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Total ionizing dose estimation of ITER upper port remote handling equipment Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-15 Shanshuang Shi, Chang-Hwan Choi, Taku Yokoyama, Hongtao Pan
This study delineates a comprehensive estimation of the integrated Total Ionizing Dose (TID) for the ITER Upper Port Remote Handling Equipment (UPRHE). The primary objective is to assess the radiation impact and devise strategies to attenuate its effects on the UPRHE components. Radiation levels in various operational zones are deduced from the ITER radiation map. This analysis includes a meticulous
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Evaluation of the impact of plasma operation scenarios on the fusion reactor blanket design using an integrated numerical plasma and neutronics analysis suite Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-15 Rin Choi, Youngin Kim, Sangjin Park, Boseong Kim, Min Ki Jung, Myeongseop Jeon, Chan-Young Lee, Hyung Jin Shim, Mu-Young Ahn, Jisung Kang, Yong-Su Na
A very limited operation scenario has been used for fusion neutronics for the design of tritium breeding blankets. However, many advanced operation scenarios are under development and the impact of choosing a plasma operation scenario for fusion neutronics, which requires integration of neutronics analysis with plasma analysis, needs to be evaluated. An automated process from plasma analysis to neutronics
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Continuous deuterium extraction from falling lithium-lead droplets in a vacuum Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-15 Fumito Okino, Yukinori Hamaji, Juro Yagi, Teruya Tanaka
This study experimentally verifies continuous deuterium extraction efficiency from falling liquid lithium-lead (LiPb) droplets in a vacuum, as an alternative to tritium The obtained efficiencies, from five runs, were between 0.59 and 0.69. The corresponding dispersion coefficient of deuterium in LiPb droplets is (2.6 + 0.8 – 0.6) × 10 ms. These results closely resemble those obtained in a proof-of-principle
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Highly efficient and direct recovery of low-pressure hydrogen isotopes from tritium extraction gas by PdY alloy membrane permeator Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-15 Xianglin Wang, Xingwen Feng, Lei Yang, Yongtao An, Yong Yao, Kelin Chen, Jiangfeng Song, Yan Shi, Changan Chen, Wenhua Luo
Low-pressure hydrogen recovery is crucial for the fusion reactor deuterium-tritium fuel cycle, and the usual treatment is achieved through multiple steps of enrichment and separation, which has the drawbacks of complex processes and high energy consumption. In this work, a membrane permeator including 12 sets of PdY at% alloy tubes has been investigated for the permeability and extraction of hydrogen
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Estimation of nuclear properties in a deuteron–lithium target using JENDL/DEU-2020 for IFMIF and similar irradiation facilities Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-15 Takeo Nishitani, Sachiko Yoshihashi, Yuuki Tanagami, Shota Amano, Akira Uritani, Saerom Kwon, Keitaro Kondo, Kohki Kumagai, Kentaro Ochiai, Satoshi Sato
An accelerator-based neutron source powered by deuteron–Li (d–Li) reactions is among the most promising neutron sources for fusion materials irradiation facilities, such as IFMIF, IFMIF-DONES, and A-FNS. The yield estimation of neutrons and other particles is critical to the design of these facilities. Presently, the McDelicious code is employed for the aforementioned yield estimations, although it
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Design study on fuel buffer system for continuous processing of tritium processes considering burn and dwell operation scenarios Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-15 Jae-Uk Lee, Junyoung Hur, Min Ho Chang, Hyun-goo Kang, Dong-you Chung, Pil-Kap Jung, Sei-hun Yun
The most important purpose of the tritium processes in the fusion fuel cycle is to stably supply tritium and deuterium to reduce operational uncertainty in the Tokamak. To achieve this, it is desirable for the process operation to be designed as a continuous process with a steady flowrate, pressure and composition. In this study, we aim to propose a modified concept of storage and delivery system,
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Applicability study of mechanical multi-pipe connections for DEMO breeding blanket maintenance concept Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-14 A. Azka, K.J. Büscher, M. Mittwollen, C. Bachmann, G. Janeschitz
Maintenance of DEMO breeding blanket includes the removal and replacement of plasma facing components, for example the breeding blankets (BB). For access, multiple coolant pipes need to be removed. As an option to reduce downtime and increase maintenance speed, a so-called mechanical multi-pipe connection (MPC) concept is developed to allow removal of multiple pipes at the same time using remotely
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Approach to spatial integration on a novel and complex major project – STEP concept Tokamak Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-14 Adam Woods, Alex Hellend, Jonathan Keep, Emily Kimbrey, James Hagues
The purpose of this work was to establish an approach to drive the development, at pace, of a commercially relevant Spherical Tokamak Concept. Building on the table of parameters and idealised geometry generated from the work of S. Muldrew et al. , applying core principles, tools, and approaches. The generation and utilisation of a reflective 3D Spatial Model to represent the current concept, now forms
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Molecular dynamics simulations of the effect of porosity on heat transfer in Li2TiO3 Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-14 Megha Sanjeev, Mark R. Gilbert, Samuel T. Murphy
Heat transfer is a key consideration in the development of tritium breeder blankets for future fusion reactors. For solid tritium breeder materials there is a fine balance to be struck between high levels of porosity to encourage tritium release and minimising it to maintain the thermal and mechanical properties. Therefore, in this work we employ molecular dynamics simulations to understand how the
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Investigation of applicability of long-distance LP system for Li target diagnostics in Fusion Neutron Source Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-14 Takafumi Okita, Shori Okunaka, Taiki Kojima, Eiji Hoashi, Makoto Oyaizu, Kentaro Ochiai
The beam target for generating a fusion neutron field in the Advanced-Fusion Neutron Source (A-FNS) is currently designed as a one-side free-surface liquid lithium (Li) jet flowing along a vertical concave wall. Characteristics of its thickness fluctuation were diagnosed in the EVEDA (Engineering Validation and Engineering Design Activity) Lithium Test Loop (ELTL), which has almost the same scale as
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Design of a Plasma neutraliser for a Fusion reactor or as an upgrade to the ITER heating neutral beam injectors Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-14 R S Hemsworth, P Veltri
The global efficiency of neutral beam injectors to be used on a fusion reactor needs to be of the order of, or greater than, 60 %, which is considerably higher than the efficiency of presently operating NBIs. The required efficiency can only be achieved by increasing the neutralisation efficiency from the ≈56 % obtained with the gas neutralisers currently used, which can be done using a plasma or photon
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Analysis of scattered light from multi-blade and V-grooved laser dumps in Thomson scattering diagnostic Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-14 Shumei Xiao, Qing Zang, Xiaofeng Han, Jian Zhou, Jianwen Liu, Liqun Hu
Laser dump is an essential optical component in Thomson scattering diagnostic. Both multi-blade and V-grooved dumps can increase the ability of absorbing laser energy by using multiple surfaces to reduce the energy density on the wall surface. Still, the wall edges of the dump easily produce scattered light. The distribution and intensity of scattered light escaping from these two dumps, as well as
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Preliminary nuclear analysis of HYLIFE-III: A thick-liquid-wall chamber for inertial fusion energy Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-13 Francisco Ogando, Michael T. Tobin, Wayne R. Meier, Gonzalo Farga-Niñoles, Jaime Marian, Susana Reyes, Javier Sanz, Conner Galloway
This paper provides neutronics analyses of the Xcimer Energy Corporation (XEC) HYLIFE-III Inertial Fusion Energy Power Plant concept. This design is based on the thick-liquid-wall HYLIFE-II reactor, but with much larger fusion yield, due to enhanced driver energy. Although HYLIFE-II neutronics was extensively studied, the differences between the two concepts suggested new analyses are required. Further
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Data driven modeling of heavy-duty joint system for DEMO manipulators: An initial study from MPD joint simulation Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-13 Ming Li, Huapeng Wu, Changyang Li, Zhixin Yao, Qi Wang, Heikki Handroos, Tom Deighan, Brace William, Olive Crofts
This study investigates the subspace modeling method in developing surrogate model of a heavy-duty joint system in a muti-purpose deployer. The joint system is constructed in the Matlab SIMULINK Simscape environment, which includes the multibody dynamics with friction models, meshing loss models and viscous models. The input and output data generated in simulation are used to develop the subspace model
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Numerical and experimental analysis of magneto-convective flows around pipes Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-13 C. Mistrangelo, L. Bühler, C. Courtessole, C. Koehly
Liquid metal flows exposed to intense magnetic fields play a fundamental role in the development of blankets for nuclear fusion reactors, where they serve to produce the plasma-fuel component tritium and transport the generated heat. When the liquid metal circulates in the blanket, it interacts with the plasma-confining magnetic field leading to the induction of electric currents. Electromagnetic forces
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Thermofluid-dynamic and thermal–structural assessment of the EU-DEMO WCLL “double bundle” Breeding Blanket concept left outboard segment Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-12 P.A. Di Maio, I. Catanzaro, G. Bongiovì, F.M. Castrovinci, P. Chiovaro, S. Giambrone, A. Gioè, F.A. Hernández, I. Moscato, G.A. Spagnuolo, A. Quartararo, E. Vallone
As part of the activities performed by the DEMO Central Team (DCT) to study alternative configurations of the Water-Cooled Lead Lithium (WCLL) Breeding Blanket (BB), to overcome the open issues that emerged at the end of the Pre-Conceptual Design phase, a new concept, the WCLL BB “double bundle” (db), was developed. This concept adopts an array of db tubes poloidally distributed inside the breeding
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Kinematics analysis of a novel hybrid manipulator for optomechanical modules assembly Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-12 Xiaoyong Wu, Shulin Wu, Congzhe Wang, Jiufei Luo, Yujin Wang
A novel hybrid manipulator is proposed in this work, in order to assemble optomechanical modules in an inertial confinement fusion facility (SG-Ⅲ). Differ from general serial and parallel manipulators, both the inverse and forward kinematics problems of hybrid manipulators are difficult to be solved. Based on an analytic method, the position and orientation mapping functions between the optomechanical
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Interactive simulations as a tool for logistics and maintenance of IFMIF-DONES Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-12 Andrea Benito-Fuentes, Fernando Arranz, Eduardo Ros, Jesús A. Garrido
Pre-configured virtual reality (VR) simulations of the logistics and maintenance processes have proven to be useful for identifying potential design issues as well as planning operations during an early design phase of facility. But VR simulations can also be used to deeply explore the feasibility of these procedures in a more interactive manner, so that we can identify potential risks and difficulty
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Design status of the neutron and gamma-ray diagnostics for the Divertor Tokamak Test facility Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-12 D. Marocco, M. Angelone, F. Belli, F. Caruggi, G. Croci, B. Esposito, G. Gandolfo, G. Gorini, G. Grosso, M. Nocente, F. Panza, M. Pillon, F. Pompili, D. Rigamonti, G. Rocchi, J. Scionti, M. Tardocchi
In the frame of the design activities of the Divertor Tokamak Test (DTT) facility the development of a comprehensive set of neutron and gamma-ray diagnostics is on-going in order to enable measurements of: neutron yield, neutron yield rate, neutron emissivity over a poloidal section through the plasma; neutron emission spectrum; runaway electrons induced bremsstrahlung radiation and gamma-ray emission
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Waste management strategy for EU DEMO: Status, challenges and perspectives Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-12 S. Rosanvallon, P. Kanth, J. Elbez-Uzan
One of the top-level safety objectives for EU DEMO design and operation is to protect workers, the public and the environment from harm and thus to minimize radioactive waste hazards and volumes and ensure that the legacy to the future generation is limited.
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Numerical simulation of buoyant flow in a vertical channel for a plasma-facing component Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Shin-ichi Satake, Hitoshi Nagasawa, Ryunosuke Imai
Cooling the plasma-facing component of fusion reactors requires investigations of the characteristics of the thermofluids used in the reactors' high-temperature plasma. The working fluid (pressurized water) influences the design of advanced fusion reactors. We considered the vertical flow geometry with single-side wall heating as a flow model to examine the influence of buoyancy in the system, where
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An obstacle avoidance path planning algorithm to simulate hyper redundant manipulators for tokamaks maintenance Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Sara Buonocore, Andrea Zoppoli, Giuseppe Di Gironimo
The present work proposes an obstacle avoidance path planning algorithm for virtual simulations of hyper-redundant manipulators, with the possibility to customize the optimization criteria to select the best trajectory given in output. For test purposes, the effectiveness of the proposed Inverse Kinematic algorithm has been tested by simulating the Remote Maintenance (RM) tasks conducted by the HyRMan:
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Analysis of the integral zeolite molecular sieve process for helium CPS application Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Vincenzo Narcisi, Alessia Santucci
Zeolite Molecular Sieves (ZMSs) are commonly adopted in tritium handling facilities for impurity removal from gaseous streams, particularly for tritiated water trapping. Several activities have been conducted to characterize the adsorption and desorption behaviour of the sieving materials. Instead, less attention has been put on the analysis of the integral process comprising the adsorption of the
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Experimental activities in view of hydrogen- deuterium equilibrium reaction in dielectric barrier discharge reactor: Effects of plasma treatment parameters and reactor design Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-11 Jiamao Li, Sheng Liu, Xin Wang, Junyan Wang, Chao Chen, Xiulong Xia, Lei Yue, Jiao Gao, Jingwei Hou, Houwen Huang, Chengjian Xiao
Cryogenic distillation (CD) is a crucial process for separating and concentrating tritium in the tritium fuel cycle of fusion reactors. The equilibrator plays a vital role in decomposing HD, DT, and HT to improve the separation efficiency of hydrogen isotopes in CD. As the equilibrium reaction temperature decreases, a larger proportion of HD, HT, and DT decomposes into H, D, and T, which enhances the
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Nuclear analyses for the integration of ITER equatorial Port 2 Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-11 S. Noce, R. Villari, A. Colangeli, D. Flammini, N. Fonnesu, P. Gaudio, M. Gelfusa, E. Grasso, J. Guirao, G. Mariano, F. Mercuri, F. Moro, A. Previti, P. Shigin, S. Soro, V.S. Udintsev, I. Wyss
The present work is devoted to nuclear analyses in support of the ITER diagnostic Equatorial Port 2 (EP#2) integration. ITER EP #2 is a diagnostic port based on the long-modular Diagnostic Shielding Module (DSM) housing the following systems: Disruption Mitigation System (DMS) in DSM#1 and #3 and X-Ray Crystal Spectroscopy Core (XRCS-Core) in DSM#2. Ensuring adequate radiation shielding is a major
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Applicability to DEMO breeding blankets of neutron measurements techniques from fission reactors Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-10 P. Filliatre
Neutron measurements are routinely performed in fission reactors. A lot of experience has been gained at CEA in designing, prototyping and testing neutron sensors and associated acquisition systems in several nuclear facilities with a wide range of demanding conditions. In this paper, the applicability of this experience to neutron diagnostics in the breeding blankets of DEMO is addressed, as a complement
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Thermo-mechanical analysis of the components in the launcher transmission line of ITER ex-port plug collective Thomson scattering Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-10 Esther Rincón Rincón, Emilio Blanco, Mercedes Medrano, Juan José Imaz, Yago Villalobos, Laura Maldonado, Paulo Varela, Yong Liu, Victor Udintsev, Stefan Schmuck
ITER Collective Thomson Scattering (CTS) diagnostic system is designed to analyze the alpha particles resulting from Deuterium-Tritium fusion reactions. It consists of one launcher and nine receiver transmission lines. The launcher line transports the high-power microwave emission of 1.2 MW from the gyrotron source to the front-end, while the receiver lines transport the collected microwave emission
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Engineering design status of IFMIF-DONES High Energy Beam Transport line and Beam Dump system inside the TIR and RIZ Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-10 Llorenç Macià, Oriol Nomen, Manel Sanmartí, Daniel Sánchez-Herranz, Claudio Torregrosa-Martin, Ivan Podadera, Concepción Oliver, David Jiménez-Rey, Fernando Arranz, Gioacchino Miccichè, Francisco Ogando, Yuefeng Qiu, Arkady Serikov, Philippe Cara, Angel Ibarra
(International Fusion Materials Irradiation Facility – DEMO Oriented Neutron Source) is a fusion materials testing facility that is currently being designed under the framework of a work package of the EUROfusion Consortium. It will use a 125mA at 40MeV deuteron beam to generate a high neutron flux through stripping nuclear reactions in a liquid lithium target. The High Energy Beam Transport line (HEBT)
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Fusion waste requirements for tritium control: Perspectives and current research Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-10 Mark R. Gilbert, Žilvinas Zacharauskas, Philippa Almond, Naomi Scott-Mearns, Stephen Reynolds, Mikhail Yu. Lavrentiev
The successful realisation of energy production through the fusion of deuterium and tritium will necessarily lead to the generation of waste contaminated with tritium. Not only will some of the tritium fuel permeate into components of fusion reactors and their wider fuel cycle, but tritium will also be generated directly in materials exposed to the high neutron fluxes via nuclear break-up reactions
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Development of the high temperature PbLi experimental facility for CFETR Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Kecheng Jiang, Yue Yu, Lei Chen, Juancheng Yang, Jiandong Zhou, MingJiu Ni, Songlin Liu
Chinese Fusion Engineering and Test Reactor (CFETR) aims to demonstrate the fusion energy production up to 200MW, and finally reach commercial electricity power level 1GW. As one key component, the blanket is in charge of tritium breeding, neutron shielding and energy conversion. The liquid breeder blanket has many important advantages of simpler structure, tritium extraction and fuel replenishment
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Nuclear analyses in support of ITER ex-port Radial Neutron Camera design Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Fabio Moro, Daniele Marocco, Francesco Belli, Dariusz Bocian, Giorgio Brolatti, Silvia Cesaroni, Andrea Colangeli, Davide Flammini, Nicola Fonnesu, Davide Falco, Giada Gandolfo, Ryszard Kantor, Jerzy Kotula, Domenico Marzullo, Enrico Occhiuto, Rafal Ortwein, Alberto Previti, Dustin Sancristóbal, Rosaria Villari, Basilio Esposito
The Radial Neutron Camera (RNC) is a diagnostic system located in the ITER Equatorial Port 1 (EP01) composed by two sub-systems (i.e.: in-port and ex-port RNC) probing a poloidal section of the plasma through a set of fan-shaped Lines of Sight (LOS). The RNC is designed to provide a time resolved measurement of the neutron and alpha particles source profiles and of the total neutron source strength
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Fast Shutter optioneering study for the ITER Disruption Mitigation System Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-08 A. Zsákai, D.I. Réfy, E. Walcz, D. Nagy, D.Z. Oravecz, L.R. Csiszár, S. Jachmich
ITER needs a Disruption Mitigation System (DMS) for protection from the consequences of plasma disruptions during high-power operation. The current DMS for ITER is based on the Shattered Pellet Injection (SPI) technology. This works on the basic principle of a cryogenic pipe gun with around a half-meter-long acceleration barrel. It is important to avoid any contact between the pellet and the internal
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Design and Development of Hydrogen Isotopes Extraction System at IPR Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Rudreksh B. Patel, Pragnesh B. Dhorajiya, Sudhir Rai, P.A. Rayjada, Deepak Sharma, Aditya Verma, Amit Sircar, Rajendra Bhattacharyay, Paritosh Chaudhuri
Tritium breeding, recovery and safe storage are very important for achieving self-sustainability in a nuclear fusion reactor. Indeed, it is one of the essential task in terms of safety and fuel cycle aspects. Usually, tritium breeding is performed in the blanket module. In Indian blanket module, there are mainly two types of tritium breeder materials, one is Lithium based ceramic pebbles LiTiO (Lithium
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Experimental investigation of the corrosion behavior of Eurofer97 steel in contact with Lithium ceramic breeder pebbles under specific Helium Cooled Pebble Bed breeding zone atmosphere Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-08 Regina Krüssmann, Bradut-Eugen Ghidersa, Mario Walter, Thi Tra My Nguyen, Viktoria Weber, Philipp Heger, Frederik Arbeiter, Georg Schlindwein, Guangming Zhou, Francisco A. Hernández Gonzalez
For sub-sized fatigue specimens made of EUROFER97 in unconstrained contact to ceramic breeder pebbles, exposed to purge gas conditions for different durations in an oven, a chemical surface attack was observed which led to a significant reduction of fatigue lifetime [Aktaa et al., 2020]. To better reproduce the flow of the purge gas in the breeding zone of a Helium-Cooled Pebble Bed Breeding Blanket
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An extension of NASCA for SDDR analysis of mobile activated components with global variance reduction Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-07 Yu Zheng, Peng Lu, Xiaokang Zhang, Yuefeng Qiu, Songlin Liu, Shanliang Zheng
During the plasma operation period of fusion reactor facilities, neutrons generated by DT reaction can penetrate deeply into structures and induce strong activation. Highly activated components have to be transferred from the on-load position to the hot cell for the purpose of maintenance. This movement produces a high-dose field, originating from the mobile radiation source, which could affect the
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Brazing SMART tungsten alloys to RAFM steels by Titanium-Zirconium-Beryllium brazing alloy Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-07 V.O. Kirillova, N.S. Popov, J.A. Gurova, D.M. Bachurina, X. Tan, I.V. Fedotov, A.N. Suchkov
Various DEMO designs imply the use of tungsten as plasma facing material and RAFM steels as structural material. LOCA concerns have shifted the focus from pure tungsten to tungsten smart alloys (SA), thus presenting a new challenge of obtaining joints of SA with RAFM steels. In this work, high-temperature brazing via TiZr4Be amorphous alloy is suggested as a prospective production technology. Use of
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Final opto-mechanical design of ex-vessel components for the Wide Angle Viewing System diagnostic for ITER Equatorial Port 12 Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-07 M. Medrano, C. Pastor, A. Soleto, R. Carrasco, F. Lapayese, A. de la Peña, A. Pereira, E. Rincón, S. Cabrera, A. Fernández, F. Ramos, C. Rodríguez, F. Mota, V. Queral, R. Lopez-Heredero, S. Vives, F. Le Guern, J.J. Piqueras
The ITER WAVS (visible and infrared Wide Angle Viewing System) provides measurements of surface temperature for the plasma facing components by infrared thermography. One of its most important roles of WAVS is to protect the plasma facing components from damage. It also takes images of the plasma visible spectral range emission. The WAVS diagnostic consists of 15 views distributed in four equatorial
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Final design analysis of the optical hinge of the wide angle viewing system for ITER equatorial port 12 Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-07 S. Cabrera, E. Rincón, M. Medrano, F.Le Guern, E. Rodríguez
The ITER Visible/Infrared Wide Angle Viewing System (WAVS) is a machine protection optical diagnostic that is being developed for ITER. The first two ex-vessel components of WAVS are the Optical Hinge (OH) and Optical Relay Unit (ORU) and both share a common support structure. Equatorial Port 12 (EP12) WAVS should be operational for the first plasma and its facing its Final Design Review (FDR). It
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Structural and thermal fluid dynamics analyses of the ITER Pressure Suppression System considering no stable steam condensation regimes Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Luca Berti, Alessio Pesetti, Francesco d'Errico, Donato Aquaro
At the University of Pisa, an experimental campaign was performed with the purpose of qualifying the ITER Vacuum Vessel Pressure Suppression System (VVPSS) and studying the Direct Contact Condensation at sub-atmospheric conditions. With financial support from ITER Organization, a Large-Scale Experimental Facility was designed and built to investigate the steam condensation in the operation conditions
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Characteristics of purge gas flow in regular and random closed packed pebble beds through CFD-DEM Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Shreyaksh Kashoudhan, Rishabh Patel, Raghuram Karthik Desu, P Kaushik
The flow characteristics of the purge gas in fusion pebble beds play an essential role in the robust design of fusion breeder units. Various factors, viz. packing fraction, pebble size and distribution, and packing structure, influence the flow characteristics. Recent investigation in pebble packing in breeder units has shown to achieve near to regular closed through various filling strategies viz
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Engineering design and testing of rotary joints’ driving systems for the CFETR multi-purpose overload robot Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Hongbin Huang, Haoyin Wang, Yiming Wang, Youmin Hu, Bo Wu, Jie Liu, Ping Su, Hongtao Pan, Yang Cheng, Yong Cheng
This study focuses on the engineering design and testing of the rotary joints' driving systems for the China Fusion Engineering Test Reactor (CFETR) Multi-Purpose Overload Robot (CMOR). CMOR is a critical component in the CFETR. The CFETR tokamak serves as an engineering test reactor in China for magnet confinement fusion research. CMOR, a 7-DOF manipulator, is specifically designed for internal maintenance
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The influence of the extraction of gas/liquid samples on the isotopic separation regime of the CECE process Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Anisia Mihaela Bornea, Marius Valentin Zamfirache
At National R&D Institute for Cryogenics and Isotopic Technologies (ICSI), in the framework of a research project within the national research program "Core", an experimental installation was created and put into operation for the development of the CECE (Combined Electrolysis Catalytic Exchange) process, both for the separation of tritium in view of waste decontamination and its recovery, and for
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Hydrogen extraction from a helium purge gas comparison of the ZrCo/ZAO getter material Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 N. Bekris, M. Sirch, I. Cristescu, R. Groessle
From the known getter materials proposed for the storage, supply, and recovery of hydrogen isotopes the interalloy ZrCo has been selected as reference material. However, because of disproportionation occurring after prolonged thermal cycling, ZrCo loses its ability to reversibly absorb, or desorb tritium. Therefore, several other interalloys such as ZAO have been suggested as potential candidate materials
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Galvanic process for Cu-infiltration of W fibre-reinforced heat sinks Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Patrick Junghanns, M. Busch, A.V. Müller, S. Roccella, K. Hunger, J.-H. You, R. Neu, J. Riesch, J. Boscary
A challenging aspect in view of the realization of a future magnetic confinement fusion reactor is the design and manufacture of highly loaded divertor target plasma-facing components (PFCs) which have to sustain intense particle, heat and neutron fluxes. In this context, tungsten-copper (W-Cu) composites are currently being investigated as potentially advanced heat sink materials for PFCs. The development
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Evaluation of the strength of FeCrAl alloy/surface Oxide film interface by micro double-notch shear test Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-06 Naoko Oono-Hori, Angela Dannea Anak Andria, Xiangyu Wu, Hao Yu, Ryuta Kasada
The shear strength of the film formed by pre-oxidation treatment of FeCrAl(-ODS) alloys, likely to be used as pipework material for liquid LiPb breeder blankets, was evaluated by micro double-notch shear testing. Fracture generally occurred at the interface between the oxide film and the alloy substrate, but the pop- in the stress-strain curves was difficult to interpret, so all the test results in
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Impact of trace silicon on irradiation hardening and embrittlement of RAFM steel subjected to neutron irradiation Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-05 Lei Zhang, Yifan Zhang, Wentuo Han, Xiaoou Yi, Pingping Liu, Kenta Yoshida, Takeshi Toyama, Qian Zhan, Yasuyoshi Nagai, Somei Ohnuki, Akihiko Kimura, Farong Wan
The content of silicon (Si) element in reduced activation ferritic/martensitic (RAFM) steels varies significantly among different candidates of fusion reactor structural materials. Si is not intentionally added to the RAFM steel but rather remains as a residue during the deoxidation process in steel fabrication. To control and reduce the Si content, removing processes and special treatments are required
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LITEC: An experimental facility for the validation of the IFMIF-DONES Impurity Control System Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-05 B. Garcinuño, A. García, M. Sánchez-Arenillas, D. Rapisarda
The Impurity Control System (ICS) of IFMIF-DONES is devoted to assuring lithium purity. After its circulation through the target, some undesired products will contaminate the lithium: deuterium, hydrogen, tritium, and beryllium. In addition, due to the high velocity of the flowing lithium, the target assembly, piping system, and other auxiliaries are subjected to a significant corrosion rate, thus
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DEMO toroidal field coil fast discharge unit integration studies Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-04 Thomas Franke, Janos Balazs Bajari, Aljaž Čufar, Alberto Ferro, Curt Gliss, Roberto Guarino, Dieter Leichtle, Pietro Zito
The Fast Discharge Units (FDUs) of the Superconducting (SC) Toroidal Field (TF) coils in the European demonstration fusion power plant DEMO warrant the machine integrity over its full lifetime against severe failure events, such as SC coil quenches or any other plant events requiring the safe TF magnet system discharge. A low (75 kA) and a high current (105 kA) configuration are under study for the
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Nuclear analyses for the ITER Diagnostic Equatorial Port 8 Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-04 A. Colangeli, A. Chaudhary, D. Flammini, N. Fonnesu, J. Guirao, K. Gupta, S. Kalwale, G. Mariano, S. Noce, F. Moro, A. Previti, M. Quatrevaux, P. Shigin, V.S. Udintsev, R. Villari
The ITER diagnostic Equatorial Modular Port Plugs (EPPs) are located in the equatorial vacuum vessel ports. Port plugs consist of structural components, shielding elements, services and tenant systems and they are designed to operate in a harsh nuclear environment. Main functions of the port plugs are to provide nuclear shielding and to host tenants systems. Shielding performance optimization and calculation
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Investigation of polarizer design with grooves of arbitrary profile by the coordinate transformation method for the ECRH system Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-04 Xixuan Chen, Sifen He, Donghui Xia, Houxiu Xiao, Zhijiang Wang, Yuan Pan
The coordinate transformation method is widely utilized in the design of polarizers. However, the numerical solution of polarizers with discontinuous structures, such as rectangular grooves, poses a challenge for the coordinate transformation method. To solve this problem, we expand the profile functions of the grooves with the Fourier series by considering the periodic groove structure. In this way
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Assessment of radiation dose rate in main building area of A-FNS facility Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-04 Saerom Kwon, Masayuki Ohta, Kentaro Ochiai, Ryota Sakamoto, Satoshi Sato
Fusion neutron source facilities intended for material irradiation are characterized by their high-intensity neutron generation, underscoring the paramount significance of dose assessment in facility design. As one of radiation shielding analyses taking into account the facility building, the radiation dose rate in the A-FNS has been calculated by Monte Carlo simulation. The calculation model was created
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Overview of Broader Approach activities Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-02 H. Dzitko, P. Barabaschi, P. Cara, Y. Carin, S. Clement Lorenzo, S. Davis, E. Di Pietro, B. Fourestié, D. Gex, Y. Ishii, M. Hanada, K. Hasegawa, Y. Ikeda, S. Ishida, N. Nakajima, H. Shirai, K. Takahashi, H. Takenaga, M. Taniguchi, M. Yagi, the Broader Approach Integrated Project Teams
The Broader Approach (BA) activities aim to complement the ITER project and to promote the early realization of fusion energy through research, development, and tests of technologies supporting the future demonstration fusion reactor (DEMO). These activities are implemented under the BA agreement, which was signed and ratified in 2007 between Euratom and the Government of Japan. In essence, the BA
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Impact of helium working pressure on the thermal-hydraulic performance of FW/Blanket for fusion reactor Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-02 Jing Chen, Xiaoyong Wang, Zhou Zhao
The pressure of helium is a critical parameter for the helium coolant, and it has a significant impact on the performance of the FW (First Wall)/Blanket. To achieve a deeper understanding of how helium pressure affects the thermo-hydraulic performance of the FW/Blanket, this study examines its influence on both the physical properties of helium coolant and its overall thermo-hydraulic performance.
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Design and analysis of actively-cooled, edge-transport diagnostic for long-pulsed operation in WEST Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-02 Arnold Lumsdaine, Michael DeVinney, Ezekiel Unterberg, Brendan Quinlan, Jessica Wysocki
Next step fusion devices that will operate in steady-state will require complex plasma-facing components (PFCs) that can survive the harsh environment over long timescales not common in current devices. This will require robust plasma facing surfaces that are integrated with active cooling systems. In a collaboration between CEA and ORNL, a plasma-interacting diagnostic is being designed for the W
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Structural assessment of the contamination protection structure placed on the top of the DEMO bioshield roof Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-01 P. Marek, M. Sienkiewicz, Ch. Bachmann, Ł. Ciupiński
In the DEMO power plant design, the bioshield roof above the tokamak is a steel structure with radial beams and concrete inserts. A contamination protection structure (CPS) is placed on the bioshield roof to protect other areas against the potential release of any radioactive materials from inside the reactor during maintenance. It also provides support to the cask, which is used to transfer the heavy
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Overview of progress on water cooled ceramic breeder blanket in Japan Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-01 Yoshinori Kawamura, Takanori Hirose, Wenhai Guan, Yuya Miyoshi, Takuya Katagiri, Atsushi Wakasa, Yoji Someya, Kentaro Ochiai, Hiroyasu Uto, Jae-hwan Kim, Yuki Koga, Motoki Nakajima, Takashi Nozawa, Yoshiteru Sakamoto, Hiroyasu Tanigawa, Takumi Hayashi
National institutes for Quantum Science and Technology (QST) are the implementing body of ITER project in Japan and takes the lead in fusion DEMO reactor development in Japan. The primary candidate of the breeder blanket for JA DEMO is Water Cooled Ceramic Breeder (WCCB) blanket. ITER-Test Blanket Module (TBM) program is an opportunity for the demonstration of DEMO breeding blanket concept under actual
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Maintenance and optimization of the TCV power supply Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-01 U. Siravo, J. Dubray, H. Elaian, D. Fasel, D. Velasco
The TCV tokamak is powered by a flywheel generator to supply the magnetic coils and the auxiliary heating systems. The generator has just undergone its fourth major overhaul to make it ready for the next ten years, after more than thirty years of almost trouble-free operation.
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Concept for HCPB TER using non-evaporable getters for tritium recovery Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-01 George Ana, Ovidiu Balteanu, Ion Cristescu, Radu Ana
The reference technology for DEMO Helium Cooled Pebble Bed breeding blanket for the tritium extraction is based on tritium release in a He purge gas at 0.2 MPa, followed by a 2-stage tritium recovery process from the purge gas in the Tritium Extraction and Recovery system. The recovery process is based on adsorption of QO on reactive molecular sieve beds, while Q is trapped using cryogenic molecular
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Commissioning and initial operation of HL-3 vacuum system Fusion Eng. Des. (IF 1.7) Pub Date : 2024-03-01 Chengzhi Cao, Yanfeng Xie, Xiangmei Huang, Yi Hu, Hong Ran, Chenghe Cui, Jun Zhou, Xiaoyan Gao, Xiao Cai, Xiaoquan Ji, Zeng Cao, Wulyu Zhong, Min Xu, HL-3 team
HL-3 is a medium-sized tokamak and aims at developing the high performance plasma and engineering toward ITER and the fusion reactor. As the critical system of HL-3, vacuum system consists of ultra-high vacuum pumping system, wall conditioning system, gas injection system and auxiliary system. Based on the characteristics of these systems, engineering commissioning had been performed for functional
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Control of electrochemical parameters for hydrogen permeation of CLF-1 steel Fusion Eng. Des. (IF 1.7) Pub Date : 2024-02-29 Fantao Meng, Long Wang, Zhihao Hong, Pinghuai Wang, Chunhai Liu, Zikai Liu, Qixiang Cao, Baoping Gong, Youzhi Wang, Long Zhang, Yuxiang Zhao
The electrochemical hydrogen permeation method is reliable for evaluating the intrinsic hydrogen permeation resistance due to its room-temperature operation. However, the control of electrochemical parameters for hydrogen permeation testing has rarely been reported. In this study, we discussed the effects of anode background current, cathode current density, and cycle times on the electrochemical permeation