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  • Multi-channel analog lock-in system for real-time motional Stark effect measurements
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-29
    Hanmin Wi; Jinseok Ko; Jinil Chung

    The motional Stark estark effect (MSE) diagnostic system developed to measure the plasma current density distribution in the Korea Superconducting Tokamak Advanced Research performs data analysis by applying Fourier transform algorithm using the Interactive Data Language (IDL) software after measurement and digital archiving. However, in order to realize advanced plasma control aiming at the development

  • Overview about cryogenic distillation control and safety approach for Isotopes Separation Facility (ISF)
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-28
    Iuliana Stefan; Ovidiu Balteanu; Ciprian Bucur; Mihai Vijulie; Nicolae Sofilca; Carmen Moraru; Liviu Stefan; Anisia Bornea; Marius Zamfirache

    Isotopes separation technologies are used in both Tritium Removal Facilities associated with CANDU reactors (to remove tritium from heavy water) and fusion applications (eq. ITER Isotope Separation System) to recover and enrich tritium for further operation. The expertise gained during design, manufacturing, building, commissioning and operation of the Cryogenic Distillation System (CDS) with proprietary

  • Effect of thermal cycles on structure and deuterium permeation of Al2O3 coating prepared by MOD method
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-27
    Changda Zhu; Wei Zhang; Long Wang; Zhien Ning; Yongjin Feng; Kaiming Feng; Jiali Liao; Yuanyou Yang; Ning Liu; Jijun Yang

    The effect of thermal cycles on the microstructure and deuterium permeation of the Al2O3 coatings prepared by the MOD method was investigated. An as-deposited Al2O3 coating showed a uniform, smooth and dense surface morphology and was composed of γ-Al2O3 and α-Al2O3 phases. After 550 ℃ thermal cycles, all the Al2O3 coatings exhibited good phase stability. The Al2O3 coating remained intact and dense

  • Progress of manufacture technology of CFETR WCCB blanket at ASIPP
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-26
    Wanjing Wang; Jichao Wang; Chunyi Xie; Qiang Li; Sigui Qin; Shiwei Xu; Guohui Liu; Mingzhun Lei; Songlin Liu; Guang-Nan Luo

    As one candidate of blankets for China Fusion Engineering Test Reactor (CFETR), the Water Cooled Ceramic Breeder (WCCB) blanket is proposed and designed in Institute of Plasma Physics, Chinese Academy of Sciences (ASIPP). Supported by the National Key R&D Program of China, we formulated a roadmap on its manufacturing. In this work, the research and development (R&D) on bonding technology for W/Steel

  • RAMI evaluation of the beam source for the DEMO neutral beam injectors
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-26
    P. Agostinetti; T. Franke; U. Fantz; C. Hopf; N. Mantel; M.Q. Tran

    DEMO is a first-of-a-kind DEMOnstration fusion power plant [1], [2] and is intended to follow the ITER experimental reactor. The main goal of DEMO will be to demonstrate the possibility to produce electric energy for the grid from the fusion reaction early in the second half of the century. The injection of high energy neutral (1 MeV) particle beams is one of the main tools to heat the plasma up to

  • Use of virtual actuators in ASDEX Upgrade control
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-26
    O. Kudláček; W. Treutterer; B. Sieglin; F. Janky; F. Felici; I. Gomez-Ortiz; A. Gräter; T. Maceina; M. Maraschek; T. Zehetbauer

    Actuator management is an essential part of a modern tokamak plasma control system. It has to deal with a large number of control task simultaneously, needs to be able to operate close to stability limits and to avoid disruptions. In the ASDEX Upgrade tokamak experiment, the process of actuator management development is ongoing. As a first step, we have removed direct assignement of physical actuators

  • Design and Analysis of liquid nitrogen cooled sorption cryopump for SST-1 Tokamak
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-25
    Vishal Gupta; Ranjana Gangradey; Samiran S. Mukherjee; Jyoti Shankar Mishra; Pratik A. Nayak; Paresh Panchal; Vipul L. Tanna; Yuvakiran Paravastu; Dilip C. Raval; Ziauddin Khan; Siju George; Prashant L. Thankey

    SST-1 is a Steady State Superconducting Tokamak for a proposed plasma discharge of long duration. This plasma device has two vacuum vessels, the cryostat and the vacuum vessel (VV). Cryostat houses the superconducting magnet systems (TF and PF coils), LN2 cooled thermal shields and hydraulics for these circuits and the plasma will be confined inside the VV [1]. To have plasma discharges, ultra-high

  • Effect of Cr and V coatings on W base material in W-Eurofer brazed joints for fusion applications
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-25
    J. de Prado; M. Sánchez; G. Stan; A. Galatanu; A. Ureña

    Titanium diffusion in tungsten is an undesirable phenomenon that may cause the drop of mechanical and thermal fatigue properties of tungsten base material and components in future fusion reactors. To avoid such as problematic, the effectiveness of two different diffusion coatings, deposited onto W base materials by means of RF magnetron sputtering (Cr and V layers), has been studied to analyze its

  • Overview of recent ITER TBM Program activities
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-24
    Luciano M. Giancarli; Xavier Bravo; Seungyon Cho; Marco Ferrari; Takumi Hayashi; Byoung-Yoon Kim; Artur Leal-Pereira; Jean-Pierre Martins; Mario Merola; Romain Pascal; Iva Schneiderova; Qian Sheng; Amit Sircar; Yuri Strebkov; Jaap van der Laan; Alice Ying

    The ITER Test Blanket Module (TBM) Program has significantly evolved since 2018. The number of equatorial ports allocated for operating the Test Blanket Systems (TBSs) has been reduced from three to two. As consequence, four TBSs can be simultaneously installed and operated, versus six previously. Since the dedicated space in the various rooms of the Tokamak Complex has been kept unchanged, the existing

  • Developments on the tritium extraction and recovery system for HCPB
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-23
    Ion Cristescu; Mirela Draghia

    The tritium release from the ceramic materials of the DEMO Breading Blanket (BB) is a complex mechanism due mainly to the chemistry of the interface between the breeder material and the purge gas. The tritium atomic form produced inside the ceramics structure is converted into tritiated water at the surface of the breeder material due to interaction with residual moisture / water layer on the ceramics

  • Parametric design study of a substrate material for a DEMO sacrificial limiter
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-22
    R. de Luca; P. Fanelli; S. Mingozzi; G. Calabrò; F. Vivio; F. Maviglia; J.H. You

    Towards the realization of nuclear fusion, a future reactor must provide efficient and safe power exhaust through both the divertor and FW. Recent studies suggest that the greatest engineering challenges of plasma-facing components (PFCs) may arise from the occurrence of plasma transients, when extreme heat fluxes are expected. In severe cases, extensive surface vaporization, melting and re-solidification

  • Development of specific software for hydrogen isotopes separation by cryogenic distillation of ICSI Pilot Plant
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-22
    Alina Elena Niculescu; Anisia Bornea; George Ana; Marius Zamfirache

    ITER Isotope Separation System (ISS) is based on cryogenic distillation process and is one of the main systems within the ITER Fuel Cycle. Similar to ITER ISS, one of the key systems of a detritiation plant for low level tritiated light or heavy water is the cryogenic distillation. Provisions are being made in ICSI to design and build an experimental facility based on CECE (Combined Electrolytic Catalytic

  • CFD evaluation and optimization of the HEMJ divertor cooling design
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-21
    M. Zhao; B. Ghidersa; R. Stieglitz

    As part of the EU-DEMO prospective design studies, a He-cooled divertor concept has been developed by Karlsruhe Institute of Technologies (KIT). Using multi-jet cooling, the aim is to provide a robust solution that is capable of coping with heat fluxes as high as 10 MW/m2 under stationary loading conditions. Using the CFD simulations the cooling performance was evaluated and a configuration providing

  • Design of GPU-based parallel computation architecture of Thomson scattering diagnostic in KSTAR
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-21
    Seung-Ju Lee; Jongha Lee; Taehyun Tak; Taegu Lee; Jaesic Hong

    The Thomson scattering (TS) System is a diagnostic system to measure electron temperature and density profiles of tokamak plasma. The TS system requires measurement of many input signals, and the amount of raw data has significantly increased since the TS data acquisition (DAQ) system was upgraded to a fast digitizer. Research has been done on applying artificial neural network (ANN) to TS data analysis

  • Neutronics related integration studies of EU-DEMO pellet injection system
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-20
    A. Colangeli; R. Villari; F. Moro; D. Flammini; A. Frattolillo; F. Lucca; A. Marin; F. Viganò; F. Cismondi

    In the frame of the EU DEMO development program within the EUROfusion Consortium, the integration of the in-vessel components is crucial even at an early stage of the design process. The auxiliary, heating and fueling systems have to be integrated into the Breeding Blanket and, thus, they will undergo to a harsh nuclear environment during operation, and have a significant impact on the Tritium Breeding

  • Design of locating system on EAST horizontal Thomson scattering diagnostic
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-20
    Shumei Xiao; Qing Zang; Ailan Hu; Xiaofeng Han; Jiahui Hu; Xiao Zhang; Yongqi Gu

    Locating optical fibers in collection optics and confirmations of target scattering positions are critical to achieve accurate measurement for Thomson scattering diagnostic. Because the port size is limited on EAST, such procedures are difficult to be developed for horizontal Thomson scattering diagnostic. In order to solve this problem, a set of novel locating system has been designed and installed

  • Development of flow regime maps for lead lithium eutectic–helium flows
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-20
    E. Mas de les Valls; A. Cegielski; M. Jaros; M. Pérez-Ferragut; L. Batet; T. Sandeep; V. Chaudhari; J. Freixa

    Institute for Plasma Research (IPR), Gandhinagar (India) is currently involved in the design and development of its Lead-Lithium Ceramic Breeder (LLCB) module for testing in the International Thermo-nuclear Experimental Reactor (ITER). In order to fulfill the ITER safety requirements, some postulated events need to be analyzed. Among them, the internal loss of coolant accident is being studied using

  • Membrane gas-liquid contactor for tritium extraction from Pb-Li alloys
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-20
    Silvano Tosti; Alfonso Pozio; Luca Farina; Marco Incelli; Alessia Santucci; David Alique

    Gas-Liquid Contactors (GLC) have been studied for the extraction of tritium from Pb-Li blankets, typically consisting on bubble, packed or spray columns. Recent researches propose the use of Permeators-Against-Vacuum (PAV) in which dense metal membranes (V, Nb, Ta) are immersed into flowing Pb-Li: here the hydrogen isotopes pass through the membrane wall and are collected in the permeate side with

  • Preliminary engineering assessment of alternative magnetic divertor configurations for EU-DEMO
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-20
    Domenico Marzullo; Roberto Ambrosino; Antonio Castaldo; Giuseppe Di Gironimo; Samuel Merriman

    One of the main challenges in the roadmap to the realization of fusion energy is to develop a heat and power exhaust system able to withstand the large loads expected in the divertor of a fusion power plant. The challenge of reducing the heat load on the divertor targets is addressed, within the mission 2 ‘Heat-exhaust systems’, through the investigation of divertor configurations alternative to the

  • Numerical analysis of sub-atmospheric steam condensation in suppression tank with SIMMER IV code
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-19
    Alessio Pesetti; Rosa Lo Frano; Donato Aquaro

    One of the key safety components for nuclear fusion plants is the suppression tank, which is designed to protect the Vacuum Vessel (VV) against accidental pressurization events, e.g. Loss Of Coolant Accident (LOCA). In this framework the attention is focused on the Vacuum Vessel Pressure Suppression System (VVPSS), made of water tanks in which the pure steam, or eventually mixed with incondensable

  • Effect of titanium on the precipitation behaviors of transmutation elements in tungsten-titanium alloys from first-principles calculations
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-19
    Zelin Cao; Min Pan; Kaige Hu; Zheng Huang; Yini Lv; Shulong Wen; Yong Zhao; Huiqiu Deng

    Tungsten (W) and W-based alloys will produce a series of transmutation elements (TEs) such as osmium (Os), rhenium (Re), tantalum (Ta) and hafnium (Hf) under high-energy fusion neutron irradiation, which will lead to radiation-induced precipitation. As an alloying element, the effect of titanium (Ti) on the precipitation behaviors of transmutation products is investigated based on first-principles

  • The DTT secondary cooling water systems
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-18
    A. Rydzy; G. Barone; A. Marini

    The new Tokamak machine DTT (Divertor Tokamak Test), planned under construction by Enea Frascati Research Center, is a machine actively cooled by water. Although DTT is an intermittently operating machine, the thermal power that must be cooled is more or less 127 MW emitted within 100 seconds. Geographically the DTT site, at the Enea Frascati center, doesn’t allow the construction of water basins and

  • Measurements of the angle-dependent reflectivity of plasma-facing components and assessment of the impact on the estimations of coverage of the IVVS measurements of the ITER VV
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-18
    Paloma Matia-Hernando; Thomas Siegel; Efstathios Kolokotronis; Philip Bates; Marino Maiorino; Maria Pilar Urizar; Parthena Symeonidou; Carlo Damiani; Gregory Dubus; Adrian Puiu; Marta de la Fuente; Andres Cifuentes

    The In-Vessel Viewing System (IVVS) is a metrology instrument designed for deployment inside the ITER Vacuum Vessel (VV) to assess any damage, erosion or displacements of the Plasma-Facing Components (PFCs) through the lifetime of the ITER experiment. The latest developments of the IVVS have identified the Incidence-Angle-Dependent Reflectivity (IADR) of the PFCs to be critical to the assessment of

  • Preliminary sensitivity analysis for an ex-vessel LOCA without plasma shutdown for the EU DEMO WCLL blanket concept
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-16
    Matteo D’Onorio; Fabio Giannetti; Maria Teresa Porfiri; Gianfranco Caruso

    In this early development phase of the DEMO design the uncertainty affecting many operational and design parameters can modify main outcomes of accident scenario aiming at studying the critical conditions for the vacuum vessel and the contiguous containment volumes. The aim of this paper is to perform a preliminary sensitivity analysis of an accident progression predicted by MELCOR code considering

  • Heat treatment optimization of China low-activation ferritic/martensitic steel with cerium addition
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-15
    Jialin Gong; Hongbin Liao; Kuan Zhang; Xiaoyu Wang

    Chinese Low-activation Ferritic/martensitic steel (CLF-1) with rare earth (RE) cerium (Ce) addition (RE-CLF-1) was prepared by a two-step method: vacuum induction melting (VIM) and vacuum arc re-melting (VAR) processes. The effect of heat treatment parameter on the microstructure and thermal mechanical property of the obtained RE-CLF-1 was studied. The heat treatment process of the RE-CLF-1 was finally

  • Design of the 3.7 GHz, 1 kW CW solid state driver for LHCD system of the SST-1 tokamak
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-07
    Sandeep R. Sainkar; Alice N. Cheeran; Manjunath Reddy; Harish V. Dixit; Promod K. Sharma

    Advances in solid state device technologies have seen its use foray into areas such as microwave heating and powering the RF cavities of an accelerator. This paper explores the possibility of use of a solid state source/driver for current drive systems of tokamaks. The design of a modular solid state source to be used as a driver for klystrons in the LHCD system of SST-1 tokamak is presented. It comprises

  • Concept design of 100 kA hybrid DC breaker on China fusion engineering test reactor
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-05-07
    Hu Xingguang; Li Hua; Song Zhiquan; Ren Zhigang; Wang Shusheng; Tang Cunwen; Fu Peng

    China Fusion Engineering Test Reactor (CFETR) is the next device for the Chinese magnetic confinement fusion program, which is now under conceptual design phase. This paper describes the concept design of the high current hybrid breaker for the quench protection of superconducting magnets. The rated currents and maximum reapplied interruption voltages for breaker are 100 kA and 20 kV. The concept solution

  • Service Joining Strategy for the EU DEMO
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-30
    Tristan Tremethick; Simon Kirk; Keelan Keogh; Alexander O’Hare; Emily Harford; Ben Quirk

    As part of the European Research Roadmap to the Realisation of Fusion Energy, the DEMO reactor aims to show the feasibility of a fusion power plant. Due to the loss of revenue created by downtime and the potential for a breakdown to render a reactor inoperable, maintenance is “mission critical” for a power plant. The harsh environment of a fusion reactor dictates that maintenance must be carried out

  • Experimental study on effect of the toroidal magnetic field ripple on the ECR layer in TABAN tokamak
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-28
    D. Rostamifard; R. Amrollahi; D. Iraji

    The study of ECR assisted breakdown on TABAN tokamak (BT = 7000 G, R = 45 cm, a = 15 cm) has been directed to find the best conditions for ECR pre-ionization and plasma generation. The toroidal magnetic field ripple (δ) was investigated to obtain optimum pressure at pre-ionization phase. The internal chamber can be divided into two areas considering the δ value: underneath and between the coils.The

  • Insulated fixation system of plasma facing components to the divertor cassette in Eurofusion-DEMO
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-28
    Vito Imbriani; Ugo Bonavolontà; Giuseppe Di Gironimo; Samir El Shawish; Mike Fursdon; Louis Giannone; Domenico Marzullo; Giuseppe Mazzone; Eliseo Visca; Jeong-Ha You

    The design activities of an insulated Plasma Facing Components-Cassette Body (PFCs-CB) support has been carried out under the pre-conceptual design phase for Eurofusion-DEMO Work Package DIV-1 "Divertor Cassette Design and Integration" - Eurofusion Power Plant Physics & Technology (PPPT) program. The Eurofusion-DEMO divertor is a key in-vessel component with PFCs which directly interact with the plasma

  • DTT Thermal Shield: Preliminary thermal analysis
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-27
    Gianluca Barone; Selanna Roccella; Emanuela Martelli; Eliseo Visca

    The Thermal Shield (TS) in DTT is devoted to minimizing the heat loads from the tokamak warm components, to the superconducting magnets operating at 4.5 K. The TS is subdivided into three main regions covering respectively the vacuum vessel, the ports and the cryostat. Along the toroidal direction, the TS is arranged in 18 electrically insulated segments (20-degrees each), composed of several cooling

  • An Evaluation of the Global Effects of Tritium Emissions from Nuclear Fusion Power
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-25
    George Larsen; Dave Babineau

    Tritium, like all hydrogen isotopes, is difficult to confine and easily diffuses through most materials. As currently planned, fusion power plants will process and handle large quantities of deuterium and tritium as fuel, and therefore, will become sources of tritium input into the environment. Tritium releases from a worldwide distribution of tritium sources (fusion or fission) will lead to higher

  • Calibration study for a “fission electron-collection” neutron detector
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-25
    Dong Wang; Jianhua Zhang; Fenni Si; Xingyu Peng; Qingyuan Hu; Yiping Cai; Xuebin Zhu; Lizong Wang; Jianlun Yang; Chuanfei Zhang

    The “fission electron-collection” neutron detector (FECND) is a fission-based neutron detector with flat and fast responses, and is a promising choice for the diagnostics of fusion neutron flux and spectrum. Due to its low sensitivity, the detector output is weak and is disturbed easily in calibration experiments. The backgrounds in the calibration primarily come from scattered neutrons and the γ rays

  • Measurement of effective thermal conductivity of lithium metatitanate pebble bed by transient hot-wire technique
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-25
    Maulik Panchal; Abhishek Saraswat; Shrikant Verma; Paritosh Chaudhuri

    In future fusion reactor lithium metatitanate (Li2TiO3) as a functional material in the form of packed pebble beds have been selected in breeding blanket concepts to generate and release tritium. The effective thermal conductivity (keff) of Li2TiO3 pebble beds are needed to be well characterized for the design and analysis of breeding blankets under fusion relevant conditions. The transient hot-wire

  • Impact of plasma thermal transients on the design of the EU DEMO first wall protection
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-25
    Francesco Maviglia; Christian Bachmann; Gianfranco Federici; Thomas Franke; Mattia Siccinio; Christian Vorpahl; Raffaele Albanese; Roberto Ambrosino; Emiliano Fable; Mehdi Firdaouss; Jonathan Gerardin; Vincenzo Paolo Loschiavo; Massimiliano Mattei; Francesco Palermo; Maria Lorena Richiusa; Fabio Villone; Zsolt Vizvary

    The protection of the EU-DEMO first wall (FW) during plasma transients represents one of the main challenges of the current pre-concept design phase. While the present DEMO FW design heat load capability is of the order of ≈1−2 MW/m2 in steady state, this limit is overcome during plasma transients for both normal and off-normal events leading to a plasma-wall contact. A strategy to protect the FW is

  • Numerical investigation of purge gas flow through binary-sized pebble beds using discrete element method and computational fluid dynamics
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-25
    Youngmin Lee; Dongkwon Choi; Seon-Pil Hwang; Mu-Young Ahn; Yi-Hyun Park; Seungyon Cho; Dongwoo Sohn

    The main function of the breeder materials, such as Li2TiO3, Li4SiO4, and Li2ZrO3, in a solid type breeding blanket is to produce tritium, which is transferred to the fuel cycle system by helium purge gas and is used as a fuel for nuclear fusion reactions. Since the configuration of the breeder should be designed considering the purge gas flow, the form of pebble beds is mainly adopted, thereby leading

  • Restraining method for delta ferrite in 9Cr-RAFM welded joint
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-25
    Jian Wang; Jinpeng Zhou; Xiaofeng Lu; Lei Cui

    Reduced activation ferritic/martensitic (RAFM) steels are important candidates for the first wall and blanket structure in future ADS and fission and fusion energy system. The development of welding process for fusion-centered applications is critical when evaluating the applicability of RAFM steels as structural materials for fusion pressure equipment. Delta ferrite is likely to emerge after welding

  • European DEMO first wall shaping and limiters design and analysis status
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-24
    Z. Vizvary; W. Arter; C. Bachmann; T.R. Barrett; B. Chuilon; P. Cooper; E. Flynn; M. Firdaouss; T. Franke; J. Gerardin; R. Gowland; M. Kovari; F. Maviglia; M.L. Richiusa; E. V. Rosa Adame; C. Vorpahl; A. Wilde; Y. Xue

    The anticipated heat flux limit of the European DEMO first wall is ∼1−2 MW/m2. During transient and off normal events, the heat load deposited on the wall would be much larger than the steady state heat load and exceed the first wall limit, therefore the breeding blanket first wall needs to be protected. This involves dedicated discrete limiters in certain regions of the machine that would take the

  • Realization of ODS-Cu/T91 Tube-to-tube Joining with Rotary Friction Welding
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-23
    Sixiang Zhao; Minjing Wang; Shengzhong Kou; Zhi Jia; Wanjing Wang; Qiang Li; G.N. Luo

    Non-melting tube-to-tube joining techniques between copper and steel are of significance for the development of monoblock type divertor components for future DEMO. In this study, tube-to-tube joining between T91 and ODS-Cu (reference materials), which is difficult to accomplish by traditional fusion welding, has been successfully realized with a rotary friction welding manner. The junctions are free

  • Thermal-hydraulic modeling and analysis of the Water Cooling System for the ITER Test Blanket Module
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-23
    Cristiano Ciurluini; Fabio Giannetti; Amelia Tincani; Alessandro Del Nevo; Gianfranco Caruso; Italo Ricapito; Fabio Cismondi

    The Water Cooled Lithium Lead (WCLL) is one of the selected breeding blanket (BB) concepts to be investigated in the EUROfusion Breeding Blanket Project (WPBB), and it was also recently chosen as one of the mock-up for ITER Test Blanket Module (TBM) program. The program foresees the test of different BB mock-ups, called Test Blanket Modules, with all the related ancillary systems. A pre-conceptual

  • Adsorption of hydrogen and deuterium on A-type zeolites at 77 K after various heat treatments
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-23
    Norihiro Ikemoto; Tomohiko Kawakami; Kazuo Yonehara; Yuri Natori; Katsuyoshi Tatenuma; Masanori Hara

    Molecular sieves (MSs) are a potentially useful means of hydrogen isotope recovery as well as the temporary storage and supply of such isotopes. In this study, the hydrogen adsorption isotherms for various MS samples were acquired after applying different water removal processes to activate the MSs. MS-3A sample was found not to adsorb hydrogen regardless of the activation conditions. MS-4A sample

  • The simulation of the application of quasi-snowflake magnetic configuration on the EAST new lower tungsten divertor
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-22
    Chen Zhang; Chaofeng Sang; Liang Wang; Zhengping Luo; Guosheng Xu; Daoyuan Liu; Dezhen Wang

    In this work, the predicted application of the quasi-snowflake (QSF) magnetic equilibrium on the EAST new designed lower divertor shape is presented and compared with the conventional magnetic configuration by using SOLPS-ITER simulation. The performance of heat flux control, power dissipation of the QSF divertor (QSFD) are studied on the potential EAST new upgrade lower tungsten (W) divertor. The

  • Numerical simulation for operational state switching processes of CN HCCB TBS helium cooling system
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-17
    Xiaoyong Wang; Xiaoyu Wang; Hongxiang Zhang; Xingfu Ye; Jie Liu; Yongjiang Yan; Zhengning Zhao

    Understanding the thermal hydraulic behaviors of switching processes between operational states is one of key issues in the design and optimization of the helium cooling system (HCS) of China helium cooled ceramic breeder test blanket system (CN HCCB TBS). The thermal hydraulic model of HCS considering the influence of heat structural and thermal insulation layer was established, and the main switching

  • Retarding field analyzer for the wendelstein 7-X boundary plasma
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-20
    M. Henkel; Y. Li; Y. Liang; P. Drews; A. Knieps; C. Killer; D. Nicolai; D. Höschen; J. Geiger; C. Xiao; N. Sandri; G. Satheeswaran; S. Liu; O. Grulke; M. Jakubowski; S. Brezinsek; M. Otte; O. Neubauer; J. Cai

    A bi-directional multi-channel retarding field analyzer (RFA) probe has been successfully developed for the first time on the Wendelstein 7-X (W7-X) stellarator boundary plasma. Modifications to the RFA prototype hardware and its upgrade for the two W7-X island divertor campaigns are presented, including the electronics. In this paper the experiences and challenges operation and customizing an RFA

  • Phase detection system based on digital signal processing in millimeter wave interferometer for fusion plasma diagnostics
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-23
    Alpesh Vala; Hiren Mewada; Amit Patel; Umesh Nagora; Suryakumar Pathak

    The design and development of a processing unit for the microwave interferometer is proposed here. The Microwave interferometer is well known for its use in several applications, and one of them is to measure the electron density of fusion plasma. An indigenous effort has been made to build a phase detection system that can be used to calculate fusion plasma electron density for millimeter-wave interferometer

  • Optimization mechanism of bonding strength of laser melting deposited tungsten/reduced activation steel heterogeneous interface via addition of Y2O3 by magnetron sputtering
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-31
    Wenjuan Jiang; Zhixin Xia; Jiachao Xu; Dong Zhao; Shaoqiu Xia; Liang Wang

    In this study, magnetron sputtering Y2O3 layer has been used as the nucleating agent at tungsten/reduced activation steel interface for solving the bonding difficulty between tungsten and reduced activation steels owing to the formation of FeW phase and the huge residual stress at tungsten/reduced activation steel interface fabricated by laser melting deposition. Moreover, SEM, EBSD and nanoscratch

  • Nonlinear effects of FCI electrical conductivity on the MHD flow in DCLL blanket
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-31
    Long Chen; Le Hao; Ming-Jiu Ni; Nian-Mei Zhang

    In a Dual Coolant Lead Lithium (DCLL) blanket, flow channel insert (FCI) with low electrical conductivity and low thermal conductivity is introduced to reduce the MHD pressure drop and improve the heat transfer efficiency. In the present work, we aim at performing a direct simulation of the magneto-thermal-fluid-structure multi-physical fields in a typical poloidal duct with different electrical conductivities

  • Crack propagation analysis of ITER Vacuum Vessel port stub with Radial Basis Functions mesh morphing
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-31
    Edoardo Pompa; Gabriele D’Amico; Stefano Porziani; Francesco Giorgetti; Corrado Groth; Alfredo Portone; Marco E. Biancolini

    The ITER Vacuum Vessel (VV) is one of the most important components of the tokamak machine. The severe operating conditions impose a design to withstand strong dynamic loads. A special focus is put on defects embedded in the component that, due to the not total accessibility of the VV to non-destructive examination (NDE), but also to identify their minimum safe dimension, must be assessed through Fracture

  • Study on the irradiation damage in Fe-based metallic glasses induced by Ne10+ ions
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-31
    Yachao Suo; Lisong Zhang; Tong Guan; Xianxiu Mei; Xiaonan Zhang; Chonghong Zhang; Yitao Yang; Xingzhong Cao; Jianbing Qiang; Younian Wang

    Due to the long-range disordered structure, metallic glass could be used as a candidate for radiation-resistant materials. The article showed the influence of Ne10+ ions irradiation on the microstructure, atomic arrangement and defects of Fe-based metallic glasses. Metallic glasses Fe80Si7B13 and Fe68Zr7B25 remained amorphous after irradiation at different fluence. However, the atomic arrangement of

  • Development of an assistant program for CAD-to-cosRMC modelling
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-04-14
    Hua Du; Yue-Tong Luo; Chengcun Han; Lei Lu; Yiman Yan; Yeshuai Sun; Yixue Chen; Songlin Liu

    cosRMC is a three dimensional neutron photon transport Monte Carlo code based on the stochastic statistical method of neutral particle transport. At present, the cosRMC geometry model is imported by writing the input file manually, which leads to time-consuming, inefficient and error-prone problems. This will limit application of cosRMC in fusion neutronics calculations because of the complex geometry

  • Shielding concept and neutronic assessment of the DEMO lower remote handling and pumping ports
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-19
    Aljaž Čufar; Christian Bachmann; Tim Eade; Davide Flammini; Curt Gliss; Ivan A. Kodeli; Domenico Marzullo; Giuseppe Mazzone; Christian Vorpahl; Andrew Wilde

    Within the EUROfusion Power Plant Physics and Technology Department the DEMOnstrational fusion power plant (DEMO) is being developed. One of the fundamental challenges is the integration of ports in the vacuum vessel. The lower port of the DEMO machine is particularly challenging due to tight space constraints imposed by the toroidal field (TF) coils and the requirement to provide a large open duct

  • Progress in development of advanced D1S dynamic code for three-dimensional shutdown dose rate calculations
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-19
    Giovanni Mariano; Davide Flammini; Nicola Fonnesu; Fabio Moro; Rosaria Villari

    The Advanced Direct 1-Step (AdD1S) is one of the most validated tools for the evaluation of Shutdown Dose Rate in complex fusion tokamak machines. The present work is built on the experience in analyzing tokamaks using prior versions of the code, where the activation analysis was limited to the first-step reactions. In the light of the neutron irradiation scenarios foreseen for fusion power reactors

  • Neutronic analyses in support of the conceptual design of the DTT tokamak radial neutron camera
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-19
    Barbara Caiffi; Maurizio Angelone; Andrea Colangeli; Davide Flammini; Nicola Fonnesu; Raul Luìs; Giovanni Mariano; Daniele Marocco; Fabio Moro; Marco Tardocchi; Rosaria Villari

    The Divertor Tokamak Test (DTT) facility, whose design phase is currently under finalization, is an Italian project aimed to investigate alternative power exhaust solutions for DEMO. It is designed to operate with significant power loads and enough flexibility to test innovative divertor configurations, different plasma edge and bulk conditions approaching, as much as possible, those planned for DEMO

  • Recovery from a hot water leakage at the tokamak ASDEX upgrade
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-25
    Volker Rohde; Albrecht Herrmann; Martin Balden; Detlef Bösser; Katja Hunger; Gerd Schall; Michaela Uhlmann; Thomas Vierle

    During a regular vessel baking after a manned access a leakage of a heating/cooling pipe released about 100 l of hot water into the vacuum vessel of ASDEX Upgrade (AUG). Erosion of a ten years old Cu gasket by water during baking causes the leak. At plasma facing components the water steam forms white remnants as it reacted with the boron-hydride layers used for wall conditioning and cause locally

  • A new array eddy current testing probe for inspection of small-diameter tubes in Tokamak fusion devices
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-21
    Yingsong Zhao; Pan Qi; Zheng Xie; Peigen Bai; Hong-En Chen; Shejuan Xie; Shusheng Liao; Zhenmao Chen

    Nuclear power plant (NPP) is the final goal of the research and developments on the controlled nuclear fusion technologies. Small-diameter tubes, such as the mono-block tubes in divertors, are widely used in Tokamak fusion reactors and other NPPs. To ensure safety of controlled fusion reactors, pre-service and in-service non-destructive testing (NDT) for small-diameter tubes are of great importance

  • Progress of the conceptual design of the European DEMO breeding blanket, tritium extraction and coolant purification systems
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-27
    F. Cismondi; G.A. Spagnuolo; L.V. Boccaccini; P. Chiovaro; S. Ciattaglia; I. Cristescu; C. Day; A. Del Nevo; P.A. Di Maio; G. Federici; F. Hernandez; C. Moreno; I. Moscato; P. Pereslavtsev; D. Rapisarda; A. Santucci; M. Utili

    In the frame of the EUROfusion consortium activities the Helium Cooled Pebble Bed (HCPB) and the Water Cooled Lithium Lead (WCLL) concepts are being developed as possible candidates to become driver Breeding Blanket (BB) for the EU DEMO, which aims at the tritium self-sufficiency and net electricity production. The two BB design options encompass water or helium as coolants and solid ceramic with

  • Evaluation of thermal profile in catalytic reactor by exothermic hydrocarbon feed into detritiation system
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-27
    Yuki Edao; Yasunori Iwai

    Air detritiation system of a fusion facility consists of catalyst reactors for tritium oxidation and a system to remove tritiated vapor from air. In the event of fire in the facility, the impact of gaseous impurities produced by polymeric materials on catalytic oxidation of tritium is one of the evaluation items. The point is the impact of reaction heat needs to be detected since gaseous impurities

  • Conceptual design of test modules for DEMO blanket, diagnostic device, and RI production for A-FNS
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-31
    Masayuki Ohta; Satoshi Sato; Makoto M. Nakamura; Saerom Kwon; ChangHo Park; Kentaro Ochiai; Youji Someya; Atsushi Kasugai

    In the conceptual design activity of advanced fusion neutron source A-FNS, a variety of test modules for fusion DEMO reactor are planned. In these modules, progresses of design activities on Blanket Nuclear Property Test Module (BNPTM) and Diagnostic and Control Device Test Module (DCDTM) were reported. The BNPTM is a module in order to evaluate accuracies of nuclear analyses of the DEMO blanket such

  • Analysis of flow channel insert deformations influence on the liquid metal flow in DCLL blanket channels
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-29
    Dionisio Di Giulio; Daniel Suarez; Lluis Batet; Elisabet Mas de les Valls; Laura Savoldi

    The dual coolant lithium lead (DCLL) is a candidate to be an effective breeding blanket (BB) concept for nuclear fusion technologies. One critical point of this design is the magnetohydrodynamic (MHD) effects involving Lorentz damping force which produces relevant pressure drop in the eutectic flow. In the framework of the European DEMO, the application of sandwich-like steel–alumina–steel flow channel

  • Helium blanketing and gas circulation effect of hydriding reaction with ZrCo
    Fusion Eng. Des. (IF 1.457) Pub Date : 2020-03-31
    Pil-Kap Jung; Min Ho Chang; Dong-you Chung; Hyun-Goo Kang; Jea-Uk Lee

    The tritium storage bed receives and supplies tritium gas for fusion facilities. Inert gases, such as helium-3, which is produced by tritium beta decay, disturb the absorption behavior of tritium in the bed and this is called the Helium Blanketing Effect (HBE). This study focused on the HBE and gas circulation as a way of breaking the HBE for the hydriding reaction with ZrCo. The hydriding reaction

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