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Verification of shielding calculation capability of cosRMC with SINBAD fusion benchmarks Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-23 Rui Che, Shichang Liu, Zhuo Tian, Yixue Chen
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,
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Preliminary accident analyses of in-vessel LOCA for the chinese fusion test reactor HCCB blanket concept Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-22 Bing Zhou, Bo Hu, Yanling Wang, Xiaoyu Wang, Long Zhang, Xinghua Wu
For the helium-cooled ceramic breeder (HCCB) blanket concept improved in 2018 and the primary heat transfer system following China Fusion Engineering Test Reactor (CFETR) in 2019, an in-vessel loss of coolant accident (LOCA) has been investigated with the assumption of a break of the first wall (FW) coolant channels and a double-ended guillotine break of a helium cooling system (HCS) pipe in vacuum
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Fabrication and load test of ITER assembly tools for lifting heavy components Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-20 Jinho Bae, Kyoungo Nam, Min-Su Ha, Yujin Kim, Jingyun Kim
Korea Domestic Agency (ITER Korea, KODA) has procured 44 kind's assembly tools for ITER project. The assembly tools are classified by the steel structure, lifting table and lifting accessory. ‘PF 1, 6 and Central Solenoid Lifting Frame’ (CS Lifting Frame) and ‘Sector Lifting Tool’ (SLT) are the representative assembly tool classified the lifting accessory. The assembly tools handle the heavy component
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Design progress of DTT divertor fixation system Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-20 D. Marzullo, N. Massanova, F. Giorgetti, B. Riccardi, G. De Sano, S. Roccella
One of the main challenges for the construction of DEMO, the first fusion demonstration reactor, is the power exhaust issue. To deal with it, EUROfusion has undertaken the construction of DTT (Divertor Tokamak Test) facility which will be held in Frascati, Italy. It aims to test various divertor models which can be integrated and used under different plasma confinement configurations. The current design
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Design of a position monitoring system for the ITER radial neutron camera Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-18 S. Cesaroni, D. Marocco, D. Marzullo, F. Moro, F. Belli, G. Brolatti, C. Centioli, E. Occhiuto, M. Riva, G. Rocchi, B. Esposito
The Radial Neutron Camera (RNC) is an ITER diagnostic system devoted to the radial measurement of the plasma neutron emissivity during ITER operation. In particular, plasma core measurements will be performed using 48 detectors located in the Ex-Port RNC subsystem and viewing the plasma through 16 lines-of-sight (LOS). Since discrepancies concerning the position of the LOS and the related optical path
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Managing the complexity of plasma physics in control systems engineering Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-18 T.F. Beernaert, M.R. de Baar, L.F.P. Etman, I.G.J. Classen, M. de Bock
The magnetized nuclear fusion plasma is a non-linear dynamic system with limits and constraints. It requires a sophisticated plasma control system with a wide variety of functions and components to ensure optimal and safe performance.
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Combined retrieval of multiple discharge signal waveforms based on distributed architecture Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-18 Hao Wang, Zhenshan Ji, Qiping Yuan, Ying Chen, Wenhui Hu, Ruirui Zhang, Bingjia Xiao
Processing time-series data, such as discharge signal waveforms, is essential for conducting data analysis in tokamak discharge experiments. Typically, researchers manually identify and filter shots by observing waveforms, which can be a time-consuming task involving data preprocessing and classification. In addition, there is often no intuitive interface for complex retrieval of distinct characteristic
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Preliminary design and analysis activities of the deuteron accelerator target for tritium breeding unit test Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-17 Sungjin Kwon, SeongHee Hong, Mu-Young Ahn, Hyun Wook Kim, Hyoseong Gwon, Yoo Lim Cheon, Nam Il Her, Seungyon Cho
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Manufacturing, installation, commissioning and operation of endoscopes for monitoring water-cooled divertor in Wendelstein 7-X Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-17 Joris Fellinger, Mathias Schülke, Marco Krause, Yu Gao, the W7-X team
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.
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Thermal-hydraulic assessment of the ITER IBED PHTS Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-16 E. Vallone, G. Agnello, L. Basili, A. Ciampichetti, P.A. Di Maio, D. Lioce, A. Quartararo, F.L. Venturi
One of the key elements for the success of ITER is the efficient removal of the thermal power generated by the deuterium-tritium fusion reaction within the confined plasma. The Integrated Blanket ELM/VS Coils and Divertor (IBED) Primary Heat Transport System (PHTS) is a pressurized, closed-loop cooling system designed to supply cooling water to the in-Vessel components. To support process design of
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Adaptive fuzzy PID control of high-speed on-off valve for position control system used in water hydraulic manipulators Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-16 Xing Yang, Defa Wu, Chenglong Wang, Chuanqi Gao, Heng Gao, Yinshui Liu
In contrast to oil hydraulics, water hydraulics presents distinct advantages, such as minimal environmental pollution, reduced operational cost, and no chemical reactions. Water hydraulic manipulators are instrumental in maintaining and handling fusion reactors. The valve-controlled hydraulic cylinder system (VCHCS) is the core of the manipulator's actuator. This study focuses on the control of a water
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Design and analysis of redundant electro-hydraulic-driven manipulator for tokamak vacuum vessel Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-15 Daoan Kang, Bangmeng Wang, Chao Huang, Dingan Song, Zhiyuan Zhang, Jindou Liu, Rongrong Luo, Pengyuan Li
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
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A remote detection method for leak location of external ports in a tokamak: Theory and preliminary experimental validation Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-15 Hailin Bi, Kunru Fu, Chunpeng Cheng, Guizhong Zuo, Huidong Zhuang, Jun Zhang, Wudi Wang
The vacuum vessel of fusion devices, equipped with an array of external ports characterized by intricate flange connections and welding joints, inherently carries a notable risk of leakage. With the impending introduction of deuterium-tritium fusion reactions, restrictions on technician access necessitate a more innovative approach. This paper introduces a groundbreaking leak detection method for fusion
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Position estimation of current-carrying filament using different magnetic sensors in ADITYA-U tokamak Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-15 Rohit Kumar, Harshita Raj, Suman Aich, Tanmay Macwan, Devilal Kumawat, S.K. Jha, Praveenlal Edappala, Kumarpal Jadeja, Kaushal Patel, R.L. Tanna, J. Ghosh
ADITYA-U tokamak entails a major upgrade of installing a new divertor coil system to attempt shaped-plasma experiments. Such modification required the installation of a new circular vessel and a full set of magnetic diagnostics. Therefore, an in-situ calibration of the magnetic sensor is performed in ADITYA-U by installing a current-carrying filament inside the vacuum vessel. Different Mirnov sensors
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Toward a high-fidelity tritium transport modeling for retention and permeation experiments Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-13 Masashi Shimada, Pierre-Clément A. Simon, Casey T. Icenhour, Gyanender Singh
Tritium Migration Analysis Program version 8 (TMAP8), the latest version of TMAP, was developed within the framework of the Multiphysics Object-Oriented Simulation Environment (MOOSE). Created at Idaho National Laboratory (INL), MOOSE is an open-source, dimension-agnostic, fully coupled, and fully implicit multiphysics platform featuring massively parallel computation capabilities. Using TMAP8, tritium
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Measurement of pressure drop of purge gas flow in unitary and binary pebble beds Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-13 Yong Liu, Yong Chen, Cong Wang, Lei Chen, Songlin Liu
This paper presents a study on the characteristics of purge gas pressure drop in unitary and binary pebble beds, considering different void fractions and diameter ratios. To model tritium purge gas in the breeder zones of blankets, a helium loop is constructed to provide a flow at 0.1–2.0 MPa with a maximum flow rate of 80 Nm/h. The experimental results show that the pressure drop increases linearly
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Monitoring of ozone production and depletion rates in a tritium-compatible system Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-13 Dominic Batzler, Max Aker, Robin Größle, Daniel Kurz, Alexander Marsteller, Florian Priester, Michael Sturm, Peter Winney
As surfaces are exposed to tritium, they will inevitably accumulate it, leading to the tritium memory effect. In order to reduce this effect, e.g. in analytic systems, decontamination methods are required. UV/ozone decontamination is known to be an efficient method, but its fundamental mechanism is not well known. In a dedicated UHV-compatible experiment, this method will be investigated systematically
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Status of the ITER TBM Program and overview of its technical objectives Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-12 Luciano M. Giancarli, Mu-Young Ahn, Seungyon Cho, Yoshinori Kawamura, Artur Leal-Pereira, Mario Merola, Yves Poitevin, Italo Ricapito, Qian Sheng, Hiroyasu Tanigawa, Hisashi Tanigawa, Jaap G. van der Laan, Xiaoyu Wang
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).
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Investigation of radiation power and carbon evaporation in infrared sensor bolometer Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-12 Dongjae Kwak, Dong-Kwon Kim, Min Uk Lee, Jaewook Kim, Byron Peterson, Jayhyun Kim, Gunsu S. Yun
The thin-foil infrared sensor bolometer (IRSB) has been developed on the KSTAR to measure the rapid change of intense radiation during plasma disruption. Despites its successful commissioning, several issues have been identified regarding the interpretation of the measured time scale of the radiation and the durability of the foil. Concerning data interpretation, we propose an extended analysis model
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Signal-distribution-based crack detection for divertor monoblock inspection using eddy current testing Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-12 Fanwei Yu, Takuma Tomizawa, Noritaka Yusa, Masayuki Tokitani
This study evaluated the applicability of eddy current testing to the surface inspection of a divertor plasma-facing unit in a Tokamak fusion reactor. Artificial slits were introduced on the surfaces of some samples. The eddy current testing was performed to detect these slits. A quantitative detection evaluation method based on evaluating the distribution differences is proposed. The results reveal
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Design of a multi-energy soft X-ray diagnostic for profile measurements during long-pulse operation in the WEST tokamak Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-12 O. Chellaï, L.F. Delgado-Aparicio, J. Wallace, T. Barbui, D. Bishop, R. Ellis, K.W. Hill, N.A. Pablant, B. Stratton, J. Wisniewski, R. Dumont, P. Lotte, P. Malard, WEST team
The W Environment in Steady-state Tokamakwas designed and built to test ITER-like tungsten plasma facing components in a long pulse (1000 s) scenario. Recently, a multi-energy soft x-ray diagnostic (MESXR) was installed in the WEST (W Environment in Steady-state Tokamak), to understand the sources, transport and confinement of high-Z impurities. The purpose of this work is to describe the engineering
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Requirements analysis for fusion reactor safety analysis software development based on COSINE Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-12 Yifei Wang, Tianze Bai, Xuebing Peng, Xinyuan Qian, Yuntao Song, Yixue Chen, Zhengze Wu
Systematic accident analysis is a significantly important and necessary phase for reactor primary heat transfer system (PHTS) design. Over the past several decades, many systematic accident analysis software and modules, such as RELAP5, CHATHARE, MELCOR, etc., have been developed to be applied on fission reactor accident simulation. However, there is no dedicated systematic safety analysis software
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Reliability research of hypervapotron under electromagnetic load for CFETR divertor Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-10 Shijun Qin, Shangxian Zhao, Jin Cao, Qingfeng Wang, Xing Feng, Shu Chen
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
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Methodology for radiofrequency electromagnetic analysis in the engineering of ITER Electron Cyclotron Heating Upper Launcher Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-10 Celia Gómez, Alexander Avilés, Ane Miren Larrea, Ander San Vicente, Iñigo Eletxigerra, Aymar du Rusquec, Olivier Dailly, Tindaro Cicero, Sandra Julià, Muriel Simon, Eduard Carbonell, Jose Manuel Arroyo, Melanie Preynas, Natalia Casal
Framed in the Final Design of ITER Electron Cyclotron Upper Launchers, a methodology has been developed for performing robust, reliable and efficient electromagnetic analysis of the millimeter waves and their interaction with the quasi-optical elements, launcher structures and finally the plasma. After a thorough research on the physics, mathematical formulations and Computational Electromagnetic Methods
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Computational Thermal-Fluid Dynamics analyses of borated water distribution in the Vacuum Vessel of the Divertor Tokamak Test facility Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-10 Roberto Bonifetto, Gianluca Barone, Mauro Dalla Palma, Antonio Froio, Emanuela Martelli, Federico Vair, Roberto Zanino, Andrea Zappatore
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Demonstration and evaluation of negative triangularity equilibria in the ARC fusion pilot plant concept Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-10 N. de Boucaud, T. Golfinopoulos, A. Marinoni
A numerical workflow is developed to explore the viability of running multiple plasma configurations in the ARC (Affordable, Robust, Compact) fusion pilot plant. Suitable cost functions were derived to evaluate more than 350,000 poloidal field coil sets based on currents required in the coils, induced stresses, and flexibility in producing plasma configurations. It is shown, for the first time, that
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ECH microwave measurement using in-vessel thermocouple array on JT-60SA Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-10 Satoshi Yamamoto, Manabu Takechi, Daigo Tsuru, Takayuki Kobayashi, Hibiki Yamazaki, Takaaki Iijima, Shigetoshi Nakamura, Akihiko Isayama
The in-vessel first wall made up of inertially cooled graphite tiles supported on stainless steel pedestals was installed in the vacuum vessel to protect the in-vessel components from the plasma and injected electron cyclotron eating (ECH) microwave in the integrated commissioning phase of JT-60SA. Sixteen thermocouples were installed in the inboard first wall, which is one of the in-vessel first walls
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Development of a remote automatic in-vessel calibration system for KSTAR motional Stark effect diagnostics Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-09 Juyoung Ko, Jinseok Ko, Myungkyu Kim
One of the essential calibration procedures for the motional Stark effect (MSE) diagnostic system includes obtaining the mapping responses of the angle evaluated from the demodulation of the photoelastic-modulated raw signals in the MSE polarimeter to the incident polarized light either parallel or perpendicular to the Lorentz electric field originated from a flux surface of a tokamak plasma. This
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Development and testing of lab-scale atmospheric molecular sieve bed with zeolite 4A adsorbent Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-09 Deepak Yadav, V. Gayathri Devi, Pragnesh Dhorajiya, Amit Munia, Amit Sircar, R. Bhattacharyay
Tritium Extraction System (TES) is one of the most important systems of fusion blanket. The two basic systems of TES are Atmospheric Molecular Sieve Bed (AMSB) and Cryogenic Molecular Sieve Bed (CMSB). AMSB removes ppm levels of water vapour (QO), while CMSB removes hydrogen isotopes (Q), oxygen, and nitrogen from helium purge gas. Though detailed parametric study of AMSB is very important for fuel
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Dispersion interferometry diagnostic at Globus-M2 Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-09 S.V. Ivanenko, A.l. Solomakhin, P.V. Zubarev, A.N. Kvashnin, Yu.V. Kovalenko, E.A. Puryga, V.V. Solokha, G.S. Kurskiev, N.S. Zhiltsov, K.D. Shulyatiev, A.D. Khilchenko, V.B. Minaev, P.A. Bagryansky
The dispersion interferometry diagnostic (DI) based on the CO-laser was commissioned at the Globus-M2 tokamak for absolute measurements of the line-integrated electron density (IED) along the chord in the equatorial plane. IED measurement error corresponding to the intrinsic noise of the acquisition device was below 10 with time and spatial resolution 20µs and 2 cm, respectively. Inference of the IED
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Circuit fault diagnosis for 12-pulse power converter in HL-3 based on Meta Pseudo Labels Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-09 Xiaoyan Wang, Weibin Li
The application of deep learning in the field of power electronics fault diagnosis has garnered considerable attention and research interest. In the coil power supply of the HL-3 tokamak device, there exist practical challenges in dealing with fault data from 12-pulse thyristor power converters, including large data volume, variable frequency operation, and sensor noise, among others. To solve these
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Selective adsorption properties of layered titanate for tritiated water Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-09 Yuki Edao, Yasunori Iwai, Hiroyoshi Mori, Nobuki Itoi, Toshiki Goto, Nobuhiro Kumada
Inorganic materials, such as concrete used in many fusion reactor facilities, adsorb tritiated water, which makes decontamination difficult and causes the generation of a large amount of nonflammable radioactive waste. Water strongly adsorbed by surface with inorganic materials causes tritiated water to be adsorbed more strongly than light water as an isotope effect, resulting in the accumulation of
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Properties test of 1D carbon-carbon composite tiles for the diagnostic calorimeter of CRAFT NNBI test platform Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-08 Yongjian Xu, Yue Hu, Ling Yu, Wuxiong Yin, Zengbo Yi, Xufeng Peng, Yahong Xie, Chundong Hu, Yuanlai Xie
An important aspect of the Comprehensive Research Facility for Fusion Technology (CRAFT) project is to conduct research on neutral beam injection based on accelerated negative ions, which is one of the main auxiliary heating methods. To evaluate beam characteristics, a diagnostic calorimeter based on a unidirectional carbon fiber-carbon matrix (CFC) composite has been developed in the negative ion
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Blanket test facility for mockup testing on water cooled ceramic breeder blanket system Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-06 Takuya Katagiri, Wenhai Guan, Atsushi Wakasa, Motoki Nakajima, Jae-Hwan Kim, Yuki Koga, Yuya Miyoshi, Takashi Nozawa, Takanori Hirose, Yoshinori Kawamura, Hiroyasu Tanigawa
Intensive research and development activities have been conducted for designing a water-cooled ceramic breeder (WCCB) blanket system. In this system 325 °C/15.5 MPa of pressurized water removes heat injected and accumulated in blanket containers made of a reduced activation ferritic/martensitic steel, F82H. The container has lithium ceramics and beryllium pebbles as tritium breeder and neutron multiplier
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Development of lower manipulator system for breeding blanket maintenance within large port-based tokamaks Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-06 Marcell Málics, László Poszovecz, Aditya Sinha, Ian Chiang, Imre Katona
Large port-based tokamaks for energy production are expected to be developed around breeding blanket technology, the purpose of which is to breed Tritium, provide shielding to the vessel walls, and carry heat transfer loops. These blankets are expected to be split into five segments in every torus section , each weighing up to 60 tons, and will require periodic replacement through the Upper Port for
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Deuterium enrichment by proton exchange membrane water electrolysis with electrolyte circulation Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Ibuki Sato, Koichiro Furusawa, Mikito Ueda, Hisayoshi Matsushima
Hydrogen isotopes will be new energy sources in nuclear fusion. In this study, a batch system for producing heavy water was developed in which pure water with deuterium (D) was circulated during proton exchange membrane water electrolysis. Deuterium was enriched by up to 94.7 at.% from 51.1 at.%. The D components of water and gas were analyzed during electrolysis. The relationship between the concentration
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Construction of GVR weight windows maps from very low density transport simulations Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-05 Gonzalo Farga-Niñoles, Francisco Ogando, Javier Alguacil, Patrick Sauvan
Fusion-related facilities present relevant neutron radiation fields even after penetrating through a considerable thickness of shielding material. Neutronic analyses performed via Monte Carlo codes, then, need Global Variance Reduction (GVR) techniques so that low statistical uncertainty is reached efficiently throughout the geometry.
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Reducing the user burden when running MELCOR for accident analysis for a tokamak Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-05 G. Karajgikar, J.J. Nebrensky
MELCOR is a software tool for accident analysis with a long history in the fission sector. A model has been developed for use in MELCOR, using proposed layouts of EU DEMO and assumptions based on current design choices, to facilitate safety design and development for fusion power plants.
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Progress and challenges of the ECH transmission line design for DTT Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-05 A. Moro, A. Bruschi, F. Fanale, P. Fanelli, E. Gajetti, S. Garavaglia, G. Granucci, S. Meloni, A. Pepato, P. Platania, A. Romano, A. Salvitti, L. Savoldi, S. Schmuck, M. Scungio, A. Simonetto, M. Turcato, E. Vassallo
The design of the Transmission Line (TL) as a part of the Electron Cyclotron Heating (ECH) system for Divertor Tokamak Test facility (DTT) is approaching the conceptual design maturity. With an ECH system of 16 MW installed for the first phase and with a total of 32 gyrotrons (170 GHz, 1 MW, 100 s) the TL design is undertaking the challenge of an evacuated Multi-Beam TL (MBTL) concept to deliver the
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Study on hot isostatic pressing conditions of ARAA for fabrication of the breeding blanket first wall Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-04 Hyoseong Gwon, Yunsong Jung, Yi-Hyun Park, Jae-Sung Yoon, Mu-Young Ahn, Suk-Kwon Kim, Seungyon Cho
The helium-cooled concept with solid breeding materials is one of the candidates for demonstration fusion power reactor (DEMO) breeding blanket. The first wall (FW) is a core component of the breeding blanket. The FW should keep the structural integrity during the normal plasma operation which is subject to high surface heat flux, neutron wall load, and high coolant pressure simultaneously. Various
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Thermal and structural analysis of the DONES target system under steady state and transient loading conditions Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-04 I. Catanzaro, P. Arena, D. Bernardi, G. Bongiovì, T. Dezsi, P.A. Di Maio, F.S. Nitti, S. Giambrone, M. Giardina, S. Gordeev, A. Quartararo, E. Tomarchio, E. Vallone
One of the crucial steps in the European Roadmap to the realisation of fusion energy is the design and construction of the Demo-orientated NEutron Source (DONES) facility. DONES is a fusion-like neutron source, based on the International Fusion Materials Irradiation Facility (IFMIF) concept, aimed at qualifying and testing the materials to be used in fusion reactors. The Target System (TSY) is a crucial
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Deuterium permeation of low-activation vanadium alloys possible for reuse in a short time in fusion reactors Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-04 Aoi Yoshizawa, Yuji Yamauchi, Takuya Nagasaka, Satoshi Tomioka, Yutaka Matsumoto, Naoki Higashi
Low-activation vanadium (V) alloys are alternative to reduced-activation ferritic/martensitic steels for the blanket structural material. To make the reusage of the alloys possible in a short time, new-design concepts of V alloy have been suggested in National Institute for Fusion Science. In the present study, the deuterium permeation properties of the new-designed V alloys were investigated.
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Neural network-based source biasing to speed-up challenging MCNP simulations Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-04 E. Martínez-Fernández, J. Alguacil, J. Sanz, R. Juárez
Nuclear analysis of fusion facilities, especially in the context of ITER, is complex due to the need for precise modeling of complex geometry and radiation sources. Monte Carlo (MC) codes, such as MCNP, are used in this context due to their high precision and capability to deal with these cases. To speed up calculations, variance reduction (VR) techniques are crucial for carrying out the simulations
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Assessment of different lifting devices for the shipping bays for DONES main building Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-04 Yan Wang, Martin Mittwollen
During the installation and maintenance phase of IFMIF-DONES (International Fusion Materials Irradiation Facility-DEMO Oriented Neutron Source), several components need to be transported vertically between different floors inside the DONES main building. There are totally four shipping bays inside the main building, two of which are mainly responsible for components transportation. The lifting devices
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Material flow planning in fusion test facilities Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-03 Timo Lehmann, Georg Fischer, Felix Rauscher, Sebastian Köhler, Fernando Arranz
Fusion test facilities are expansive, intricate structures designed to test materials for fusion power plants. For a seamless testing process, it's crucial that the machinery and equipment within these facilities are consistently maintained and promptly replaced if they fail. This necessitates the transportation of large, heavy machinery and equipment through the facility's narrow corridors, doors
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A review of pipe cutting, welding, and NDE technologies for use in fusion devices Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-03 Yao Ren, Robert Skilton
Piping systems that transport coolant and tritium breeding fluid are naturally an essential part of the support system of nuclear fusion power plants. Following a campaign of operations, the reactor is required to be shut down and maintained. Pipes connected to the reactor components are to be cut, re-welded or re-joined, and inspected using non-destructive methods. This paper outlines the candidate
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Development of real-time density feedback control on MAST-U in L-mode Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-03 G.L. Derks, B. Kool, C. Vincent, S. Elmore, S.S. Henderson, J.T.W. Koenders, J. Lovell, G. McArdle, B. Parry, R. Scannell, R. Sarwar, K. Verhaegh, M. van Berkel, EUROFUSION TOKAMAK EXPLOITATION TEAM, MAST-U team
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Development of potassium doped tungsten plate for fusion reactor applications Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Shuhei Nogami, Naoya Matsuta, Kei Miura, Kenta Okutani, Akira Hasegawa, Shigekazu Yamazaki, Seiji Nakabayashi, Tomohiro Takida
Since fusion reactors including DEMO and beyond are expected to be operated for a longer period of time than ITER, plasma-facing materials (PFM) will suffer by longer term heat load and non-negligible neutron irradiation damage. Therefore, for tungsten (W) materials as PFM, improvement of low temperature brittleness and suppression of recrystallization embrittlement and neutron irradiation embrittlement
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First-principles study of the energetics of Fe interstitial clusters in vanadium Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Xing Wang, Pengbo Zhang, Guofeng Li, Mingliang Wei, Haichuan Ji, Yichao Wang
The energetics of Fe interstitials and their clusters as well as the effect of Fe on the migration of self-interstitial (SIA) and vacancies in vanadium (V) are investigated by first-principles calculations. We determined the formation energies of mono-/di- Fe-V and Fe-Fe interstitial pairs by typical structures (<111>, <110> and <100>), and found that the <110> and mixed <111> dumbbell are the lowest
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Development of a real-time digital pulse acquisition and processing algorithm for compact neutron spectrometer on EAST Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Yong-Qiang Zhang, Li-Qun Hu, Wei Lu, Guo-Qiang Zhong, Hong-Rui Cao, Jin-Long Zhao, Li Yang, Rui-Xue Zhang, Ming-Yuan Xu, Qiang Li
Neutron emission spectroscopy (NES) diagnosis is an important technique in high-power thermonuclear fusion experiments for studying fast ions. In the fusion NES measurement on the Experimental Advanced Superconducting Tokamak (EAST), the liquid scintillator detector was employed. This selection was based on the detector's high sensitivity, fast time-response and superior γ discrimination properties
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Free boundary equilibrium determination of HL-3 discharges Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Y.Y. Zhong, L. Xue, W.L. Zhong, J.X. Li, T. Hoang, J. Garcia, J.F. Artaud, X. Song, Z. Yan, R. Ma, L. Liu, N. Wu, H. Heumann
In 2022–2023, HL-3 (previously known as HL-2M) realized the divertor discharge after its first plasma campaign. To analyze equilibrium configurations of HL-3, a free boundary equilibrium code FEEQS, based on the finite element method, has been used in recent experiments. Free boundary equilibrium determinations are carried out for discharges performed in both limiter and divertor configurations. Compared
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Overview of European efforts and advances in Stellarator power plant studies Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Felix Warmer, J. Alguacil, D. Biek, T. Bogaarts, G. Bongiovì, V. Bykov, J.P. Catalán, R.K. Duligal, I. Fernández-Berceruelo, S. Giambrone, C. Hume, M. Hrecinuc, R. Kembleton, J. Lion, T. Lyytinen, J.A. Noguerón Valiente, I. Palermo, V. Queral, D. Rapisarda, W.J. Rutten, L. Sanchis, X. Sarasola, K. Sedlak, A. Snicker, D. Sosa, F.R. Urgorri
The stellarator concept is a promising candidate for a steady-state fusion power plant, but currently lacking behind the tokamak developments. In order to bring the stellarator concept to maturity, a new EUROfusion Task has been established within the Work Package Prospective Research & Development (WP-PRD) called Stellarator Power Plant Studies (SPPS). This task addresses the stellarator-specific
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Proposal of new electrode supports in NBI for breakdown incidence reduction Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 Vincenzo Variale, Marco Cavenago, Vincenzo Valentino
In Neutral Beam Injectors (NBI) for fusion applications very large ion currents (40 A) at High voltage HV (1 MeV) will be generated and transported to the neutralizer. Keeping the accelerating electrodes at high voltage requires advanced electrical technologies and complex interfaces, tested at 700 kV and 1 MV, with frequent breakdowns, triggered also by local failures or voltage fluctuations. The
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Re-evaluation of shielding ability and induced activity of KSTAR with increased neutron yields Fusion Eng. Des. (IF 1.7) Pub Date : 2024-04-01 UkJae Lee, Nam-Suk Jung, Hee-Seock Lee, Hee-Soo Kim
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Damage assessment of the Faraday Shield in the RF driver of the negative ion source test stand SPIDER Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-30 Alessandro La Rosa, Giovanni Berton, Sylvestre Denizeau, Mauro Pavei, Daniel López Bruna, Enrico Miorin, Francesco Montagner, Valentina Zin
The SPIDER experiment, hosted in the Neutral Beam Test Facility of Padua, Italy, is the full-scale prototype of the negative ion source for ITER neutral beam injectors. Inside the ion source, eight radio frequency (RF) drivers are responsible for producing plasma via electromagnetic induction. The Faraday Shield lateral wall (FSLW) is a cup-shaped component of SPIDER RF drivers made out of electrodeposited
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Key technologies of manufacturing and testing for the ITER blanket shield block Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-30 Sa-Woong Kim, Si-Kun Chung, Jun-Sung Chang, Duck-Hoi Kim, Hyun-Soo Kim, Hyeon-Gon Lee, Ki-Jung Jung
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
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Eddy current actuated fast valve development for disruption mitigation applications Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-29 D. Nagy, D.I. Réfy
Shattered Pellet Injection (SPI) is a common technique used in Disruption Mitigation Systems to prevent or minimize the effect of plasma disruptions in Tokamaks. To operate an SPI system, a fast-acting valve is needed for accelerating pellets. An eddy current actuated high-pressure fast valve was developed in 2021, dedicated to the SPI system of the ITER DMS Support Laboratory at the Centre for Energy
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Analysis of transport-activation internal coupling method implemented in Monte Carlo code cosRMC Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-29 Shengzhe Wang, Shichang Liu, Jun Wu, Yixue Chen
When fusion nuclear reactor is in operation, neutron activation causes a large number of radioisotopes. Activation calculation is a very important step in both of reactor shielding calculation and radiation safety analysis. The capability of transport-activation coupling calculation was developed in Monte Carlo code cosRMC though the built-in burnup solver Depth. In the process of neutron transport
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Induction brazing of DEMO's large bore cooling pipes Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-29 Jenő Kádi, Miklós Palánkai, Márton Gregor, József Szőke, Tétény Baross, Gábor Veres, Oliver Crofts, Tristan Tremethick
The DEMO first wall blanket components will be exposed to high neutron fluence and high levels of thermal energy, the latter being removed by a coolant fluid transmitted out of the reactor via service piping. Due to the thermal and nuclear loading on the first wall, the components will need to be periodically removed and replaced. Installation of new blankets requires the joining of their service pipes
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In-vessel inspection system: Development and testing activities of high vacuum and temperature technologies for fusion remote handling Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-29 ManoahStephen M, Naveen Rastogi, Ravi Ranjan Kumar, Krishan Kumar Gotewal, Jignesh Chauhan, Yuvakiran Paravastu, Dilip Raval, Siju Geroge
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