Fusion Engineering and Design最新文献

{"title":"H-Mode density limit disruption avoidance in ASDEX Upgrade, TCV and JET","authors":"B. Sieglin ,&nbsp;M. Maraschek ,&nbsp;A. Gude ,&nbsp;F. Klossek ,&nbsp;F. Felici ,&nbsp;M. Bernert ,&nbsp;O. Kudlacek ,&nbsp;A. Pau ,&nbsp;L. Piron ,&nbsp;M. Lennholm ,&nbsp;K. Kirov ,&nbsp;S. Hongjuan ,&nbsp;A. Parrott ,&nbsp;P. Fox ,&nbsp;O. Sauter ,&nbsp;C. Stuart ,&nbsp;C. Ashe ,&nbsp;A. Huber ,&nbsp;W. Treutterer ,&nbsp;ASDEX Upgrade Team ,&nbsp;TCV Team ,&nbsp;JET contributors ,&nbsp;EUROfusion WPTE Team","doi":"10.1016/j.fusengdes.2025.114961","DOIUrl":"10.1016/j.fusengdes.2025.114961","url":null,"abstract":"<div><div>In recent years a strong effort has been made to investigate disruption avoidance schemes in order to aid the development of integrated operational scenarios for ITER. Within the EUROfusion programme the disruptive H-Mode density limit (HDL) has been studied on ASDEX Upgrade, TCV and JET. This paper presents the results from these studies in the view of the plasma control system (PCS).</div><div>The successful migration of disruption avoidance schemes based on direct measurements as well as state space based HDL observers across the devices has been demonstrated. HDL disruption avoidance using continuous control is now routinely available at TCV and ASDEX Upgrade, using both observer types. Exception handling based disruption avoidance has been utilized on all three devices. All three control systems allowed the implementation of the desired schemes with minimal adaptions, implementing the equivalent detection algorithms. Using the same experimental methodology it is confirmed that the XPR/MARFE movement, which is preceding the disruptive HDL, is similar on all three devices. The main difference lies in the associated times scales which increase with machine size. This would be beneficial for large devices, such as ITER and DEMO, since it would give the control system more time to apply the appropriate disruption avoidance action. As actuator auxiliary heating, gas flow and plasma shaping have been successfully demonstrated.</div><div>This work gives input for the requirements of future tokamak control systems which should enable the necessary disruption avoidance schemes. All disruption avoidance schemes developed during the studies can be implemented in the ITER PCS with its currently foreseen capabilities.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114961"},"PeriodicalIF":1.9,"publicationDate":"2025-03-18","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143641723","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Research on gas target neutralization processes for negative ion beam","authors":"Haohua Xia,&nbsp;Chen Zuo,&nbsp;Hanyang Liu,&nbsp;Ming Ding,&nbsp;Junbin Zhao","doi":"10.1016/j.fusengdes.2025.114931","DOIUrl":"10.1016/j.fusengdes.2025.114931","url":null,"abstract":"<div><div>Neutral Beam Injection (NBI) is one of the most important heating methods in magnetic confinement fusion. It needs to convert high-energy charged particles into neutral particles as efficiently as possible, and gas neutralization is currently widely used due to its simplicity and maturity. The target gas density distribution is a critical parameter for gas neutralizers, which affects the neutralization efficiency and beam transmission. This paper focuses on the processes of negative ion beam neutralization, based on a coupled simulation model of gas density distribution and beam transmission. The influence of gas injection methods and neutralizer length on the evolution of beam components were analyzed in detail. The results indicate that by maintaining the gas target thickness at the optimal value, the gas injection methods significantly affect the distribution of target gas density and the symmetry of beam particles at the beam transmission section. Increasing the length of the neutralizer can significantly decrease the required gas inlet rate, which can further reduce the vacuum system load and re-ionization of beam particles in the vacuum chambers. The influence of beam divergence on the power load distribution and transmission efficiency were also analyzed, providing references for the design of gas neutralizers in NBI systems.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114931"},"PeriodicalIF":1.9,"publicationDate":"2025-03-17","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143636631","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Virtual simulator for the ITER Agile Robot Transporter with an inverse kinematic algorithm considering joint limits and collision avoidance","authors":"Chang-Hwan Choi ,&nbsp;Jonas Kynde ,&nbsp;Soumik Sarkar ,&nbsp;David Hamilton ,&nbsp;Raphael Hery ,&nbsp;Jim Palmer","doi":"10.1016/j.fusengdes.2025.114927","DOIUrl":"10.1016/j.fusengdes.2025.114927","url":null,"abstract":"<div><div>Remote Handling (RH) systems are required to maintain the in-vessel components inside the ITER Vacuum Vessel (VV) because of high radiation. The Agile Robot Transporter (ART) is under development to cope with the in-vessel maintenance needs. This paper describes a virtual simulator of the Transporter to verify its kinematics such as joint ranges and body lengths whether it is suitable to carry out the blanket first wall maintenance. Independent joint control and sequential joint control algorithms are implemented, which are used to simulate the deployment sequence of the Transporter. End-effector control, null space control, automatic trajectory generation and control algorithms are implemented by using inverse kinematic algorithms, which are used to produce various postures of the Transporter to access the blanket first walls. The control algorithms consider the physical constraints such as the joint limits, internal collision avoidance among the Transporter’s bodies, and external collision avoidance between the Transporter’s bodies and the blanket first walls. The simulation results demonstrate that the Transporter can be deployed from the port into the VV and it can access all blanket first walls while respecting the physical constraints.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114927"},"PeriodicalIF":1.9,"publicationDate":"2025-03-15","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143628678","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Dose assessment of gaseous tritium release from deuterium-tritium fusion devices based on steady-state plume model","authors":"Weijie Cui ,&nbsp;Zaixin Li ,&nbsp;Qixiang Cao ,&nbsp;Bo Cao","doi":"10.1016/j.fusengdes.2025.114965","DOIUrl":"10.1016/j.fusengdes.2025.114965","url":null,"abstract":"<div><div>The deuterium-tritium fusion represents a promising avenue for achieving commercial controlled nuclear fusion, but the usage of radioactive tritium poses significant environmental and radiological risks to personnel. To evaluate the radiological impact on nearby staff and the public from atmospheric tritium released during the steady-state operation of fusion devices, this study developed a dose assessment code based on the steady-state Gaussian plume model. Using tritium water vapor release rates from the ITER safety report as the source term, we calculated the tritium doses absorbed via inhalation and skin contact for various age groups, utilizing 2023 ECMWF reanalysis meteorological data for an inland basin site in China, a hypothetical site in Daya Bay, and the ITER site. The results indicate that the maximum public dose near the inland basin site is 0.030 mSv/yr, near the Daya Bay site is 0.015 mSv/yr, and near the ITER site is 0.018 mSv/yr, all of which are well below the ITER project's limit of 0.1 mSv/yr. Thus, fusion devices are radiologically safe for the surrounding personnel under normal operation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114965"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621483","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Implications of the measured BRDF for the ITER divertor impurity monitor","authors":"Hiroki Natsume ,&nbsp;Kunpei Nojiri ,&nbsp;Shin Kajita ,&nbsp;Hirohiko Tanaka ,&nbsp;Tomohiro Yokozuka ,&nbsp;Yoshihiko Nunoya","doi":"10.1016/j.fusengdes.2025.114926","DOIUrl":"10.1016/j.fusengdes.2025.114926","url":null,"abstract":"<div><div>This study employed a goniophotometer to investigate the three-dimensional bidirectional reflectance distribution function (BRDF) of tungsten, a material utilized in the ITER divertor. The measured BRDFs were incorporated into ray-tracing simulations to assess the impact of reflected light on the divertor impurity monitor (DIM), a spectroscopic measurement for ITER. The results demonstrate that surface conditions, which are expected to be modified by the plasma-wall interaction, exert a considerable influence on the characteristics of reflected light. In comparison to mirror-polished surfaces, the sputtered and recrystallized surfaces exhibited broader reflection patterns. In addition, the presence of a fuzzy surface resulted in a notable reduction in reflected light, including multiple reflections. Furthermore, this study investigated the potential for errors to arise from the use of optical reflection models with a small number of free parameters. It was found that discrepancies in the BRDF at shallow angles result in errors in the reflected light observed in the DIM. However, these discrepancies became relatively small in DIM measurements when the surface exhibited low specular reflection, such as sputtered or fuzzy surfaces. These results indicate that while some discrepancy exists between the predicted and actual BRDFs, the impact on DIM diagnostics may not be as significant as the error in the BRDF itself.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114926"},"PeriodicalIF":1.9,"publicationDate":"2025-03-14","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143621482","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Temperature dependence of sputtering yields of steels with various W content for plasma facing applications","authors":"S. Möller ,&nbsp;M. Reinhart ,&nbsp;B. Kuhn ,&nbsp;A. Kreter","doi":"10.1016/j.fusengdes.2025.114950","DOIUrl":"10.1016/j.fusengdes.2025.114950","url":null,"abstract":"<div><div>Materials are a key issue in any nuclear fusion reactor. For plasma-facing materials irradiation and sputtering limit their lifetime in a fusion power reactor. Steels allow for lower costs, reduced activation, and proven long irradiation lifetimes compared to tungsten at the expense of a lower sputtering lifetime.</div><div>Several studies investigated the connected effects showing W-fuzz like surface morphology and a W enrichment through preferential sputtering, opening up a potentially significant increase of this sputtering lifetime of steels.</div><div>The role of the W content in the steel is an open question. The presented exposure of 3 steels with W content from 0.5 to 2 at. % in deuterium plasma at 693–843 K sample temperature aims at resolving this. The experiments combine in-situ infrared imaging and passive spectroscopy with ex-situ ion-beam analysis.</div><div>At the given conditions, a flux density independent exponential W enrichment within 70 s time-constant followed by a 1400 s time-constant nano-structure formation is observed. The overall effect is a reduction of the Fe sputtering yield by a factor 4.4, 3.0, and 2.5 for 2 %, 1.1 %, 0.5 % W steels, respectively. This factor reduces slightly with temperature by up to 20 % towards higher temperatures.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114950"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143609577","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Design, development and manufacture of the ITER Torus and cryostat cryopumps","authors":"Robert Pearce ,&nbsp;J. Buckerfield ,&nbsp;A. Iannetti ,&nbsp;F. Chitu ,&nbsp;A. Cobalt ,&nbsp;S. Nasluzov ,&nbsp;E. Quinn ,&nbsp;F. Canadell ,&nbsp;M. Dremel ,&nbsp;L. Mora ,&nbsp;G. Moner Pallas ,&nbsp;J. Zeutschel","doi":"10.1016/j.fusengdes.2025.114939","DOIUrl":"10.1016/j.fusengdes.2025.114939","url":null,"abstract":"<div><div>Eight large bespoke cryo-sorption pumps have been successfully manufactured for the pumping of ITER's divertor and cryostat. These pumps were a European in kind supply to the ITER project. They each have a nominal pumping speed of ∼100m³/s, weigh ∼8 tonnes each and will operate in synchronization to provide continuous high vacuum pumping of the vacuum vessel and cryostat in many different operational modes.</div><div>The cryopumps are some of the most complex components on the ITER tokamak, as vacuum/cryogenic components operating down to 4.5K and up to 470K, as precision mechanical components with moving parts on account of their integral 800mm all metal isolation valves and being nuclear components providing first confinement for tritium and nuclear shielding.</div><div>The time between the start of development of the pumps to the production pump deliveries is &gt; 20 years involving many parties and dedicated efforts. Both a half size pump and a pre-production pump were built and tested as part of the development. Qualification and validation have been essential to ensure the pumps meet their performance requirements, nuclear safety functions and are built with a quality to function over tens of years.</div><div>This paper summarizes the experience gained and lessons learned from the design, validation and manufacturing process.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114939"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143609578","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Design specifications for the ion cyclotron heating amplifier system in the Divertor Tokamak Test facility","authors":"C. Salvia ,&nbsp;P. Bettini ,&nbsp;A. Cardinali ,&nbsp;S. Ceccuzzi ,&nbsp;S. Greco ,&nbsp;D. Van Eester ,&nbsp;F. Zanon","doi":"10.1016/j.fusengdes.2025.114916","DOIUrl":"10.1016/j.fusengdes.2025.114916","url":null,"abstract":"<div><div>This paper provides an overview of the basis of design of the high-power amplification for the Ion Cyclotron Heating system of the Divertor Tokamak Test facility. Notably, this system will be entirely based, for the first time, on solid-state technology, marking a significant milestone in the development of advanced fusion energy systems. The amplifier operates within a frequency range of 60–90 MHz, delivering MW-level power and supporting long pulse durations. The design specifications of amplifiers under different operating conditions, such as plasma experiments, wall cleaning, and conditioning are provided. By explaining the reasons behind the adoption of solid-state amplifiers, the paper highlights the role of solid-state technology in advancing fusion research.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114916"},"PeriodicalIF":1.9,"publicationDate":"2025-03-13","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143609579","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Using programming by demonstration tools for DEMO maintenance: A framework proposal with an experimental validation","authors":"O. David ,&nbsp;V. Dhédin ,&nbsp;J. Dumora ,&nbsp;V. Padois ,&nbsp;A. Rebbouh ,&nbsp;F. Milella","doi":"10.1016/j.fusengdes.2025.114935","DOIUrl":"10.1016/j.fusengdes.2025.114935","url":null,"abstract":"<div><div>The level of radiation expected in the DEMO fusion reactor facility will make it impossible to send operators during maintenance phases of the reactor as well as of a large number of ancillary components. During each scheduled shutdown phase, specifically designed robotized machines will replace consumable components in the course of complex and carefully studied procedures. However, accurate predictions for a large volume of tasks such as inspection or replacement of faulty components, are very difficult. These tasks will therefore require more versatile tools (e.g. manipulator with carrier). The resulting sum of potential errors (manufacturing, positioning…) makes it complex to carry out these tasks in a fully automated mode. Man-in-the-loop teleoperation techniques are therefore being considered, as they provide fault tolerance and adaptability. Previous studies have shown that the generation of operator assistance by model-based control tools does not only provides comfort when executing tasks, but also significantly improves their execution quality and allows hybrid modes switching between man in the loop, automated or semi-automated control modes. However, design of such models remains complex. Inspired by “programming by demonstration” (PbD) tools used in robotics, the feasibility of a framework creating the model directly during the execution phase is investigated, thanks to measurements taken locally. The model is built-up on-site, today with help of probes and in the future with 3D scanners. Then, the model is used to generate operator assistance. After reviewing the literature on the subject, we detail the particularities of our approach focusing on the interface level between a tool and its target and we explain why it is relevant in contexts such as fusion facilities. Then an experimental validation in the form of a pipe inspection with a camera mounted on a manipulator arm is presented.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114935"},"PeriodicalIF":1.9,"publicationDate":"2025-03-11","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592955","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}

{"title":"Characterization of a new 4He scintillator detector prototype for the ITER Radial Neutron Camera","authors":"S. Cesaroni ,&nbsp;D. Marocco ,&nbsp;V. Anagnostopoulou ,&nbsp;F. Belli ,&nbsp;A. Colangeli ,&nbsp;N. Fonnesu ,&nbsp;S. Loreti ,&nbsp;F. Moro ,&nbsp;G. Pagano ,&nbsp;E. Pirovano ,&nbsp;F. Pompili ,&nbsp;M. Pontesilli ,&nbsp;A. Zimbal ,&nbsp;B. Esposito","doi":"10.1016/j.fusengdes.2025.114948","DOIUrl":"10.1016/j.fusengdes.2025.114948","url":null,"abstract":"<div><div>In ITER, the reconstruction of the neutron emissivity profile along radial Lines-Of-Sight (LOS) is performed with the Radial Neutron Camera (RNC) diagnostic system. Neutrons produced in the plasma core through D(d, n)³He and T(d, n)⁴He reactions are measured by the RNC Ex-Port subsystem using three different types of detectors located within modules positioned at the end of each LOS collimator: a plastic scintillator, a ⁴He gas scintillator and a single-crystal diamond matrix.</div><div>Considering plasma scenarios foreseen in ITER, measurements during full-power operations rely on the ⁴He scintillator detector. A new prototype of this device, specifically designed for the ITER RNC system, has been developed and experimentally characterized. This work presents the results of the measurements performed at the Physikalisch-Technische Bundesanstalt (PTB) Ion Accelerator Facility (PIAF) in Braunschweig, focusing on the detector's response to different neutron energies, and at ENEA Frascati Neutron Generator (FNG), where its performance under high neutron fluxes was studied. The new ⁴He scintillator detector prototype has proven to be suitable for its installation in the ITER RNC Ex-Port detector modules, and capable of reliable operation during the full-power scenarios of ITER operation.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114948"},"PeriodicalIF":1.9,"publicationDate":"2025-03-10","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143592393","PeriodicalName":null,"FirstCategoryId":null,"ListUrlMain":null,"RegionNum":3,"RegionCategory":"工程技术","ArticlePicture":[],"TitleCN":null,"AbstractTextCN":null,"PMCID":"OA","EPubDate":null,"PubModel":null,"JCR":null,"JCRName":null,"Score":null,"Total":0}