop and the first MHD experiments with Pb-Li have been performed so far at 320 °C in a test mock-up of a basic circular flow geometry having two 90° bends. So far till now, the isothermal MHD experiments have been conducted in the presence of a uniform transverse magnetic field of 0.62T (Ha ∼ 322) and 1.06T (Ha∼551) for the ranges of Reynolds number 20,000–50,000. During the MHD experiments, flow rates, temperature, pressure, and induced wall electric potential have been recorded. The MHD effects on the pressure drop and flow rate has been noticed. The 3D MHD numerical simulation has also been performed, using add on MHD module of ANSYS FLUENT. Both simulation and experimental results of the induced wall electric potential have been compared. [ABSTRACT FROM AUTHOR]

Fast digitizers are employed in a variety of experimental contexts, including for microwave measurements from fusion plasma diagnostics. However, most existing commercial digitizers used for this purpose are severely limited by their onboard memory. Here we present a system developed from mostly commercially available hardware components capable of acquiring essentially indefinitely (here ∼ 10 s) while meeting the target performance of 5 GHz analog bandwidth with a rate of 10 billion samples per second and 8 bits per sample. At its core is a field-programmable gate array (FPGA) receiving data from a high-performance analog-to-digital converter (ADC). The data are continuously streamed with a maximum throughput of 120 Gb/s from the FPGA to a computer over optical fiber in the form of raw Ethernet packets, allowing the use of entirely standard networking hardware in the PC. Whilst this technology is transferable to a range of applications, we are motivated by the demands of microwave scattering measurements, for which the new digitizer increased the acquisition duty cycle from 6% to 100%. In this paper we describe our digitization system, demonstrate its capability, and then use it to acquire data from microwave diagnostics at the ASDEX Upgrade and Wendelstein 7-X fusion experiments. • Presents a continuous ultrafast digitizer (10 GSa/s) with mostly COTS components. • Flexible architecture is future-proof as it can be easily modified or upgraded. • Data streamed on the fly to storage enables acquiring virtually indefinitely. • Successful tests on AUG and W7X reveal previously unidentified plasma signals. [ABSTRACT FROM AUTHOR]

• The results of research and preparation of the KTM tokamak high-frequency plasma heating system for launch. • Calculations of the effective absorption of high-frequency radiation by the plasma of the KTM tokamak. • The results of measurements of the characteristics of the antenna feeder device. • More efficient plasma heating on a KTM tokamak can be obtained at a frequency of 14 MHz. The paper presents and discusses the results of the research on the preparation of the HF plasma heating system of the KTM tokamak for the start-up. The calculations have been performed to determine the parameters at which the most effective absorption of the HF radiation by the plasma of the KTM tokamak was expected, the results of measurements of the frequency response of the antenna-feeder device of the HF heating system generator of the KTM tokamak plasma have been presented. The result of the carried out work have shown that more efficient plasma heating on a KTM tokamak can be obtained at a frequency of 14 MHz. [ABSTRACT FROM AUTHOR]

• Plasma neutralisers can significantly improve the global efficiency of a negative ion based neutral beam injector, which is a requirement for an injector to be used on a fusion reactor. A beam driven plasma neutraliser has many advantages over a neutraliser using alternative methods to create the plasma in the neutraliser. However, due to the recombination of electrons and molecular ions, an acceptable plasma density may not be achieved with D 2 as the target gas in the neutraliser, and it has been suggested that the use of argon as the target gas should be considered. The advantages and inconveniences of using argon are discussed. • The equations governing the species change in a high energy H- beam traversing an argon plasma are identified and solved. • A careful evaluation of the cross sections needed to calculate the change in beam species and the resulting neutralisation efficiency of an argon plasma neutraliser has been carried out, and the cross sections given as a function of the incident beam energy over the appropriate energy range. Where no measurements have been found of some of the required cross sections, assumed values are used and the effect of different assumed values evaluated. • Calculated neutralisation efficiencies are given as a function of the ionisation fraction in the neutraliser as well as the necessary line densities. • The used of a beam driven argon plasma neutraliser as an upgrade of the ITER diagnostic neutral beam injector is discussed. A beam driven plasma neutraliser (BDPN) has been proposed [ 1 ] as a relatively simple way to increase the efficiency of neutral beam injectors based on the acceleration and neutralisation of H- or D-. Initial calculations showed that sufficient levels of ionisation could be achieved with H 2 or D 2 as the initial gas target in the neutraliser if a plasma confinement time of the plasma in the neutraliser is >0.5 ms could be achieved. However, later calculations [ 2 , 3 ] show that the afore-mentioned results were very optimistic as the calculations did not take into account the recombination of molecular ions with electrons. Hence, it has been suggested [ 4 ] to use an argon (Ar) plasma as the molecular ion content in such a plasma will be negligible. The use of an Ar plasma neutraliser has been suggested previously [ 5–7 ], but few details have been given of the obtainable neutralisation efficiency, the necessary line density for that neutralisation efficiency, or of the calculations and cross sections used to deduce those quantities. This paper briefly discusses why Ar could be a good choice for a plasma neutraliser, then the reactions occurring between the H-/D- beam and the particles in the Ar plasma, and the available cross section data for those reactions. Subsequently, the differential equations for the change of the species in the beam as it traverses an Ar plasma are deduced and solved, and the solutions used to calculate the achievable neutralisation efficiency as a function of the line density in the neutraliser etc. Results are given of the neutralisation efficiency as a function of the degree of ionisation in the neutraliser, the required line density, and the species changes in the beam for accelerated H- beams with 100, 500 and 870 keV energies. [ABSTRACT FROM AUTHOR]

In support of the ITER disruption mitigation system (DMS), a highly flexible, triple-barrel shattered pellet injection (SPI) system was installed at ASDEX Upgrade and tested in 240 dedicated discharges in the 2022 experimental campaign. Prior to the tokamak experiments, the system was commissioned and characterised in a laboratory environment. In this paper we discuss the experience gained from 2000 pellet launches on the freezing and launching process for 4 mm and 8 mm diameter pellets made from deuterium, neon and their mixtures. Different amounts of neon inside the pellet as well as the pellet heating recipes show a significant impact on pellet quality and successful launching. Pellet recipes were developed pushing the lower boundary of successful pellet launches. Depending on pellet size and composition, velocities between 60 m/s and 750 m/s were achieved. • Overview of the laboratory commissioning of the ASDEX Upgrade SPI system. • Insights on desublimation process & potential issues based on ultra-high-speed videos. • Presentation of a pellet detachment model for the AUG SPI system & experimental validation. • Reproducible pellet production & prediction of velocities with a precision of up to 1 m/s. [ABSTRACT FROM AUTHOR]

• Designed a multiuser lumped cooling water system for ECRH and ICRF testing bench, which is different from the traditional one system corresponding to one cooling water system. • The water supply test of two users with 400t/h flow regulation is realized to verify the operation mode of the future fusion reactor. • The integrated cooling water system can greatly reduce the construction and maintenance costs of the system and the redundancy of the cooling water capacity. In EAST(Experimental Advanced Superconducting Tokamak), the heat produced by ECRH (Electron Cyclotron Resonance Heating) and ICRF (Ion Cyclotron range of Frequencies Heating) should be respectively removed through independent cooling water system. Their capacities are typically proposed based on the full power of the ECRH and ICRF. Indeed, the operational duration of the wave system is limited, which necessitates a larger cooling water capacity redundancy, occupies a significant amount of space, and requires a substantial budget. In this paper, the cooling water system for ECRH and ICRF of CRAFT (Comprehensive Research Facility for Fusion Technology) testing bench is designed according to its requirements.Initially, the optimization of the cooling water process is carried out in accordance with the specified parameters. Thereafter, the entire process is simulated employing the AFT Fathom code.Finally, a multiuser integrated cooling water system is designed, which is different from the traditional majority one-to-one mode. And the results show that all the parameters can meet the system requirements, meanwhile the redundancy of the cooling water capacity can be greatly reduced by flow regulation.The cooling water integral structure will provide a new idea for the design of fusion reactor cooling water systems. [ABSTRACT FROM AUTHOR]

• A realistic assessment of the DT fuel cycle indicates that ∼327 g tritium will be required for successful plant startup. • Efficient operation of a fusion power plant will require a high level of direct internal recycling. • Improvements to fueling efficiency and burn fraction can greatly reduce startup inventory but may be technologically challenging to achieve. • Plant doubling time is defined to inform the time it takes for a new plant startup inventory to be bred. With the increasing efforts to commercialize fusion power, private and government organizations are investing heavily in the development of technology to support a viable fusion power plant. Deuterium-Tritium (DT) fueled reactors are more prevalent than other proposed designs, requiring tritium processing and handling technology for safe operations and self-sufficiency. Further, each fusion power plant will need a specific-to-design startup inventory of tritium to begin operations. This startup inventory is required prior to breeding and is the minimum tritium inventory required to fill each processing component in the fuel cycle, to offset radioactive decay losses, and to avoid a zero-fuel situation for continuous operation. We present an approach to calculate the startup tritium inventory for a 500 MW th reactor, with considerations for reserve inventory for maintenance and commissioning. A baseline startup inventory was calculated to be approximately 327 gs. This value was obtained using modest assumptions about the technology and operating parameters of a fusion power plant. The required operating reserve inventory or the inventory necessary to keep a fusion power plant operational using only direct internal recycling for 24 h for the same plant design is approximately 642 gs. The approach and findings of this paper will enable fusion energy stakeholders to better utilize the existing scarce global tritium supply. [ABSTRACT FROM AUTHOR]

• This study demonstrates the successful performance of a small ambient molecular sieve bed (AMSB) in adsorbing low partial pressure vapor, confirming its suitability for use in ancillary systems of breeding blanket. • The vapor partial pressure is identified as the dominant factor influencing the full adsorption performance of the AMSB, even at reduced size and low vapor concentration levels. • The cyclic operation of the AMSB is successfully validated, Note the importance of complete desorption to maintain optimal adsorption performance and efficiency in future operating scenarios. This paper investigates the adsorption performance of an ambient molecular sieve bed (AMSB) intended to be designed for ancillary systems of breeding blanket in a nuclear fusion reactor. The study focuses on MS5A as a promising adsorbent for removing low partial pressure Q 2 O impurities in the ancillary systems of breeding blanket. While the absorbent's isotherm is well established for huge columns in the chemical industry, the competence of a small MS5A bed in handling low Q 2 O concentrations within the ancillary systems for breeding blanket is investigated. The adsorption performance on low partial pressure vapors has been investigated through experimental evaluations using a small AMSB with a Research Apparatus for Vapor Adsorption and Desorption, RAVAD. The results show that the adsorption performance of the small AMSB, despite its reduced size, closely follows the absorption isotherm based on the chemical industry. The vapor partial pressure was identified as the dominant factor determining the adsorption performance. Accordingly, an adsorption experiment was performed under the Q 2 O partial pressure of the TES, and although the size was small, 99 % adsorption performance without breakthrough was confirmed for 11 days. In addition, the cyclic operation of the AMSB was successfully validated through a comprehensive evaluation of the total vapor uptake. These experimental results provide considerations for the development of AMSB designs and their operational scenarios for the ancillary systems of breeding blanket. Overall, this study provides a basis for the effective use of small AMSBs for the removal of Q 2 O impurities within fusion reactor blanket ancillary systems. The experimental results are expected to contribute to improve design efficiencies and operational strategies to enable better performance in the removal of Q 2 O impurities from the feed gas. [ABSTRACT FROM AUTHOR]

• The PHA diagnostic installed on W7-X, despite its primary purpose of providing information on plasma impurities, can be also useful in determination of 〈T e〉. • The potential results related to line-averaged electron temperature 〈T e〉 provided by the PHA diagnostic could be also used as comparative material for other systems – such as TS or ECE - that are directly dedicated to determination of central electron temperature T e0. • Under specific plasmas conditions both, the shape of electron density and temperature profiles, as well as, its boundary values (T e0 , T e1 , n e0 , n e1) affect estimated line-averaged electron temperature 〈T e〉. • Under W7-X experimental condition characterized by low electron temperature in the plasma centre, line-averaged electron temperature estimated from PHA spectra is more comparable to the central electron temperature T e0. Electron temperature (T e) is one of the main plasma parameters, which is routinely measured during fusion experiments. The major systems for the Te determination used for tokamak and stellarator plasmas are Thomson scattering (TS) and Electron Cyclotron Emissions (ECE). Measurements of T e can be also performed by other diagnostics such as Pulse Height Analysis (PHA), which is mainly intended for impurity identification. Due to the fact that PHA system is able to provide an information about line-averaged electron temperature (), there is a need to specify experimental conditions under which 〈T e〉 determined from PHA spectra is comparable to the electron temperature from plasma centre (T e0). In this paper, the analysis related to the effect of electron temperature (T e) and density (n e) profiles on 〈T e〉 estimated from spectra collected by PHA diagnostic installed at Wendelstein 7-X (W7-X) is presented. The studies described in this work consists of simulation and experimental results, on the basis of which 〈T e〉 has been estimated for a wide range of experimental condition, that differ in combination of T e and ne profiles. [ABSTRACT FROM AUTHOR]

Fusion reactors are complex systems of interconnected parts, many of which rely on novel technologies. Understanding the patterns of connectivity is essential for many design aspects: assessing technical risks, leveraging modularity, preventing rework and enabling functional integration. The enabling design artefact is the system's architecture, its conceptual model of functions, components and interfaces. In this paper, we present two architecture modelling and analysis methods based on Dependency Structure Matrices (DSM). Firstly, the bottom-up method builds an architecture model from decentralized documentation of all ITER diagnostics and their interfaces. The DSM highlights a modular system topology, integration challenges and opportunities for managing complexity. The limitations of this method, however, motivate the development of a top-down alternative. The top-down method is suited for early-stage design definition, therefore generating valuable insights throughout the development lifecycle. The architecture model serves as a central storage of design information, from which viewpoints can be automatically generated: DSMs, functional flow diagrams, cabling diagrams, interface reports and requirements propagation tables. This model-based information paradigm drastically improves design quality, project planning, change management and stakeholder communication. The top-down method is demonstrated for the Visible Spectroscopy Reference System, one of ITER's first diagnostics, but promises a more widespread application. • A modelling method to define the multi-level components, functions and interfaces of fusion devices. • Automatically generated design documents that improve communication with multidisciplinary stakeholders. • Recommendations for the organization and management of sets of diagnostic systems. [ABSTRACT FROM AUTHOR]