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  • The development, testing and comparison of unfolding methods in SPECTRA-UF for neutron spectrometry
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-19
    R. Worrall; B. Colling; M.R. Gilbert; E. Litherland-Smith; C.R. Nobs; L.W. Packer; C. Wilson; A. Zohar

    Spectrum unfolding is a key tool used together with diagnostics in the determination of nuclear fields that are associated with a range of nuclear technologies spanning fusion, fission, nuclear medicine and accelerator domains. The underlying process requires a mathematical method for solving the Fredholm integral equation of the first kind. This paper discusses the development, testing and comparison

    更新日期:2020-10-19
  • Thermal neutron measurement by single crystal CVD diamond detector applied with the pulse shape discrimination during deuterium plasma experiment in LHD
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-19
    Makoto I. Kobayashi; Maurizio Angelone; Sachiko Yoshihashi; Kunihiro Ogawa; Mitsutaka Isobe; Takeo Nishitani; Siriyaporn Sangaroon; Shuji Kamio; Yutaka Fujiwara; Tomomi Tsubouchi; Akira Uritani; Minoru Sakama; Masaki Osakabe

    The Pulse shape discrimination (PSD) technique for single crystal CVD diamond detector (SDD) was applied to the real-time thermal neutron measurement during the 3rd campaign of Deuterium-Deuterium (D-D) plasma experiment in the Large Helical Device (LHD). The PSD method is based upon the different shape of electrical pulses produced in diamond by gamma-ray (triangular-shaped pulse) and energetic ions

    更新日期:2020-10-19
  • Preliminary insulation design for transmission line of CFETR NNBI test platform
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-19
    ChunDong Hu; RiXin Wang; CaiChao Jiang; YongJian Xu; YaHong Xie; YuanLai Xie

    In order to satisfy the requirements of China Fusion Engineering Test Reactor (CFETR) for neutral beam injector system, a Negative-ion Based Neutral Beam Injector system (NNBI) test platform will be established by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP). The high voltage Transmission Line (TL) plays an important role in the NNBI system, connecting the ion source and the power

    更新日期:2020-10-19
  • On the impact of the heat transfer modelling approach on the prediction of EU-DEMO WCLL breeding blanket thermal performances
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-19
    Francesco Edemetti; Emanuela Martelli; Alessandro Del Nevo; Fabio Giannetti; Pietro Arena; Ruggero Forte; Pietro Alessandro Di Maio; Gianfranco Caruso

    The Water-Cooled Lithium-Lead Breeding Blanket is a key component of a fusion power plant, in charge of ensure Tritium production, shield Vacuum Vessel and magnets and remove the heat power deposited by radiation and particles arising from plasma. The last function is fulfilled by First Wall and Breeding Zone independent cooling systems. Several layouts of BZ coolant system have been investigated in

    更新日期:2020-10-19
  • In-situ measurement of surface modifications of tungsten exposed to pulsed high heat flux for divertor design in tokamak-type fusion nuclear reactors
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-16
    Yuki Matsuda; Shohei Yamashita; Yusei Miyamoto; Daichi Motoi; Takafumi Okita; Eiji Hoashi; Kenzo Ibano; Yoshio Ueda

    A tungsten (W) divertor is exposed to pulsed high heat caused by disruptions and Type-I Edge Localized Modes (ELMs). Therefore, it is important to investigate surface morphology of W during phase change. Heretofore, W melting dynamics under disruption-like heat load has been researched and tried to modify the optical system of our apparatus for knowing more detailed surface behavior. In this paper

    更新日期:2020-10-17
  • Negative ion beam extraction in volume mode on the RF negative ion source at ASIPP
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-14
    Qinglong Cui; Jianglong Wei; Qi Wang; Junjun Pan; Shiyong Chen; Yahong Xie; Caichao Jiang; Yuanzhe Zhao; Yongjian Xu; Yuanlai Xie; Chundong Hu

    Hefei utility negative ions test equipment with RF source (HUNTER) has been developed at the Institute of Plasma Physics Chinese Academy of Sciences (ASIPP), in order to study the RF negative ion source for neutral beam injection application. This paper reports on several recent modifications made to HUNTER, including magnetic confining field, magnetic filter field, extraction area, pumping system

    更新日期:2020-10-15
  • FEM analyses of the ITER EC H&CD torus diamond window unit towards the prototyping activity
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-14
    G. Aiello; N. Casal; P. Estébanez; V. Gràcia; M. Henderson; A. Meier; G. Saibene; T. Scherer; S. Schreck; D. Strauss; A. Vaccaro

    The chemical vapour deposition (CVD) diamond torus window unit is a sub-component of the ITER Electron Cyclotron Heating and Current Drive (EC H&CD) system used for a diverse range of applications including plasma heating and control of plasma magneto-hydrodynamic (MHD) instabilities. It consists of an ultra-low loss polycrystalline diamond disk brazed to copper cuffs and then enclosed by a metallic

    更新日期:2020-10-15
  • Irradiation tests of bolometer sensor prototypes for ITER
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-12
    H. Meister; S. Schmitt; I. Szenthe; A. Szakál; H. Albrecht; F. Gillemot

    In ITER, the bolometer diagnostic is foreseen to provide the measurements for the total plasma radiation. Sensors need to withstand the harsh environmental conditions. Most prominent among those is the nuclear environment with neutron fluxes up to 1013ncm2s at the locations of bolometers, which result in a radiation dose of up to 0.3dpa in Si3N4. Original metal resistor bolometer sensors based on Au-absorbers

    更新日期:2020-10-13
  • Development of tritium dynamic transport analysis tool for tritium breeding blanket system using Modelica
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-12
    Ruyan Li; Xiaoyu Wang; Long Zhang; Jun Wang

    Demonstration of the engineering feasibility of tritium breeding, including tritium generation, tritium extraction, tritium control and tritium safety is one of the main objectives of ITER Testing Blanket Module (TBM) Program. As one of TBM concepts, the tritium transport assessment of the China Helium Cooled Ceramic Breeder TBM and its ancillary systems (called test blanket system, TBS) is absolutely

    更新日期:2020-10-13
  • SPIDER plasma grid masking for reducing gas conductance and pressure in the vacuum vessel
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-09
    M. Pavei; S. Dal Bello; G. Gambetta; A. Maistrello; D. Marcuzzi; A. Pimazzoni; E. Sartori; G. Serianni; F. Degli Agostini; L. Franchin; M. Tollin

    SPIDER experiment is operating at the PRIMA site in Padova (I) since June 2018, with the aim of testing and optimizing the negative ion source prototype for ITER Heating Neutral Beam Injectors. In the first operational phase it was discovered that, as the in-vessel hydrogen pressure exceeds the design requirements, discharges occur on the back of the radio frequency source. A specific operational campaign

    更新日期:2020-10-11
  • Conceptual design of Doppler shift spectroscopy diagnostics for INTF
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-08
    A.J. Deka; Bharathi P.; M. Bandyopadhay; M.J. Singh; A.K. Chakraborty

    INTF (Indian Test Facility) is a negative hydrogen ion source based neutral beam test facility to characterise 100 keV,60 A Diagnostic Neutral Beam (DNB), conceptualised for ITER. Several diagnostics are under development to monitor and characterise the performance of ion source and the beam properties. Doppler Shift Spectroscopy (DSS) diagnostics shall be used to estimate the beam divergence, beam

    更新日期:2020-10-08
  • Preliminary investigation of the shutdown radioactivity for EAST based on gamma ray spectrometry
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-10-07
    Ruixue Zhang; Liqun Hu; Kai Li; Guoqiang Zhong; Ruijie Zhou; Bing Hong; Mengjie Zhou; Min Xiao; Liangsheng Huang

    Experimental Advanced Superconducting Tokamak (EAST) is capable of operating with long-pulse deuterium plasma, which will generate a lot of radiation such as fusion neutrons, gamma rays and hard X-rays. Since the structure material of the device was activated, work around and inside the tokamak would be limited from a radiation protection point of view after shutdown. Identification of radionuclides

    更新日期:2020-10-08
  • Shielding analysis of the ITER Collective Thomson Scattering system
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-30
    A. Lopes; R. Luís; E. Klinkby; Y. Nietiadi; A. Chambon; E. Nonbøl; B. Gonçalves; M. Jessen; S.B. Korsholm; A.W. Larsen; B. Lauritzen; J. Rasmussen; M. Salewski

    The Collective Thomson Scattering (CTS) system will be the ITER diagnostic obtaining the plasma fast alpha-particle velocity distribution and will be implemented in drawer #3 of the Equatorial Port Plug #12 of the reactor. In this work, a neutronics analysis is presented for the in-vessel front-end parts of the CTS system, including neutron and gamma-ray fluxes and nuclear heat loads for the main components

    更新日期:2020-09-30
  • Deformation heterogeneity in laser-welded Eurofer
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-26
    Allan Harte; Huw Dawson; David Bowden; Rory Spencer; Simon Kirk; Michael Gorley
    更新日期:2020-09-26
  • On the standards and practices for miniaturized tensile test – A review
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-23
    Pengfei Zheng; Ran Chen; Haiting Liu; Jiming Chen; Zhijie Zhang; Xing Liu; Yao Shen

    Miniaturized Tensile Test (MTT), as an important type of Small Specimen Test Technology (SSTT), has gained increasing attentions in the nuclear industry and other fields. However, MTT faces great challenges because the test results may deviate from standard specimens due to specimen size effects and may exhibit large scatters caused by fabrication imperfection. In order to improve the reliability of

    更新日期:2020-09-23
  • Extraction and separation of lithium isotopes by using organic liquid film extraction system of crown ether-ionic liquid
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-21
    Zezheng Zhang; Yongzhong Jia; Bing Liu; Huaxin Sun; Yan Jing; Quanyou Zhang; Fei Shao; Mixiang Qi; Ying Yao

    The organic liquid film extraction system, a novel extraction system, was investigated for the extraction and separation of lithium isotopes using ionic liquid as the co-extractant and Dibenzo-15-crown 5-Ether (DB15C5) as the extractant. Multiple subjects were studied during the investigation, including the effects of extraction time; gas flow rate; temperature; counter anions of lithium salt; the

    更新日期:2020-09-21
  • Influence of Lithium Mass Transfer on Tritium Behavior in Pebbles of Li2TiO3 with Excess Lithium
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-21
    Kazunari Katayama; Akito Ipponsugi; Tsuyoshi Hoshino

    Tritium breeding ceramic materials are placed at high temperatures for a long period in a fusion DEMO reactor. Therefore, the understanding of Li mass loss phenomena and its influence on tritium behavior are important. In this study, the pebbles of Li2TiO3 with excess Li were heated at 900 °C for 30 days in a 1000 Pa H2/Ar flow and tritium sorption and recovery experiments were carried out. Li mass

    更新日期:2020-09-21
  • Mechanical properties of degraded OFE copper subjected to electron irradiation
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-19
    S.R. Ghodke; B.K. Dutta; P.V. Durgaprasad; N.N. Kumar

    The mechanical and fracture properties of aged materials degrade due to metallurgical and micro- structural changes during service life. Miniaturized specimen testing methodology has been evolved over the years to quantify such degradation mechanisms. Oxygen Free Electronic (OFE) Copper has wide applications in nuclear industry, particularly in fusion reactors and accelerators. Such material when subjected

    更新日期:2020-09-20
  • Shutdown dose rate studies for the DTE2 campaign at JET
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-18
    N. Fonnesu; R. Villari; D. Flammini; P. Batistoni; U. Fischer; P. Pereslavtsev

    The EUROfusion Work Package JET3 programme, established to enable the technological exploitation of the future Deuterium-Tritium (DT) operations at JET over the next years, includes, within the NEXP subproject, a novel Shutdown Dose Rate (SDR) benchmark experiment. The measurement of the SDR due to neutron activation in a fusion machine operating with Deuterium and Tritium is of primary importance

    更新日期:2020-09-20
  • Development of an on-line sensor for hydrogen isotopes monitoring in flowing lithium at DONES
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-18
    B. Garcinuño; D. Rapisarda; F.R. Urgorri; J. Herranz; M. Malo; J. Mollá; A. Ibarra

    DONES (Demo Oriented Neutron Source) consists of an irradiation facility based in the nuclear stripping reactions that occur when one energetic deuteron beam impinges on a liquid lithium target flowing along an open channel. Since impurities enhance the erosion/corrosion effect, one of the challenging aspects on the development of the neutron source is the purity of the lithium. One group of impurities

    更新日期:2020-09-20
  • Activation analysis of the European DEMO divertor with respect to the different breeding blanket segmentation
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-20
    Andrius Tidikas; Gediminas Stankunas

    Neutron activation is an unavoidable process in large scale fusion devices operating with deuterium and tritium fuels. Neutron activation leads to the production of radioactive materials. Activated materials can generate heat and ionizing radiation in their environment. Consequently, values of activation characteristics need to be determined in order to ensure safety and performance of the fusion devices

    更新日期:2020-09-20
  • Test Results of Active Thermography Method for Plasma-Wall Interaction Studies on the КTM Tokamak
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-17
    E. Batyrbekov; B. Chektybayev; A. Sadykov; M. Skakov; E. Kashikbayev; D. Olkhovik; S. Zhunisbek

    The paper describes the main experimental results of testing the method of thermographic measurements on a test bench with a plasma-beam facility. The developed method is intended to improve the accuracy of surface temperature measurements of candidate materials of the first wall of future thermonuclear reactors during investigations on the KTM tokamak. The method is based on the use of an external

    更新日期:2020-09-18
  • Highly stable signal generation in microwave interferometer using PLLs
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-16
    Jitendra P Chaudhari; Bhargav Patel; Amit V Patel; Alpesh D Vala; Keyur K Mahant; Hiren K Mewada; Abhishek Sinha; S K Pathak

    The microwave interferometer is a device that works in the millimeter-wave frequency range, and it is used to measure the plasma density. The instability in the frequency source at the transmitter section of the interferometer affects the accuracy of the phase measurement in the receiver section. The phase-locked loop (PLL) circuit is a substantial unit in the generation of highly stable frequency

    更新日期:2020-09-16
  • Oxidation behavior analysis of a ferritic ODS steel in supercritical Water
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-16
    Qian Zhao; Zhixia Qiao; Ji Dong; Liming Yu; Chenxi Liu; Huaiwen Wang

    The corrosion behavior of 14Cr ODS steel exposed to the supercritical water (SCW) at 600 °C under 25 MPa with the dissolved oxygen content of 300 ppb was investigated. The weight gain measurement and corrosion layers analysis were used to determine the corrosion behavior and corrosion mechanism of the 14Cr ODS steel. The results indicate that the 14Cr ODS steel exposed in this work possesses better

    更新日期:2020-09-16
  • Commissioning and initial operation of the W7-X neutral beam injection heating system
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-16
    P. McNeely; S. Äkäslompolo; W. Auerweck; Y. Drider; O.P. Ford; D.A. Hartmann; B. Heinemann; S. Heinrich; C. Hopf; R. Kairys; S. Obermayer; R. Riedl; P. Rong; N. Rust; R. Schroeder; R.C. Wolf

    The first, of two planned, neutral beam injectors for the stellarator Wendelstein 7-X (W7-X) was commissioned for and participated in the experimental campaign (OP1.2b) from July to October 2018. The injector was equipped with two RF driven ion sources from which 90A of positive hydrogen ions were extracted at 55 kV. After neutralization, the two sources provided >3 MW of neutral beam heating power

    更新日期:2020-09-16
  • Safety analysis of the DONES primary heat removal system
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-11
    Pietro Arena; Pietro Alessandro Di Maio; Francesco Saverio Nitti

    The development of a neutron source able to reproduce the irradiation conditions typical of a nuclear fusion reactor, in order to test candidate structural materials, is the main goal of the Work Package Early Neutron Source (WPENS) of the EUROfusion action. This source, named Demo Oriented NEutron Source (DONES), is a facility where neutrons are produced by means of D-Li interactions. More in detail

    更新日期:2020-09-12
  • Study of lattice thermal conductivity of tungsten containing bubbles by molecular dynamics simulation
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-11
    Hongyu Zhang; Jizhong Sun; Yingmin Wang; Thomas Stirner; Ali Y. Hamid; Chaofeng Sang

    Exposed to high fluxes of helium/hydrogen isotope particles and heat, tungsten divertor plates will suffer damage thus degrading its performance such as its thermal conductivity. This paper presents a study on the effect of bubbles on the lattice thermal conductivity of tungsten at the atomic level using molecular dynamics simulations. The present study finds that empty bubbles in tungsten lead to

    更新日期:2020-09-11
  • Inhibitory effect of dislocation on helium irradiation induced damage in Fe-9 wt.Cr alloy
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-10
    Zhaokuan Zhang; Te Zhu; Shuoxue Jin; Peng Zhang; Peng Kuang; Liping Guo; Xudong An; Baoyi Wang; Runsheng Yu; Shasha Zhang; Xingzhong Cao

    Suppression of transmutation helium can be achieved by introducing trapping sites such as dislocation and grain boundary in nuclear materials. In order to understand role of dislocations in helium irradiation effect in metal, positron annihilation spectroscopy and thermal desorption measurements were performed to study the interaction of helium atoms with dislocation in deformed Fe-9 wt.Cr alloys irradiated

    更新日期:2020-09-10
  • Weighted majority rule ensemble classifier for sensor fault classification for plasma position control in Tokamaks
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-09
    Debashish Mohapatra, Bidyadhar Subudhi

    Tokamaks are the most promising devices which employ magnetic confinement of Deuterium and Tritium plasma to achieve nuclear fusion for power generation. Magnetic flux sensors are employed to measure the position of the plasma column inside the device. Faults occurring in these sensors may cause the failure of control of the plasma position, thereby terminating the fusion process. In this paper, we

    更新日期:2020-09-09
  • Rapid prototyping of advanced control schemes in ASDEX Upgrade
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-09
    B. Sieglin, M. Maraschek, O. Kudlacek, A. Gude, W. Treutterer, M. Kölbl, A. Lenz

    The integration of advanced control schemes is becoming more important as the development of fusion experiments progresses. The ASDEX Upgrade discharge control system (DCS) is designed to be adaptable via configuration, no recompilation is necessary to tailor the behaviour of the control system. In order to enable advanced control schemes the required information has to be available during the discharge

    更新日期:2020-09-09
  • Design and preliminary analyses of the new Water Cooled Lithium Lead TBM for ITER
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-08
    J. Aubert, G. Aiello, D. Alonso, T. Batal, R. Boullon, S. Burles, B. Cantone, F. Cismondi, A. Del Nevo, L. Maqueda, A. Morin, E. Rodríguez, F. Rueda, M. Soldaini, J. Vallory

    In the European strategy, DEMO is the intermediate step between ITER and a commercial fusion power plant. In this framework, one of the goal of DEMO is to be a Breeding Blanket test facility. The Breeding Blanket, which is not present in ITER, is one of the key components for the future deployment of nuclear fusion electricity as it accomplishes the functions of tritium breeding and nuclear to thermal

    更新日期:2020-09-08
  • Design of the European DEMO vacuum vessel inboard wall
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-08
    Rocco Mozzillo, Christian Bachmann, Giacomo Aiello, Domenico Marzullo

    The pre-concept design of the DEMO Vacuum Vessel is going on in view of the 2020 gate review; moreover the nuclear heat loads on the vessel inner shell were determined and found to be about one order of magnitude higher compared to ITER. A subsequent thermal-structural analysis of the vessel inner shell revealed high thermal stresses and a large temperature gradient through the inner shell thickness

    更新日期:2020-09-08
  • TFC-PREDIM: A FE dimensioning procedure for the TF coil system of a DEMO tokamak reactor
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-08
    Ilia Ivashov, Wolfgang Biel, Philippe Mertens

    The equatorial plane of the inner leg of a toroidal field (TF) coil is the most stressed part of the TF coil system and optimal usage of the radial space in this region is crucial for the design of the DEMO tokamak reactor. A procedure for initial dimensioning (pre-dimensioning) of this region developed earlier [1] is based on a simplified 2D geometry of the TF coil cross-section and a semi-analytical

    更新日期:2020-09-08
  • Three-dimensional validation and analyses of the optimized CFETR HCCB blanket
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-07
    Shijie Cui, Yueheng Lang, Dalin Zhang, Xinyu Jiang, Haoyu Wan, Wenxi Tian, G.H. Su, Xiang Gao

    The Helium Cooled Ceramic Breeder (HCCB) blanket is the critical component of the China Fusion Engineering Test Reactor (CFETR). The radial structural layout of the internal functional zones has the largest impact on the neutronics and thermal-hydraulic performance of the blanket, and it’s also the major determinant for the detailed 3D design. In the previous work, NTCOC, a Neutronics/Thermal-hydraulic

    更新日期:2020-09-07
  • Safety system of W7-X neutral beam injection heating system
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-07
    R. Schroeder, Y. Drider, D.A. Hartmann, B. Heinemann, R. Kairys, P. McNeely, R. Riedl

    In the last experimental campaign (OP1.2b: 7/2018 to 10/2018) of the Wendelstein 7-X stellarator (W7-X), the first of two neutral beam injectors (NBI) was commissioned with 2 plasma sources. The safe operation of the NBI requires the exclusion of hazards to personnel and minimization of system hazards. The NBIcontrol system consists of a standard PLC based on PCS7 from Siemens AG and a separate safety

    更新日期:2020-09-07
  • Control pathway for an advanced divertor on ITER
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-06
    J.T. Wai, P.J. Vail, E. Kolemen

    This paper presents the development of a coupled shape and divertor controller for ITER with capabilities to control the flux expansion and an advanced divertor configuration, the x-divertor (XD), in which a secondary x-point is placed in the downstream scrape-off layer. Due to the high-performance nature of ITER and its relatively few shaping coils, satisfying constraints on the coil currents, power

    更新日期:2020-09-06
  • TOKES simulations of mitigated disruption thermal quenches in ITER
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-06
    S. Pestchanyi, M. Lehnen, R.A. Pitts, G. Saibene

    TOKES simulations of mixed Ne/D2 pellet injection through 1 equatorial launcher into an ITER discharge of 280 MJ thermal plasma energy was carried out. Two kinds of pellets were simulated: ‘large’, containing 1.1∙1024 D2 molecules and ‘small’ with 2.6∙1023 D2 molecules, both with various amount of Ne and correspondingly lower quantities of D to preserve the pellet size. It was found that the physics

    更新日期:2020-09-06
  • Real-time applications of Electron Cyclotron Emission interferometry for disruption avoidance during the plasma current ramp-up phase at JET
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-04
    M. Fontana, C.D. Challis, N.J. Conway, R. Felton, A. Goodyear, C. Hogben, A. Peacock, S. Schmuck

    Data produced by an electron cyclotron interferometer diagnostic are now available to the real-time control systems of of the joint European torus (JET) tokamak. The data consist of absolutely calibrated electron temperature profiles, covering the plasma low-field side, core and part of the high-field side, for most of the range of magnetic fields used at JET. The profiles are obtained without the

    更新日期:2020-09-04
  • Design and verification of a non-self-supported cryostat for the DEMO tokamak
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-03
    Ł. Ciupiński, T. Zagrajek, P. Marek, G. Krzesiński, C. Bachmann, R. Mozzillo

    The paper presents a conceptual design and its structural verification of the cryostat for the DEMOnstration Fusion Reactor (DEMO). The cryostat is a large pressure vessel providing the vacuum required to operate the superconducting coils at cryogenic temperatures. Cryostats of existing machines typically are cylindrical and self-support the external pressure. In a nuclear machine, like DEMO, a massive

    更新日期:2020-09-03
  • The flow instability phenomenon in the loss of coolant accident of the water cooled blanket
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-02
    Dianle Wang, Changhong Peng, Yun Guo

    In the current design, the blanket plays an important role of transporting heat and breeding Tritium. It is very necessary to analyze the accident of the blanket. In this study, the RELAP5 model of the Water Cooled Ceramic Breeder blanket (WCCB) was established. The numerical simulation of the in-vessel loss of coolant accident (LOCA) on the first wall (FW) were conducted. The instability phenomenon

    更新日期:2020-09-02
  • Initial operation results of exhaust detritiation system using a polymer membrane
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-02
    Masahiro Tanaka, Naoyuki Suzuki, Hiromi Kato, Minoru Yokosawa

    An exhaust detritiation system using commercially available hollow fiber-type polyimide membrane modules, a PM system, was installed and applied for tritium recovery in the vacuum vessel purge gas of a large fusion test device. The PM system is operated annually and there were no serious malfunctions after starting the operation for approximately 2.5 y. The continuous recovery of tritiated water without

    更新日期:2020-09-02
  • Design of multipactor-suppressed high-power VFT for helicon current drive in KSTAR
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-02
    K.H. Jang, S.J. Wang, H.H. Wi, K. Saito, J.H. Kim, H.Y. Lee, J.G. Kwak

    A helicon wave was tested for coupling at low power in the KSTAR, and high-power coupling has also been tried. When RF is applied to the antenna system, the reflected power gradually increases on a sub-millisecond timescale. A slower process than usual arcing suggests that a multipactor discharge causes the reflection. It is essential to mitigate or eliminate the multipactor discharge to apply a high-power

    更新日期:2020-09-02
  • Autonomous pulse shaping method for inertial confinement fusion high power laser facility
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-02
    Xiaoxia Huang, Xuewei Deng, Wei Zhou, Huaiwen Guo, Bowang Zhao, Bo Zhang, Xiaocheng Tian, Wei Zhong, Wu Deng

    Laser pulse shape is one of the most critical parameters for the success of inertial confinement fusion (ICF) experiments. The ability to shape the laser pulse with accuracy, efficiency, and robustness is fundamental to the ICF experiments. An autonomous pulse-shaping method that is universally feasible for ICF facilities with individual characteristics and independent adjustability for each beam has

    更新日期:2020-09-02
  • Tritium distribution analysis of Be limiter tiles from JET-ITER like wall campaigns using imaging plate technique and β-ray induced X-ray spectrometry
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-01
    S.E. Lee, Y. Hatano, M. Hara, S. Masuzaki, M. Tokitani, M. Oyaizu, H. Kurotaki, D. Hamaguchi, H. Nakamura, N. Asakura, Y. Oya, J. Likonen, A. Widdowson, S. Jachmich, K. Helariutta, M. Rubel

    Tritium (T) distribution on the plasma-facing surfaces (PFSs) and inside castellation of Be limiter tiles from the JET tokamak with the ITER-like wall (ILW) was analyzed using imaging plate (IP) technique and β-ray induced X-ray spectrometry (BIXS). Regarding to PFSs, the outer poloidal limiter (OPL) showed significantly higher T concentrations than the inner wall guard limiter (IWGL) and upper dump

    更新日期:2020-09-01
  • Understanding and investigating the relationships between geometrical errors and lost particles in MCNP
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-09-01
    Davide Laghi, Marco Fabbri, Raul Pampin, Alfredo Portone

    Generating Constructive Solid Geometry (CSG) ready for MCNP input using automated programs, and the use of universe structures, may give raise to geometrical inconsistencies leading to numerical phenomena known as lost particles, which perturb the statistical reliability of the transport solution. In the context of the reference nuclear analysis models developed for ITER, there is a need to better

    更新日期:2020-09-01
  • Composition optimization of high strength and ductility ODS alloy based on machine learning
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-31
    Bing Bai, Xu Han, Quan Zheng, Lixia Jia, Changyi Zhang, Wen Yang

    To find out characteristic parameters, mine data and predict candidate materials by machine learning is an important way for rapid development of materials. Based on about 300 groups data of composition, process, test condition and mechanical properties of ODS alloy, the correlation between the key component and the ultimate tensile strength and elongation of ODS alloy is established by deep learning

    更新日期:2020-08-31
  • Assessment of potential heat flux overload of target and first wall components in Wendelstein 7-X finite-beta magnetic configurations and choice of locations for temperature monitoring
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-30
    G. Bongiovì, D. Böckenhoff, A. Carls, M. Endler, J. Fellinger, J. Geiger

    Within the framework of R&D activities on the Wendelstein 7-X (W7-X) stellarator machine, the assessment of heat loads onto the plasma facing components (PFCs) is an important aspect. So far, W7-X was operated in short pulses without water cooling of the PFCs. Presently, the device is being prepared for future operation phases with water cooling. The target plates, which receive the highest heat loads

    更新日期:2020-08-30
  • Development of high strength and high electrical conductivity Cu-Cr-Zr alloy through friction stir processing
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-30
    R. Bheekya Naik, K. Venkateswara Reddy, G. Madhusudhan Reddy, R. Arockia Kumar

    Friction stir processing (FSP), a solid-state materials processing technique, is an effective method to modify the surface properties of metallic materials. In the present study, an effort has been made to improve the hardness and wear resistance of Cu–0.62 %Cr–0.11 %Zr alloy through FSP. Single-pass friction stirring was carried out by fixing the tool rotational speed as 600 rpm and by varying the

    更新日期:2020-08-30
  • Tritium transport analysis for tokamak exhaust processing system of tritium plant
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-29
    Qin Zeng, Wei Shi, Xiande Wang, Hongli Chen

    As an important part of tritium plant, tokamak exhaust processing (TEP) system is used for extracting hydrogen isotopes from exhaust gas of tokamak which contains high concentration of tritium and deuterium. A detailed tritium transportation analysis for TEP system can help to obtain more precise and accurate results of tritium behavior in the tritium plant. In this paper, a detailed model of TEP system

    更新日期:2020-08-29
  • Modeling and analysis of tritium transport for updated WCCB blanket of CFETR
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-28
    Xueli Zhao, Lei Chen, Xuebin Ma, Songlin Liu

    Radioactivity source item assessment has become an essential part of safety evaluations for the China Fusion Engineering Test Reactor. Thus, numerical simulations of tritium transport in a water-cooled ceramic breeder blanket are performed in this study. 2D and 3D multi-physics convection-diffusion models that couple velocity and temperature fields are built to estimate tritium permeation and other

    更新日期:2020-08-28
  • The implementation and operation of the 4th version of KSTAR fast interlock system
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-28
    Myungkyu Kim, Taehyun Tak, Jaesic Hong

    In order to increase the machine protection of KSTAR, it is more important than ever to generate an event to turn off the related heating devices even when the abnormal situation occurs during the plasma discharge. KSTAR has built the Fast Interlock System (FIS) since achieving first plasma in 2008, and built and operated the 4th version of FIS which using NI c-RIO technology in 2018 [1]. We moved

    更新日期:2020-08-28
  • Seismic analysis of the CFETR CS Model Coil
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-28
    Fan Wu, Xiaogang Liu, Zhaoliang Wang, Yong Ren, Junjun Li, Jie Zhang, Yu Wu, Xiang Gao

    The China Fusion Engineering Test Reactor (CFETR) Central Solenoid (CS) Model Coil is being fabricated by the Institute of Plasma Physics Chinese Academy of Sciences, and will be cryogenically tested in 2021. In order to validate the structural integrity of the CS Model Coil under seismic loads, response spectrum (RS) and time history analyses have been carried out. This paper first introduces the

    更新日期:2020-08-28
  • Plasma characteristics of a tapered coaxial gun and its damage effect on tungsten target
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-28
    Chongxiao Zhao, Jian Song, Liangwen Qi, Chunyu Ma, Jinjuan Hu, Xiaodong Bai, Dezhen Wang

    Due to the property of high energy density, temperature and density, the plasma prodced by the coaxial gun is usually untilized to simulate the Type- I ELM in tokamak. In this work, the plasma characteristics of a tapered coaxial gun was studied and its effect on the tungsten target was investigated. By using current probe, photodetector, spectrometer and calorimeter, it is found that the plasma energy

    更新日期:2020-08-28
  • Thermal-hydraulic analysis of the DEMO WCLL elementary cell: BZ tubes layout optimization
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-26
    Francesco Edemetti, Ivan Di Piazza, Alessandro Del Nevo, Gianfranco Caruso

    The Water-Cooled Lithium-Lead (WCLL) Breeding Blanket (BB) is a key component in charge of ensuring Tritium production, shield the Vacuum Vessel and remove the heat generated by plasma thermal radiation and nuclear reactions. It relies on PbLi eutectic alloy adopted as breeder and neutron multiplier and refrigerate by subcooled pressurized water. The last function is fulfilled by two independent cooling

    更新日期:2020-08-26
  • An intelligent controller design based on the neuroendocrine algorithm for the plasma density control system on Tokamak devices
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-26
    Shuangbao Shu, Chenyao Xing, Zhipeng Chen, Ming Zhang, Jiarong Luo, Yuzhong Zhang, Xiaojie Tao, Zhixue Cui, Feng Ji, Qiaosheng Pan, Chao Liu

    Realization of real-time and accurate control of plasma electron density is one of the key points for the long-time steady-state operation of Tokamak. Based on the mechanism model of the neuroendocrine regulatory principle, this paper designs an intelligent controller and studies its application on the plasma density control system (PDCS) for the J-TEXT Tokamak. This PDCS uses the HCN laser interferometer

    更新日期:2020-08-26
  • Design and implementation of a mobile parallel robot for assembling and machining the fusion reactor vacuum vessel
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-25
    Changyang Li, Huapeng Wu, Harri Eskelinen, Heikki Handroos, Ming Li

    The present paper introduces a mobile robot machine for the fusion reactor and presents its implementation. The task of the robot is to carry out assembly work inside the fusion reactor vacuum vessel, the assembly work consists of scanning, machining, splice plate transportation, welding, nondestructive testing (NDT), defect welding point cutting and re welding. To better meet the requirements of the

    更新日期:2020-08-25
  • Magnetic shaping for a minimum B trap
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-24
    O. Ågren, V.E. Moiseenko

    A mirror machine with a minimum B field for MHD stability is an option for a steady-state compact fusion neutron source design. A previous calculation of an idealized quadrupolar mirror field in Ref Ågren and Moiseenko (2017) is extended to a magnetic field which smoothly evolves to expander regions beyond the mirror throats. In a minimum B field, the projection of a flux surface on planes perpendicular

    更新日期:2020-08-24
  • Emissivity measurement of PbLi droplets in a vacuum for heat recovery by radiation
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-23
    Fumito Okino, Satoshi Konishi

    The emissivity of a liquid lithium-lead (PbLi) droplet falling in a vacuum is theoretically analyzed and verified by infrared thermal imaging. The authors propose the concept of simultaneous recovery of heat and tritium from falling PbLi droplets in a vacuum. Permeation reduction in a heat exchanger is expected to occur by radiation heat recovery in a vacuum. The radiation power depends on the droplet

    更新日期:2020-08-23
  • Aditya Upgradation – Equilibrium study
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-21
    Deepti Sharma, R. Srinivasan, Joydeep Ghosh, P. Chattopadhyay

    The ADITYA tokamak device is used to produce circular plasma for few hundreds of milli-seconds. The edge physics study in this device led to significant contributions. The up-gradation of this device (ADITYA-U) is focused to address issues relevant to heat removal capability at the plasma edge. This requires to construct plasma equilibrium with divertor configuration. In this regard, additional pair

    更新日期:2020-08-21
  • Backwards extrapolation activation diagnostics and their dynamic range for pulsed neutron source measurements
    Fusion Eng. Des. (IF 1.692) Pub Date : 2020-08-20
    L.W. Packer, S. Allan, S.C. Bradnam, S. Jednorog, E. Łaszyńska, N.J. Roberts, C. Wilson, R. Worrall

    Activation materials implanted within radiation detectors can be used to measure pulsed neutron fields. This work develops an instrument concept with the aim to maximise sensitivity to pulsed fusion neutron fields and, using a data-rejection algorithm combined with backwards extrapolation, enable neutron fluence estimates to be made over a large dynamic range. Through high-fidelity modelling of residual

    更新日期:2020-08-20
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