Fusion Engineering and Design最新文献

Mechanism of nanocracks formation in tungsten and tungsten-nickel-iron alloy under helium ion irradiation

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-09 DOI: 10.1016/j.fusengdes.2025.115037

Haiyu Xue , Te Zhu , Xiangming Xie , Qigui Yang , Peng Zhang , Baoyi Wang , Pinjing Cheng , Xingzhong Cao , Haibiao Wu

{"title":"Mechanism of nanocracks formation in tungsten and tungsten-nickel-iron alloy under helium ion irradiation","authors":"Haiyu Xue , Te Zhu , Xiangming Xie , Qigui Yang , Peng Zhang , Baoyi Wang , Pinjing Cheng , Xingzhong Cao , Haibiao Wu","doi":"10.1016/j.fusengdes.2025.115037","DOIUrl":"10.1016/j.fusengdes.2025.115037","url":null,"abstract":"<div><div>Plasma-facing materials used in fusion energy facilities, such as pure tungsten (W) and tungsten-nickel-iron (W-Ni-Fe) alloys, must withstand helium irradiation, which can lead to nanocrack that eventually develop into surface blistering and large-size crack. The mechanisms behind this phenomenon remain elusive, highlighting the need for a detailed investigation. This study employed positron annihilation spectroscopy (PAS) and transmission electron microscopy (TEM) to analyze the development of nanocracks. The results elucidate that helium irradiation initially engenders vacancy-type defects within the material matrix. As the irradiation dose escalates, implanted helium atoms are captured by these vacancies, leading to the formation of helium-vacancy complexes that subsequently expand into larger aggregates and eventually evolve into helium bubbles. The coalescence of these bubbles results in the generation of high-pressure helium entities, which precipitate nanocrack formation once the internal pressure surpasses the structural limits of the material, ultimately culminating in surface blistering.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115037"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799761","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}

引用次数: 0

Scoping studies on ITER tungsten first wall and borated water in Vacuum Vessel

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-09 DOI: 10.1016/j.fusengdes.2025.115030

Alberto Bittesnich , Davide Laghi , Marco Fabbri , Alfredo Portone

{"title":"Scoping studies on ITER tungsten first wall and borated water in Vacuum Vessel","authors":"Alberto Bittesnich , Davide Laghi , Marco Fabbri , Alfredo Portone","doi":"10.1016/j.fusengdes.2025.115030","DOIUrl":"10.1016/j.fusengdes.2025.115030","url":null,"abstract":"<div><div>The 2023 ITER rebaselining includes the replacement of beryllium (Be) First Wall (FW) with tungsten (W). Furthermore, at the March 2023 Technical Coordination Meeting (TCM) the investigation of borated water in Vacuum Vessel (VV) – Primary Heat Transfer Systems (VV-PHTS) has been requested. In this context, scoping studies were made by using ITER C-Model, to assess the impact of these design changes on nuclear heating of different Plant Breakdown Systems (PBS). The pre-and post-processing of the calculations has been carried out using F4Enix, a new open-source Python package for Monte Carlo simulations input and output files parsing developed at Fusion for Energy (F4E). This paper wants to show and provide physical explanation of the results of the scoping studies, with particular attention to radiation shielding of superconducting magnets. The analysis evidenced that a thickness of at least 3 mm of W is needed to reach the shielding performance of 10 mm of Be, and that borated water provides relevant nuclear heating reduction mainly in the Toroidal Field Coils (TFCs), but its usability is limited by different technical reasons.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"216 ","pages":"Article 115030"},"PeriodicalIF":1.9,"publicationDate":"2025-04-09","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143799762","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}

引用次数: 0

Comparative study of PVD techniques for tungsten deposition onto graphite substrate for fusion application

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-09 DOI: 10.1016/j.fusengdes.2025.115016

Alžběta Endrychová , Slavomír Entler , Vojtěch Smolík , Tomáš Polcar , Jan Walter , Jan Podaný

引用次数: 0

Numerical investigation of the DTT cryopump performance via 3D Direct Simulation Monte Carlo modeling

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-04 DOI: 10.1016/j.fusengdes.2025.115021

C. Tantos , H. Strobel , V. Hauer , C. Day , T. Giegerich , P. Innocente

{"title":"Numerical investigation of the DTT cryopump performance via 3D Direct Simulation Monte Carlo modeling","authors":"C. Tantos , H. Strobel , V. Hauer , C. Day , T. Giegerich , P. Innocente","doi":"10.1016/j.fusengdes.2025.115021","DOIUrl":"10.1016/j.fusengdes.2025.115021","url":null,"abstract":"<div><div>DTT (Divertor Tokamak Test Facility) is a new facility, currently under construction in Frascati, Italy. The goal is to provide a basis to perform various scaled experiments for testing different magnetic configurations and alternative solutions for the power exhaust system of DEMO. The DTT pumping system design uses cryopumps as the primary pumping solution and up to 10 pumping openings. The cryopump system was developed at the Karlsruhe Institute of Technology (KIT). In the present work, the pumping capabilities of the DTT cryopump are estimated by performing a 3D numerical investigation of the neutral gas dynamics in the pumping duct of DTT including the entire complex cryopump geometry. The investigation is based on the Direct Simulation Monte Carlo (DSMC) method, which allows for a precise description of the neutral gas dynamics over the entire range of the gas collisionality. The values of the pumping probability for deuterium and neon were determined for two scenarios: open and closed divertor toroidal gaps. For open gaps, the probabilities were obtained as 0.4 for deuterium and 0.62 for neon, with a slight increase in both values observed when the gaps were closed. The results suggest that achieving the target of ten ports for deuterium pumping seems feasible, while for neon, partial elimination of toroidal leakages may be required. The importance of these simulations lies in the fact that, given the imposed simplifications, they delineate the existing pumping capabilities of the DTT particle exhaust and can act as a guide regarding the pumping capabilities among various plasma configurations. Moreover, this work demonstrates the level of geometric complexity that can be adopted in numerical modeling and highlights the effort needed to determine the values of the pumping probability, which are dependent on the chosen pumping technology.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115021"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768180","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}

引用次数: 0

On the use of LSTM-based estimation components for tokamak gas actuator control

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-04 DOI: 10.1016/j.fusengdes.2025.114932

Hudson Baker, Lucas Brown, Adam Parrott

引用次数: 0

Integrity assessment of Tokamak-type fusion reactor First Wall and Blanket structures

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-04 DOI: 10.1016/j.fusengdes.2025.114995

Mohammad Alabdullah , Nasr Ghoniem

{"title":"Integrity assessment of Tokamak-type fusion reactor First Wall and Blanket structures","authors":"Mohammad Alabdullah , Nasr Ghoniem","doi":"10.1016/j.fusengdes.2025.114995","DOIUrl":"10.1016/j.fusengdes.2025.114995","url":null,"abstract":"<div><div>This study presents a nonlinear structural analysis of the First Wall & Blanket (FW/B) of the Fusion Nuclear Science Facility (FNSF), aimed at identifying potential failure modes and estimating the component’s lifetime. The analysis focuses on the accumulation of plastic strain damage and applies fracture mechanics principles to assess the effects of radiation on mechanical properties. The study uses a database for reduced activation ferritic/martensitic steel F82H, with particular emphasis on radiation-induced changes in its mechanical properties. Safety factors for various failure modes are evaluated both at the beginning of life and after neutron irradiation under normal and off-normal operating conditions. The results indicate that the most critical failure modes are flow localization, thermal creep damage, and fast fracture due to radiation-induced reduction in fracture toughness. Based on current empirical prediction methods, we estimate that the FW/B structure will likely survive for approximately one year of operation under conditions of 1.2 MW/m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span> neutron wall loading and 0.25 MW/m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span> surface plasma thermal flux. These findings suggest that the effects of very high fluence (100–200 dpa) may be less significant than expected, with thermal creep damage and rapid fracture toughness degradation being the primary factors controlling failure. Therefore, there is a critical need to improve creep strength at temperatures above 500 °C and to mitigate fracture toughness degradation with increasing neutron fluence. Furthermore, the development of compact fusion energy sources with neutron wall loads greater than 5 MW/m<span><math><msup><mrow></mrow><mrow><mn>2</mn></mrow></msup></math></span> may require new classes of structural materials to ensure their economic viability.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 114995"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768100","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}

引用次数: 0

Development and optimization of a new control interface for the ASDEX upgrade high current converters

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-04 DOI: 10.1016/j.fusengdes.2025.115006

Friedhelm Albers, Nils Arden, Manuel Krause, Bernhard Sieglin, the ASDEX Upgrade Team

{"title":"Development and optimization of a new control interface for the ASDEX upgrade high current converters","authors":"Friedhelm Albers, Nils Arden, Manuel Krause, Bernhard Sieglin, the ASDEX Upgrade Team","doi":"10.1016/j.fusengdes.2025.115006","DOIUrl":"10.1016/j.fusengdes.2025.115006","url":null,"abstract":"<div><div>The original interface between the ASDEX Upgrade (AUG) control system and the power converters of the high current group uses multicore signal cables to transmit command, status and control loop signals. Depending on the configuration of the AUG coils and the converters, these signal cables were manually arranged via connectors on a patch panel.</div><div>The redesigned control interface will consist of two separate Ethernet links, one for command and status signals and one for control loop signals. With the digital interface, there are no more contacting problems or crosstalk of signals.</div><div>A new programmable logic controller (PLC) was installed as the central point of the network to coordinate the exchange of command and status signals between AUG and the power converters. In order to achieve short response times in the network, additional PLC hardware was replaced and the real-time capable network protocol PROFINET IO was selected.</div><div>The second Ethernet connection, which uses the User Datagram Protocol (UDP), is currently under development. Therefore, an industrial PC was installed to receive the data from AUG and forward it to the corresponding converter feedback control system. While the newer converters with digital feedback control can be connected directly via Ethernet, the older converters with analog feedback control require additional digital-to-analog converters.</div><div>This paper presents the concept, implementation and optimization of the command and status interface. Furthermore, the design of the prototype of the control loop interface for the converters with analog feedback control as well as initial test results are shown.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115006"},"PeriodicalIF":1.9,"publicationDate":"2025-04-04","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143768179","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}

引用次数: 0

Radiation effects in optical fibres measured during JET DT plasma experiments

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-03 DOI: 10.1016/j.fusengdes.2025.115011

M Maslov , D Kos , Z Ghani , A Milocco , Z Huang , B Gray , I Ipatova , the JET contributors and the EUROfusion Tokamak Exploitation Team

{"title":"Radiation effects in optical fibres measured during JET DT plasma experiments","authors":"M Maslov , D Kos , Z Ghani , A Milocco , Z Huang , B Gray , I Ipatova , the JET contributors and the EUROfusion Tokamak Exploitation Team","doi":"10.1016/j.fusengdes.2025.115011","DOIUrl":"10.1016/j.fusengdes.2025.115011","url":null,"abstract":"<div><div>In 2021 and 2023 JET has completed experiments with D/T plasma mixtures where a significant amount of 14 MeV neutrons were released. To collect experimental data on the effect of fusion relevant radiation environment on optical fibres, a 60 metres long pure fused silica fibre has been coiled and placed in the vicinity of the JET machine to accumulate neutron and gamma dose during these experiments. Existing high-OH content radiation resistant fibres were used to connect the sample fibre with the lab space outside the neutron shield, so that the change in optical properties could be measured remotely in real time during and between JET plasma pulses. Significant amount of data was collected over that period. Radiation-induced emission (RIE) was observed over the whole measured spectrum but was dominant in the blue range. Prompt radiation-induced attenuation (RIA) is also dominant the blue range of spectrum and very weak in the near-infrared. The majority of the short-term radiation-induced damage recovers within 1 h after the irradiation, but a fraction of the transmission loss remained permanent.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115011"},"PeriodicalIF":1.9,"publicationDate":"2025-04-03","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143761162","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}

引用次数: 0

Construction of 3D MHD pressure drop correlation and flow characterization in the contraction region of a fusion blanket manifold

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-02 DOI: 10.1016/j.fusengdes.2025.115005

Yuchen Jiang, Sergey Smolentsev

{"title":"Construction of 3D MHD pressure drop correlation and flow characterization in the contraction region of a fusion blanket manifold","authors":"Yuchen Jiang, Sergey Smolentsev","doi":"10.1016/j.fusengdes.2025.115005","DOIUrl":"10.1016/j.fusengdes.2025.115005","url":null,"abstract":"<div><div>Inlet and outlet manifolds are typical components of liquid metal (LM) blanket designs of a fusion power reactor to be used to distribute the LM flow into breeding channels and collect it at the exit of the blanket. High pressure loss in the magnetohydrodynamic (MHD) flows featuring abrupt geometrical changes is one of the main feasibility issues of such designs. Recently, optimization studies were conducted to construct 3D MHD pressure drop correlations for a LM flow in an electrically insulating manifold with gradual expansion. Here, the 3D computational approach developed in that study is applied to the outlet manifold featuring gradual contraction. A systematic analysis was performed with a total number of 135 flow cases computed with COMSOL Multiphysics for Hartmann numbers 1000 < <em>Ha</em> < 10,000, Reynolds numbers 100 < <em>Re</em> < 12,000, and contraction angles 45° < <em>θ</em> < 75° for a fixed contraction ratio of 4. The effects of <em>Ha, Re</em> and <em>θ</em> on the flow recirculation, development length and the total pressure drop were carefully examined. A linear regression analysis was used to determine the power rule of pressure drop coefficient <em>k</em> related to <em>Ha</em> and <em>Re</em>, demonstrating a good match with the Ludford layer theory. Eventually, a correlation for the 3D MHD pressure drop coefficient was constructed as a function of <em>Ha, Re</em> and <em>θ</em>. The results were compared against the inlet manifold. It was found that the flow in the inlet manifold exhibits larger recirculation zones. In the investigated range of <em>Ha, Re</em> and <em>θ</em>, the pressure drop coefficient <em>k</em> of the LM MHD flow in the gradual contraction is only slightly lower (< 8 %) than that in the gradual expansion.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115005"},"PeriodicalIF":1.9,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143747791","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}

引用次数: 0

Analysis of load energy feedback in pulsed coils power supply for the DIII-D tokamak

IF 1.9 3区工程技术

Fusion Engineering and Design Pub Date : 2025-04-02 DOI: 10.1016/j.fusengdes.2025.115003

Qian Jiang , Ge Gao , Shusheng Wang , Zongjie Fu

{"title":"Analysis of load energy feedback in pulsed coils power supply for the DIII-D tokamak","authors":"Qian Jiang , Ge Gao , Shusheng Wang , Zongjie Fu","doi":"10.1016/j.fusengdes.2025.115003","DOIUrl":"10.1016/j.fusengdes.2025.115003","url":null,"abstract":"<div><div>In response to the issue of power device damage in pulsed power supply system of DIII-D Tokamak device caused by the rise in DC bus voltage due to energy feedback from the superconducting load coil during physical experiments. This paper focused on the mechanism analysis of energy feedback from strongly magnetically coupled load coils. Initially, theoretical calculations are performed to assess the energy fed back from the load side to the DC side of the power supply during normal pulsed operation in the second or fourth quadrant. To ensure the continuous operation of normal pulses, an energy absorption circuit for the DC bus is designed, and the parameters of this circuit are carefully calculated. The proposed absorption circuit is capable of absorbing 100–150 kJ of energy feedback from the load coils during a 10s/10 min pulsed operation mode. Furthermore, a thyristor bypass circuit is designed to short-circuit the load coil within 3–5 ms in the event of an unexpected shutdown or termination of the experiment, thereby protecting the electrical equipment. Finally, the proposed solution is validated through simulation to ensure the power supply system can reliably handle the energy feedback from the load.</div></div>","PeriodicalId":55133,"journal":{"name":"Fusion Engineering and Design","volume":"215 ","pages":"Article 115003"},"PeriodicalIF":1.9,"publicationDate":"2025-04-02","publicationTypes":"Journal Article","fieldsOfStudy":null,"isOpenAccess":false,"openAccessPdf":"","citationCount":null,"resultStr":null,"platform":"Semanticscholar","paperid":"143747884","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}

引用次数: 0