In-vessel Components Research Articles

Machine learning powered predictive modelling of complex residual stress for nuclear fusion reactor design

Fusion In-vessel components, assembled and maintained using laser welding, one of the most promising techniques, exhibit complex distributions of residual stress, microstructures, and material properties. These residual stresses can compromise structural integrity and lifespan of critical components. Although using advanced experimental measurements can evaluate the residual stress for individual case, extending the measurements to massive number of components are costly and time-consuming. Traditional machine learning (ML) models struggle to account for the heterogeneity and anisotropy of these stress distributions. Here, we develop a novel ML framework based on the Eurofer97 steel, the structural material for in-vessel components. The ML framework is trained on high-resolution residual stress data derived from recently-developed evaluation techniques. Combining with microstructures, the model enables prediction of heterogenous and anisotropic residual stress distribution. It successfully predicts the compressive residual stress in fusion zone (∼−200 MPa) balanced by tensile residual stress in heat affected zone (∼300 MPa), aligning closely with experimental results with the R-squared value of 0.989 and the mean square error of 10−4. Unlike experiments that take hours, the ML model provides predictions within seconds. It offers valuable insights into residual stress prediction for various joints, enhancing the reliability and lifetime prediction of in-vessel components.

Design, justification, and prototyping of the visible and infrared wide angle viewing system diagnostic for ITER equatorial port 12.

The ITER equatorial visible and infrared Wide Angle Viewing System (WAVS) will play a major role in the protection of plasma facing components by providing surface temperature measurements of these components. It will also image the plasma emission in the visible range. The WAVS is composed of 15 lines of sight located in four Equatorial Ports (EPs) 3, 9, 12, and 17. Its development is being carried out by the Consortium constituted by CEA, CIEMAT, INTA, and Bertin Technologies, within grant contracts financed by F4E. In the EP12, the in-vessel and ex-vessel components of the WAVS are at their final design phase. This article presents an overview of the opto-mechanical design of the WAVS in the EP12 presented at the final design review.

Initial results of hard X-ray spectroscopy by LaBr[formula omitted](Ce) detector for runaway electron study in Thailand Tokamak-1

Thailand Tokamak-1 (TT-1) successfully achieved its first plasma operation in early 2023. Understanding the behavior of high-energy runaway electrons (RE) during plasma discharges is crucial in TT-1 due to the potential risk of significant damage to in-vessel components. To study the RE behavior and analyze its characteristics, the LaBr3(Ce) detector was employed for measuring hard X-ray emissions in TT-1. In this study, we first characterized the LaBr3(Ce) detector in the laboratory and then performed hard X-ray spectroscopy in TT-1. Calibration sources, including 133Ba, 137Cs, 22Na, and 60Co, with energies up to 1.33 MeV, were used in the laboratory. The detector was calibrated using biased high voltage of -1000 V. It was found to have an energy resolution of approximately 6.2% at an energy of 0.662 MeV. After calibration, the detector was installed at TT-1 to measure hard X-ray. We analyze the hard X-ray emission from discharge #2183 during a selected time interval. It is found that the high-energy hard X-ray emissions reach up to approximately 500 keV. Assuming a simple Maxwellian distribution of the RE population, their temperature is estimated to be 224±5 keV. These findings confirm the presence of high-energy runaway electrons during TT-1’s plasma discharges. However, to accurately derive the runaway electron energy spectrum from the hard X-ray energy spectrum, the unfolding technique is required. In future work, we plan to apply the unfolding method, conduct numerical simulations on the physics of runaway electrons, and employ Monte Carlo simulations on the hard X-ray emissions.

Research on the inverse kinematics of heavy-load manipulator and end-effector for fusion reactor

The environment in Tokamak fusion reactor is hostile to remote handling (RH) equipment that maintains in-vessel components. Many maintenance tasks include transferring components, pipe cutting, and welding in the vacuum vessel (VV). To fulfill these maintenance works, a 7- degree-of-freedom (DOF) heavy load manipulator and a 3-DOF end-effector are designed and connected together to form a 10-DOF redundant robot. The inverse kinematics for such a high redundancy robot, considering obstacle avoidance in the complex space of VV, is a challenge. To solve this problem, the Jacobian method is adopted, and a method specially designed for obstacle avoidance of redundant robots working in the VV is integrated to simplify algorithm calculation and improve inverse solution efficiency. The algorithm is applied to solve the inverse kinematics of the redundant robot for the maintenance of the divertor plasma facing components (PFC) in VV. The results show that the inverse kinematics of the redundant robot can be solved quickly with high positioning accuracy. The method can also be used for other fusion reactor maintenance robots with some adjustments. The maintenance efficiency of redundant robot can be improved with the method which can be valuable for the whole remote handling process.

Recent progress of JT-60SA project toward plasma operation

Abstract Superconducting tokamak JT-60SA plays an essential role in fusion research and development by supporting and complementing ITER project, providing directions to the DEMO design activity and fostering next generation scientists and engineers. Since the incident of the Equilibrium Field coil #1 during the Integrated Commissioning (IC) in March 2021, both EU and JA Implementing agencies (IAs) have examined how to ensure safety operation of JT-60SA by mitigating the risk of possible discharge occurrence inside the cryostat. Based on the experience of the Global Paschen tests, the IAs have established a strategy of risk mitigation measures, which is a combination of (i) reinforcement of insulation, (ii) avoiding unnecessary voltage application to the coil systems and (iii) immediate de-energization of the coils when deteriorated vacuum condition is detected. Thanks to the considerable efforts of the Integrated Project Team (IPT) members, the IC restarted in May 2023. After the confirmation of superconducting state of coil systems (TF, EF and CS), the coil energization test and the plasma operation (OP-1) starts. The first plasma was successfully achieved on 23 October 2023 with a limited value of applied voltage and current to the coils. The plasma configuration control will be also confirmed with low plasma current and low auxiliary heating power conditions. Based on the IO-F4E-QST collaboration, activities of JT-60SA have been shared with the IO and provided an important lesson learned for ITER assembly and commissioning, and will provide an outstanding contribution to fusion research at large. After OP-1, Maintenance & Enhancement phase 1 (M/E-1) starts from January 2024, in which in-vessel components are installed, and heating system and diagnostic system are extensively upgraded to allow high power heating experiment planned in OP-2. In order to make the best use of JT-60SA, newly organized JT-60SA experiment team will refine the research plan in the future high heating power operation phase.

Recent advance progress of HL-3 experiments

Abstract Since the first plasma realized in 2020, a series of key systems on HL-3 (known as HL-2M before) tokamak have been equipped/upgraded, including in-vessel components (the first wall, lower divertor, and toroidal cryogenic/water-cooling/baking/glow discharge systems, etc.), auxiliary heating system of 11 MW, and 28 diagnostic systems (to measure the plasma density, electron temperature, radiation, magnetic field, etc.). Magnet field systems were commissioned firstly for divertor plasma discharges. During the 2nd experimental campaign of HL-3 tokamak, several great progresses have been achieved. Firstly, the successful operation with plasma current larger than 1 MA was achieved under a divertor configuration. Secondly, the advanced divertor concept with two distinct snowflake configurations was realized. It is found that the distribution of ion saturation current and heat flux on bottom plate becomes wide due to magnetic surface expansion, demonstrating the advantage of such configuration in the heat flux mitigation. In addition, using the combination of NBI, ECRH and LHCD, the standard sawtoothing high confinement mode of megampere plasma was firstly accessed on the HL-3. The successful commissioning of HL-3 is beneficial for the initial operation of ITER.

Status and issues of high-temperature and high-pressure water corrosion research of fusion structural materials in Japanese DEMO reactor development

Abstract The activated corrosion product assessments of fusion structural materials are essential to designing components and evaluating workers’ radiation exposure. This paper first gives the R&D status of the high-temperature pressurized water corrosion study of reduced activation ferritic/martensitic steels and chromium–zirconium–copper (CuCrZr) alloys, which are the leading candidate materials of fusion reactor in-vessel components such as breeding blanket and divertor, which are utilized in high-temperature and high-pressure water, and the recent progress of corrosion test apparatus simulating the unique environment of a fusion reactor will also be presented.

Managing the heat: In-Vessel Components.

The Spherical Tokamak for Energy Production (STEP) programme aims to deliver a first-of-a-kind fusion prototype powerplant (SPP). The SPP plasma places extreme heat, particle and structural loads onto the plasma-facing components (PFCs) of the divertor, limiters and inboard and outboard sections of the first wall. The PFCs must manage the heat and particle loads and wider powerplant requirements relating to safety, net power generation, tritium breeding and plant availability. To enable STEP PFC concepts to be identified that satisfy these wide-ranging requirements, an iterative design ('Decide & Iterate') methodology has been used to synchronize a prioritized set of decisions, within the fast-paced, iterative, whole plant concept design schedule. This paper details the 'Decide and Iterate' methodology and explains how it has enabled the identification of the SPP PFC concepts. These include innovative PFC solutions such as a helium-cooled discrete and panel limiter design to increase tritium breeding while providing sufficient coverage and enabling individual limiter replacement; the integration of the outboard first wall with the breeding zone to enhance fuel self-sufficiency and power generation; and the use of heavy water (D2O) within the inboard first wall and divertor PFCs to increase tritium breeding within the outboard breeding zone. This article is part of the theme issue 'Delivering Fusion Energy - The Spherical Tokamak for Energy Production (STEP)'.

Visible core spectroscopy at Wendelstein 7-X.

This paper presents an overview of recent hardware extensions and data analysis developments to the Wendelstein 7-X visible core spectroscopy systems. These include upgrades to prepare the in-vessel components for long-pulse operation, nine additional spectrometers, a new line of sight array for passive spectroscopy, and a coherence imaging charge exchange spectroscopy diagnostic. Progress in data analysis includes ion temperatures and densities from multiple impurity species, a statistical comparison with x-ray crystal spectrometer measurements, neutral density measurements from thermal passive Balmer-alpha emission, and a Bayesian analysis of active hydrogen emission, which is able to infer electron density and main ion temperature profiles.

Preliminary architecture design for human-in-the-loop control of robotic equipment in remote handling tasks: Case study on the NEFERTARI project

In this work, we present a general control architecture for robotic systems dedicated to the remote handling of in-vessel components in fusion machines. This control architecture will be tested for the inspection and maintenance of the first wall components of the RFX-mod2 experiment, in the scope of the New Equipment For the Experimental Research and Technological Advancement for the Rfx Infrastructure (NEFERTARI) project. The architecture is split into low-level and high-level layers. The former is used for the low-level control of the robotics systems and to manage safety; the latter is used for the high-level planning and implements several sub-modules, such as a Virtual Reality (VR) environment for the visualisation of the robot’s digital twin. Moreover, an Application Programming Interface (API) is intended to connect the two layers, and the communication between modules and layers is provided by the ROS2 framework. A typical usage of such a framework involves a human operator who is teleoperating the real robot, data from motor encoders are used as inputs for the dynamic model module. This module is used to compute the forward dynamics in real-time, providing an accurate simulation of the robot (digital twin) in the virtual environment.

Surface temperature measurement from infrared synthetic diagnostic in preparation for ITER operations

The protection of ITER in-vessel components and the plasma-wall interaction studies will be based on a large network of infrared (IR) cameras covering 70% of the tokamak. The surface temperature measurement from IR images remains challenging due to the presence of metallic targets, with changes in surface thermo-radiative properties (emissivity) and the presence of multiple reflections. The paper provides an overview of major progress to improve the interpretation of IR image and to get more reliable surface temperature from IR synthetic diagnostics. The paper presents the latest development of (1) the forward model to include the modelling of the edge localised modes and a new advanced camera that is better adapted to experimental data (2) the inverse model to retrieve the emissivity of the targets and the surface temperature from a neural network trained exclusively from synthetic IR images. Promising results have been obtained both from simulated test images with an estimated emissivity better than 0.05 and a surface temperature better than 10%, and from WEST experimental images of ITER-like wide-angle to filter reflection patterns.

Corrosion phenomena and deposits in ITER Neutral Beam Test Facility primary cooling circuits

The ITER Neutral Beam Test Facility (NBTF) is hosted in Padua and includes two experiments: MITICA, the 1 MeV full-scale prototype of the ITER HNB injector, and SPIDER, the 100 keV full-size ITER Radio Frequency (RF) negative ion source. SPIDER and MITICA experiments are actively cooled by Ultrapure Water (UPW) to electrically insulate in-vessel components that are biased to high voltage levels. Water conductivity is an important monitored parameter to ensure components' insulation by limiting the leakage current of active cooled equipment. A very low conductivity water is especially important in MITICA, where components need to operate up to 1 MVdc, an insulation level beyond the actual industrial standard. Careful selection of suitable materials for any in-vessel (vacuum insulation) and out-vessel component (air insulation) is of utmost importance, as their interaction with water and different environment may affect their chemical and mechanical properties. Additionally, materials selection can severely influence water chemical characteristics: cooling circuits are made of metals (copper, aluminium, steel) and insulating materials (plastic, rubber) when connecting parts at different electric potentials.During the first years of cooling plant exploitation, it was shown that water degrades more quickly than estimated by design. The Primary Circuits (PCs) that showed the most severe water degradation during operation are SPIDER and MITICA power supply ones, respectively called PC01 and PC08. This paper describes the results of specific experimental tests performed on MITICA PC08 to evaluate possible causes of water degradation and detect sources of contaminants that might compromise future experimental campaigns. The cooling circuit was subdivided in different sections and water circulation tests were performed at constant temperature and flowrate. Water samples were collected and Inductively Coupled Plasma Mass Spectrometry (ICP-MS) analyses were performed to quantify the type and amount of metals released. Moreover, components samples collected along their cooling circuits were characterized by Scanning Electron Microscope (SEM) technique to detect undesired contaminants and immersed in UPW to study their corrosion behaviour using metal release tests.

An innovative coil fabrication and assembly process for the VNS tokamak based on in-situ winding

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.

Use of the OSCAR-Fusion V1.4.a code for a preliminary assessment of the ACP contamination within the main ITER water cooling circuit

One of the main objectives of ITER is to produce 500 MW of power from a deuterium-tritium plasma for several seconds. This goal presents two inherent challenges: firstly, in-vessel components will require active cooling to remove the heat coming from the fusion reaction (i.e., mainly fast neutrons and alpha particles). Secondly, the materials exposed to the neutron flux will yield activated corrosion products (ACPs) in all primary cooling circuits of ITER. From a safety point of view, ACPs are one of the contributors to the Occupational Radiation Exposure (ORE), they represent a source of radiological waste and also contribute to the source term for accidental scenarios involving the loss of primary confinement. Therefore, ACPs assessment is key to estimate radiological impact for nuclear workers and the public. ITER nuclear safety engineers adopted OSCAR-Fusion v1.4.a code to assess the ACPs inventory in the Integrated Blanket ELMs and Divertor (IBED) cooling loop. This paper describes the selection of input data, the modelling of the circuits and the operational scenarios used in OSCAR-Fusion calculations. This study also examines the outcomes of such calculations, notably in terms of ACPs inventory, emphasizing the impact on the ORE and highlighting its driving parameters. Additionally, the paper offers recommendations for better ACPs management in the context of the ITER project and in accordance with the ALARA principle.

Control of elongated plasmas in superconductive tokamaks in the absence of in-vessel coils

The roadmap for the commissioning and first operations of superconductive tokamaks envisages the possibility of running discharges with fairly elongated plasmas before the complete installation of the in-vessel components, including vertical stabilization coils, or any other specific sets of coils to be used for the magnetic control of fast transients. In the absence of dedicated actuators, the magnetic control system shall perform the essential fast control actions by using the out-vessel superconductive coils, if needed. These are typically less efficient in reacting to fast transients, due to the shielding effect of the vessel and imply a coupling with other control tasks relying on the same actuators, such as plasma current, position, and shape control. Hence, effective actuator-sharing strategies must be put in place. This paper presents an architecture and a possible control strategy that is able to cope with vertically unstable elongated plasmas subject to fast varying disturbances, in the absence of dedicated in-vessel coils. The architecture exploits a model-based actuator-sharing approach to effectively accomplish the main magnetic control objectives while minimizing the cross-couplings among the various tasks. The effectiveness of the approach is demonstrated by means of nonlinear simulations of realistic JT-60SA scenarios. In particular, an isoflux plasma shape controller is integrated with plasma current control and vertical stabilization. The proposed control approach proves to control vertical displacement events and plasma deformations due to fast variations of poloidal beta with satisfactory performance.

Radiatively tamed divertor thermal loading in resonant magnetic perturbation (RMP)-driven, ELM-crash-suppressed plasmas

Edge-localized-modes (ELMs) suppression by non-axisymmetric resonant-magnetic-perturbation (RMP) provides the way to reach high performance fusion plasmas without a threatening level of transient heat fluxes to the walls of fusion devices. The application of RMP, however, strongly modifies the heat flux pattern onto in-vessel components in contact with the plasma (especially the divertor) leading to local ‘hot spots’. Radiative dissipation by partially ionized species (impurities and deuterium) lowers the heat flux peaks on the walls but has been poorly compatible with such RMP-driven, ELM-crash-suppression. Here, we show how KSTAR has radiatively tamed divertor thermal loading down to more than a factor of 7 in the off-separatrix region without losing ELM-crash-suppression using ITER-like, three-row, RMP configurations, demonstrating its sustainment even in a partially detached plasma in the outer strike point, as required for ITER.

ParaStell: parametric modeling and neutronics support for stellarator fusion power plants

The three-dimensional variation inherent to stellarator geometries and fusion sources motivates three-dimensional modeling to obtain accurate results from computational modeling in support of design and analysis of first wall, blanket, and shield (FWBS) systems. Manually constructing stellarator fusion power plant geometries in computer-aided design (CAD) and defining the corresponding fusion source can be cumbersome and challenging. The open-source parametric modeling toolset ParaStell has been developed to automate construction of such geometries in low-fidelity. Low-fidelity modeling is useful during the conceptual phase of engineering design as a means of rapidly exploring the design space of a given device. The modeling capability of ParaStell includes in-vessel components and magnets, for any given stellarator configuration, using a parametric definition and plasma equilibrium data. Furthermore, the toolset automates the generation of detailed, tetrahedral neutron source definitions and DAGMC geometries for use in neutronics modeling. ParaStell assists rapid design iteration, parametric study, and design optimization of stellarator fusion cores. As a demonstration of the design iteration capability, the effect of the three-dimensional parameter space on tritium breeding and magnet shielding is investigated, using the WISTELL-D configuration as a design basis. Blanket and shield thicknesses are varied in three dimensions, using the space available between the plasma edge and magnet coils as a constraint. The corresponding effects on tritium breeding ratio and magnet heating are tallied using the open-source Monte Carlo particle transport code OpenMC. The inclusion of additional and higher-fidelity modeling capabilities is planned for ParaStell’s future, as well as its implementation in machine-driven optimization.

Multiscale computational study to predict the irradiation-induced change in engineering properties of fusion reactor materials

In this study, we address the impact of irradiation conditions in a tokamak on the engineering properties of materials, leading to potential degradation of in-vessel components over their lifecycle. Our approach involves a predictive model for irradiation-induced damage, employing a multiscale computational framework. This framework integrates various simulation techniques, including Monte Carlo-based neutronics (OpenMC), dislocation dynamics (DD) using MoDELib, and finite element analysis (FEA) with Code_Aster. This integration offers a versatile solver capable of analysing tokamak components exposed to different irradiation doses and temperature conditions.To showcase the utility of this multiscale computational framework, we present a case study focused on tungsten monoblock designs. We assess the failure probabilities of these designs at different stages of their lifecycle. Neutron heating and damage energy values are obtained from OpenMC neutronics simulations. The neutron heating values serve as volumetric heat sources for the FEA thermal simulation. We calculate the displacement per atom (dpa) across the monoblock at various full power days (day 0, day 100, and day 1000) using the damage energy values. The irradiation-induced defect densities, dependent on temperature and dpa, are inputs to DD microstructural simulations performed on the representative volume element (RVE) using MoDELib. This allows us to obtain the yield stress of the material. Subsequently, the thermal fields from the FEA thermal simulation, along with the dpa and temperature-dependent yield stress from the DD simulation, are implemented for FEA mechanical simulations.To evaluate the failure probability of the monoblock designs at different stages of their lifecycle, we conduct an SDC-IC assessment, incorporating a plastic flow localization rule within the current framework. This comprehensive approach provides insights into the thermo-mechanical behaviour of in-vessel components subjected to neutron irradiation, offering a predictive capability for assessing their performance over time.

Key technologies of manufacturing and testing for the ITER blanket shield block

The SB is one of the main in-vessel components manufactured with SS316L(N)-IG to play a major role in neutron shielding, to support the FW panels and to supply the FW panel with cooling water. Cooling channels are located inside of the SB to remove the neutron heat deposition while keeping the structure temperature to acceptable levels. The water coolant is routed first through the plasma-facing first wall panel which is attached to the SB and then through the SB. A number of deep slits are introduced into the SB to disrupt eddy current loops and reduce the Electro-Magnetic (EM) loads on the support system and VV. Water headers are machined on the side of the module with 7–10 mm welded cover plates. A Procurement Arrangements (PAs) for the manufacturing and testing of the SBs have been signed with Korea (50 %) and China (50 %).As part of the PA in Korean Domestic Agency (KODA), Process Qualification (PQ) to ensure the manufacturing and testing ability of the SB was completed successfully through manufacture of a Full Scale Prototype (FSP) as an essential step before starting series production. This paper reports key technologies, such as machining, welding, NDE and hot helium leak testing, of manufacturing and testing for the ITER SB series production in Korea toward successful completion of the missions.

The integrated engineering design concept of the upper limiter within the EU-DEMO LIMITER system

The EU-DEMO first wall protection relies on a system of limiters. Although they are primarily designed for facing the energy released by a limited plasma during transients, their design should safely withstand a combination of loads relevant for in-vessel components (IVCs) during steady-state operation. They are not meant to breed tritium, nor to provide plasma stability. However, sitting in place of blanket portions, they should ensure an adequate shielding function to vacuum vessel and magnets while withstanding both their dead weight and the electro-mechanical loads arising from the interaction between current induced in the conductive structure and magnetic field. During plasma disruptions they will be subjected to halo currents flowing from/to the plasma and the grounded structures, whose effects must be added to the eddy current ones. Disruption-induced electro-mechanical loads are hence IVC design-driving, despite the uncertainties in both eddy and halo currents’ magnitude and distribution, which depend on IVC design, electrical connectivity, plasma temperature and halo width.The integrated design of the limiter is made of two actively water-cooled sub-components: the Plasma-Facing Wall (PFW) directly exposed to the plasma, and the Shielding Block (SB) devoted to hold the PFW while providing neutronic shielding. The PFW design is driven by disruptive heat loads. Disruption-induced electro-magnetic loads are instead SB design drivers, meaning that the design details (i.e. geometry, electrical connections, attachments) affect the loads acting on it, which, in turn, are affected by the mechanical response of the structure.The present paper describes the design workflow and assessment of the Upper Limiter (UL), resulting from a close and iterative synergy among different fields. Built on static-structural and energy balance hand calculations based on, respectively, preliminary electro-magnetic and neutronic loads, the UL integrated design performance has then been verified against electro-magnetic, neutronic, thermal-hydraulic and structural assessment under the above-mentioned load combination. The outcome will be taken as reference for future limiter engineering designs.