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Sökning: L773:0920 3796 OR L773:1873 7196

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1.
  • Rubel, Marek J., et al. (författare)
  • Molybdenum limiters for Extrap-T2 upgrade : surface properties and high heat flux testing
  • 2000
  • Ingår i: Fusion engineering and design. - 0920-3796 .- 1873-7196. ; 49, s. 323-329
  • Tidskriftsartikel (refereegranskat)abstract
    • A vacuum vessel, a conductive copper shell and plasma-facing components of the Extrap-T2 device, a medium size reversed field pinch, are under major rebuild. The new machine is equipped with an array of 180 molybdenum limiters, which will be exposed to power loads of about 30 MW m(-2), but higher loads can not be excluded. Prior to the limiters' installation in the vessel, they were tested under high heat loads in the JUDITH electron beam facility in order to assess the possible damage to the surface and to the bulk of the material. A test limiter was irradiated with a beam of increasing power density from 15 to 1500 MW m(-2). Surface characterization was performed before and after the irradiation using electron and optical microscopy, energy dispersive X-ray spectroscopy, enhanced proton scattering and laser profilometry. Metallography studies were performed for the irradiated areas. The irradiation induced the change in surface morphology, e.g. surface melting and re-crystallization of grains, only following the 10 ms long pulses with the absorbed power density approaching 1500 MW m(-2).
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2.
  • Tanabe, T., et al. (författare)
  • Material mixing on W/C twin limiter in TEXTOR-94
  • 2000
  • Ingår i: Fusion engineering and design. - 0920-3796 .- 1873-7196. ; 49, s. 355-362
  • Tidskriftsartikel (refereegranskat)abstract
    • In order to investigate the effect of mutual contamination between tungsten (W) and carbon
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3.
  • Agredano Torres, Manuel, et al. (författare)
  • Coils and power supplies design for the SMART tokamak
  • 2021
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 168, s. 112683-112683
  • Tidskriftsartikel (refereegranskat)abstract
    • A new spherical tokamak, the SMall Aspect Ratio Tokamak (SMART), is currently being designed at the University of Seville. The goal of the machine is to achieve a toroidal field of 1 T, a plasma current of 500 kA and a pulse length of 500 ms for a plasma with a major radius of 0.4 m and minor radius of 0.25 m. This contribution presents the design of the coils and power supplies of the machine. The design foresees a central solenoid, 12 toroidal field coils and 8 poloidal field coils. Taking the current waveforms for these set of coils as starting point, each of them has been designed to withstand the Joule heating during the tokamak operation time. An analytical thermal model is employed to obtain the cross sections of each coil and, finally, their dimensions and parameters. The design of flexible and modular power supplies, based on IGBTs and supercapacitors, is presented. The topologies and control strategy of the power supplies are explained, together with a model in MATLAB Simulink to simulate the power supplies performance, proving their feasibility before the construction of the system.
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4.
  • Baron-Wiechec, A., et al. (författare)
  • Thermal desorption spectrometry of beryllium plasma facing tiles exposed in the JET tokamak
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 133, s. 135-141
  • Tidskriftsartikel (refereegranskat)abstract
    • The phenomena of retention and de-trapping of deuterium (D) and tritium (T) in plasma facing components (PFC) and supporting structures must be understood in order to limit or control total T inventory in larger future fusion devices such as ITER, DEMO and commercial machines. The goal of this paper is to present details of the thermal desorption spectrometry (TDS) system applied in total fuel retention assessment of PFC at the Joint European Torus (JET). Examples of TDS results from beryllium (Be) wall tile samples exposed to JET plasma in PFC configuration mirroring the planned ITER PFC is shown for the first time. The method for quantifying D by comparison of results from a sample of known D content was confirmed acceptable. The D inventory calculations obtained from Ion Beam Analysis (IBA) and TDS agree well within an error associated with the extrapolation from very few data points to a large surface area.
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5.
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6.
  • Batistoni, P., et al. (författare)
  • Technical preparations for the in-vessel 14 MeV neutron calibration at JET
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 117, s. 107-114
  • Tidskriftsartikel (refereegranskat)abstract
    • The power output of fusion devices is measured from their neutron yields which relate directly to the fusion yield. In this paper we describe the devices and methods that have been prepared to perform a new in situ 14 MeV neutron calibration at JET in view of the new DT campaign planned at JET in the next years. The target accuracy of this calibration is 10% as required for ITER, where a precise neutron yield measurement is important, e.g., for tritium accountancy. In this paper, the constraints and early decisions which defined the main calibration approach are discussed, e.g., the choice of 14 MeV neutron source and the deployment method. The physics preparations, source issues, safety and engineering aspects required to calibrate directly the JET neutron detectors are also discussed. The existing JET remote-handling system will be used to deploy the neutron source inside the JET vessel. For this purpose, compatible tooling and systems necessary to ensure safe and efficient deployment have been developed. The scientific programme of the preparatory phase is devoted to fully characterizing the selected 14 MeV neutron generator to be used as the calibrating source, obtain a better understanding of the limitations of the calibration, optimise the measurements and other provisions, and to provide corrections for perturbing factors (e.g., anisotropy of the neutron generator, neutron energy spectrum dependence on emission angle). Much of this work has been based on an extensive programme of Monte-Carlo calculations which provide support and guidance in developing the calibration strategy.
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7.
  • Batistoni, P., et al. (författare)
  • Technological exploitation of Deuterium-Tritium operations at JET in support of ITER design, operation and safety
  • 2016
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 109, s. 278-285
  • Tidskriftsartikel (refereegranskat)abstract
    • Within the framework of the EUROfusion programme, a work-package of technology projects (WPJET3) is being carried out in conjunction with the planned Deuterium-Tritium experiment on JET (DTE2) with the objective of maximising the scientific and technological return of DT operations at JET in support of ITER. This paper presents the progress since the start of the project in 2014 in the preparatory experiments, analyses and studies in the areas of neutronics, neutron induced activation and damage in ITER materials, nuclear safety, tritium retention, permeation and outgassing, and waste production in preparation of DTE2.
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8.
  • Batistoni, P., et al. (författare)
  • The JET technology program in support of ITER
  • 2014
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 89:7-8, s. 896-900
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper presents an overview of the current and planned technological activities at JET in support of ITER operation and safety. The scope is very broad and it ranges from analysis of components from the ITER-like Wall (ILW) to determine material erosion and deposition, dust generation and fuel retention to neutronics measurements and analyses. Preliminary results are given of the post-mortem analyses of samples exposed to JET plasmas during the first JET-ILW operation in 2011-2012, and retrieved during the following in-vessel intervention. JET is the only fusion machine capable of producing significant neutron yields, up to nearly 10(19) n/s (14.1 MeV) in DT operations. Recently, the technological potential of a new DT campaign at JET in support of ITER has been explored and the outcome of this assessment is presented. The expected 14 MeV neutron yield, the use of tritium, the preparation and implementation of safety measures will provide a unique occasion to gain experience in several ITER relevant technological areas. A number of projects and experiments to be conducted in conjunction with the DT operation have been identified and they are described in this paper.
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9.
  • Biel, W., et al. (författare)
  • Development of a concept and basis for the DEMO diagnostic and control system
  • 2022
  • Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 179
  • Tidskriftsartikel (refereegranskat)abstract
    • An initial concept for the plasma diagnostic and control (D&C) system has been developed as part of European studies towards the development of a demonstration tokamak fusion reactor (DEMO). The main objective is to develop a feasible, integrated concept design of the DEMO D&C system that can provide reliable plasma control and high performance (electricity output) over extended periods of operation. While the fusion power is maximized when operating near to the operational limits of the tokamak, the reliability of operation typically improves when choosing parameters significantly distant from these limits. In addition to these conflicting requirements, the D&C development has to cope with strong adverse effects acting on all in vessel components on DEMO (harsh neutron environment, particle fluxes, temperatures, electromagnetic forces, etc.). Moreover, space allocation and plasma access are constrained by the needs for first wall integrity and optimization of tritium breeding. Taking into account these boundary conditions, the main DEMO plasma control issues have been formulated, and a list of diagnostic systems and channels needed for plasma control has been developed, which were selected for their robustness and the required coverage of control issues. For a validation and refinement of this concept, simulation tools are being refined and applied for equilibrium, kinetic and mode control studies.
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10.
  • Biel, W., et al. (författare)
  • Diagnostics for plasma control - : From ITER to DEMO
  • 2019
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 146:A, s. 465-472
  • Tidskriftsartikel (refereegranskat)abstract
    • The plasma diagnostic and control (D&C) system for a future tokamak demonstration fusion reactor (DEMO) will have to provide reliable operation near technical and physics limits, while its front-end components will be subject to strong adverse effects within the nuclear and high temperature plasma environment. The ongoing developments for the ITER D&C system represent an important starting point for progressing towards DEMO. Requirements for detailed exploration of physics are however pushing the ITER diagnostic design towards using sophisticated methods and aiming for large spatial coverage and high signal intensities, so that many front-end components have to be mounted in forward positions. In many cases this results in a rapid aging of diagnostic components, so that additional measures like protection shutters, plasma based mirror cleaning or modular approaches for frequent maintenance and exchange are being developed. Under the even stronger fluences of plasma particles, neutron/gamma and radiation loads on DEMO, durable and reliable signals for plasma control can only be obtained by selecting diagnostic methods with regard to their robustness, and retracting vulnerable front-end components into protected locations. Based on this approach, an initial DEMO D&C concept is presented, which covers all major control issues by signals to be derived from at least two different diagnostic methods (risk mitigation).
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11.
  • Binda, Federico, 1987-, et al. (författare)
  • Calculation of the profile-dependent neutron backscatter matrix for the JET neutron camera system
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 865-868
  • Tidskriftsartikel (refereegranskat)abstract
    • We investigated the dependence of the backscatter component of the neutron spectrum on the emissivity profile. We did so for the JET neutron camera system, by calculating a profile-dependent backscatter matrix for each of the 19 camera channels using a MCNP model of the JET tokamak. We found that, when using a low minimum energy for the summation of the counts in the neutron pulse height spectrum, the backscatter contribution can depend significantly on the emissivity profile. The maximum variation in the backscatter level was 24% (8.0% when compared to the total emission). This effect needs to be considered when a correction for the backscatter contribution is applied to the measured profile. (C) 2017 The Authors. Published by Elsevier B.V.
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12.
  • Boltruczyk, G., et al. (författare)
  • Development of MPPC-based detectors for high count rate DT campaigns at JET
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 940-944
  • Tidskriftsartikel (refereegranskat)abstract
    • The products of fusion reactions at JET are measured using different diagnostic techniques. One of the methods is based on measurements of gamma-rays, originating from reactions between fast ions and plasma impurities. During the forthcoming deuterium-tritium (DT) campaign a particular attention will be paid to 4.44 MeV gamma-rays emitted in the Be-9(alpha,n gamma)C-12 reaction. Gamma-ray detectors foreseen for measurements in DT campaigns have to be able to register spectra at high count rates, up to approximately 500 kHz. For the Gamma-ray Camera at JET a new setup will be based on scintillators with a short decay time, e.g., CeBr3, and a multi-pixel photon counter (MPPC). We present two methods of output signal shortening in modules based on MPPC. A short detector output signal is necessary in order to minimize the number of pile up events at high count rates. One method uses a passive RC circuit with a pole zero cancellation, whereas an active transimpedance amplifier is used in the other one. Due to the strong dependence of MPPC properties on temperature variation, a special device MTCD@NCBJ was designed and produced to stabilize the gain in MPPC-based scintillation detectors. We show that this device guarantees stable working conditions.
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13.
  • Boyer, Helen, et al. (författare)
  • JET Tokamak, preparation of a safety case for tritium operations
  • 2016
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 109, s. 1308-1312
  • Tidskriftsartikel (refereegranskat)abstract
    • A new Safety Case is required to permit tritium operations on JET during the forthcoming DTE2 campaign. The outputs, benefits and lessons learned associated with the production of this Safety Case are presented. The changes that have occurred to the Safety Case methodology since the last JET tritium Safety Case are reviewed. Consideration is given to the effects of modifications, particularly ITER related changes, made to the JET and the impact these have on the hazard assessments as well as normal operations. Several specialized assessments, including recent MELCOR modelling, have been undertaken to support the production of this Safety Case and the impact of these assessments is outlined. Discussion of the preliminary actions being taken to progress implementation of this Safety Case is provided, highlighting new methods to improve the dissemination of the key Safety Case results to the plant operators. Finally, the work required to complete this Safety Case, before the next tritium campaign, is summarized. (C) 2016 EURATOM. Published by Elsevier B.V. All rights reserved.
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14.
  • Calabro, G., et al. (författare)
  • Divertor currents optimization procedure for JET-ILW high flux expansion experiments
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 129, s. 115-119
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper deals with a divertor coil currents optimized procedure to design High Flux Expansion (HFE) configurations in the JET tokamak aimed to study the effects of flux expansion variation on the radiation fraction and radiated power re-distribution. A number of benefits of HFE configuration have been experimentally demonstrated on TCV, EAST, NSTX and DIII-D tokamaks and are under investigation for next generation devices, as DEMO and DTT. The procedure proposed here exploits the linearized relation between the plasma-wall gaps and the Poloidal Field (PF) coil currents. Once the linearized model is provided by means of CREATE-NL code, the divertor coils currents are calculated using a constrained quadratic programming optimization procedure, in order to achieve HFE configuration. Flux expanded configurations have been experimentally realized both in ohmic and heated plasma with and without nitrogen seeding. Preliminary results on the effects of the flux expansion variation on total power radiation increase will be also briefly discussed.
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15.
  • Cavinato, M., et al. (författare)
  • Comparison of strategies and regulator design for active control of MHD modes
  • 2005
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 74:1-4, s. 549-553
  • Tidskriftsartikel (refereegranskat)abstract
    • A system of evenly spaced poloidal arrays of saddle coils was recently installed on the reversed field pinch device EXTRAP T2R to perform experiments on the active control of MHD modes. The implementation of different control strategies, such as "intelligent shell" and "mode control", was made possible by a flexible digital control system. After giving some results on the performances of the innermost coil current control loop, two versions of "mode control" recently tested on the machine are presented. In the "wise shell" approach, equilibrium related modes are ruled out and a systematic increase of the pulse length is obtained. In a second, more model based, approach, a mode estimator/controller is designed aiming at a full state feedback by including modes, which are not directly measurable due to the limited number of available real-time signals.
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16.
  • Cecconello, Marco, et al. (författare)
  • Conceptual design of a collimated neutron flux monitor and spectrometer for DTT
  • 2021
  • Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 167
  • Tidskriftsartikel (refereegranskat)abstract
    • A conceptual design and performance studies for a collimated neutron flux monitor and neutron spectrometer for the Divertor Tokamak Test (DTT) facility are presented. This study is based on the single-null divertor configuration and for “Half Power” and “Full power” scenarios with 15 MW of negative-ion NBI, 29 MW of ECH and 3 MW of ICRF heating with a maximum neutron yield of 1.5 × 1017 s−1. Fast ion distributions (both from auxiliary heating systems and fusion born) have been simulated in TRANSP/NUBEAM and the corresponding neutron energy spectra have been calculated using DRESS. Synthetic diagnostics have been implemented to determine the neutron fluxes and spectra at the detector location. Neutron emissivity profiles, plasma position, core ion temperature and the ratio of thermal and non-thermal D ion populations can be obtained with good accuracy and time resolution.
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17.
  • Cecconello, Marco, et al. (författare)
  • Neural network implementation for ITER neutron emissivity profile recognition
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 637-640
  • Tidskriftsartikel (refereegranskat)abstract
    • The ITER Radial Neutron Camera (RNC) is a neutron diagnostic intended for the measurement of the neutron emissivity radial profile and the estimate of the total fusion power. This paper presents a proof of-principle method based on neural networks to estimate the neutron emissivity profile in different ITER scenarios and for different RNC architectures. The design, optimization and training of the implemented neural network is presented together with a decision algorithm to select, among the multiple trained neural networks, which one provides the inverted neutron emissivity profile closest to the input one. Examples are given for a selection of ITER scenarios and RNC architectures. The results from this study indicate that neural networks for the neutron emissivity recognition in ITER can achieve an accuracy and precision within the spatial and temporal requirements set by ITER for such a diagnostic.
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18.
  • Cecconello, Marco, et al. (författare)
  • Strategy and guidelines for the calibration of the ITER Radial Neutron Camera
  • 2019
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 146, s. 2049-2052
  • Tidskriftsartikel (refereegranskat)abstract
    • A calibration procedure is proposed for ITER Radial Neutron Camera that relies on embedded sources, reference ITER pulses and cross-calibration with ITER fission chambers and activation system coupled to Monte Carlo simulations of radiation transport. The proposed procedure would allow to measure the neutron emissivity profile and of the fusion power with 10 % accuracy and precision, a time resolution of 10 ms and a spatial resolution of a/10 for ITER entire life-time.
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19.
  • Chiariello, Andrea Gaetano, et al. (författare)
  • A 3D electromagnetic model of the iron core in JET
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 527-531
  • Tidskriftsartikel (refereegranskat)abstract
    • The Magnet and Power Supplies system in JET includes a ferromagnetic core able to increase the transformer effect by improving the magnetic coupling with the plasma. The iron configuration is based on an inner cylindrical core and eight returning limbs; the ferromagnetic circuit is designed in such a way that the inner column saturates during standard operations [1]. The modelling of the magnetic circuit is a critical issue because of its impact on several applications, including equilibrium and reconstruction analysis required for control applications. The most used model in present applications is based on Equivalent Currents (ECs) placed on the iron boundary together with additional specific constraints, in a 2D axisymmetric frame. The (circular) ECs are chosen, by using the available magnetic measurements, to best represent the magnetic polarization effect [1]. Due to the axisymmetric assumption such approach is not well suited to deal with significant 3D effects, e.g. arising in operations with Error Field Correction Coils (EFCC). In this paper a new methodology is proposed, based on a set of 3D-shaped ECs and able to better model the actual 3D magnetization giving rise to a linear system to be solved. According to a well assessed approach [2], the 3D shape of ECs is represented by a set of elementary sources. The methodology has been successfully validated in a number of JET dry-run experiments where 3D effects are generated by EFCC currents. The new procedure has been designed to be easily coupled with equilibrium or reconstruction codes such as EFIT/V3FIT. The proposed model resulted to be very effective in representing 3D iron magnetization, especially if compared with typical 2D models.
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20.
  • Coad, J. P., et al. (författare)
  • Diagnostics for studying deposition and erosion processes in JET
  • 2005
  • Ingår i: Fusion engineering and design. - : Elsevier. - 0920-3796 .- 1873-7196. ; 74:1-4, s. 745-749
  • Tidskriftsartikel (refereegranskat)abstract
    • Estimates of erosion, deposition and H-isotope retention in JET from previous divertor campaigns have relied on analysis of in-vessel components removed at shutdowns. The components analysed have also provided an incomplete coverage of the vessel. In 2004, new diagnostics are being installed to give a more complete picture (such as smart tiles) and to provide some time resolution. The latter includes further quartz microbalances (QMB), following the successful operation of a prototype in 2002-2004 [H.-G. Esser, G. Neill, P. Coad, G.F. Matthews, D. Jolovic, D. Wilson, M. Freisinger, V. Philipps, Quartz microbalance: a time-resolved diagnostic to measure material deposition in JET, Fusion Eng. Des. 66-68 (2003) 855-860; H.-G. Esser, V. Philipps, M. Freisinger, G.F. Matthews, J.P. Coad, G.F. Neill, JET EFDA Contributors, Effect of plasma configuration on carbon migration measured in the inner divertor of JET using quartz microbalance, J. Nucl. Mater. 337-339 (2005) 84-87], which will also have temperature control. Other diagnostics include rotating collectors and deposition monitors [M. Mayer, V. Rohde, P. Coad, P. Wienhold, ASDEX Upgrade Team, JET EFDA Contributors, Carbon erosion and migration in fusion devices, Phys. Scr. T111 (2004) 55-59]. Units are also being installed to provide information on mirrors for ITER.
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21.
  • Coad, J. P., et al. (författare)
  • Material migration and fuel retention studies during the JET carbon divertor campaigns
  • 2019
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 138, s. 78-108
  • Tidskriftsartikel (refereegranskat)abstract
    • The first divertor was installed in the JET machine between 1992 and 1994 and was operated with carbon tiles and then beryllium tiles in 1994-5. Post-mortem studies after these first experiments demonstrated that most of the impurities deposited in the divertor originate in the main chamber, and that asymmetric deposition patterns generally favouring the inner divertor region result from drift in the scrape-off layer. A new monolithic divertor structure was installed in 1996 which produced heavy deposition at shadowed areas in the inner divertor corner, which is where the majority of the tritium was trapped by co-deposition during the deuterium-tritium experiment in 1997. Different divertor geometries have been tested since such as the Gas-Box and High-Delta divertors; a principle objective has been to predict plasma behaviour, transport and tritium retention in ITER. Transport modelling experiments were carried out at the end of four campaigns by puffing C-13-labelled methane, and a range of diagnostics such as quartz-microbalance and rotating collectors have been installed to add time resolution to the post-mortem analyses. The study of material migration after D-D and D-T campaigns clearly revealed important consequences of fuel retention in the presence of carbon walls. They gave a strong impulse to make a fundamental change of wall materials. In 2010 the carbon divertor and wall tiles were removed and replaced with tiles with Be or W surfaces for the ITER-Like Wall Project.
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22.
  • Coiling, Bethany, et al. (författare)
  • Testing of tritium breeder blanket activation foil spectrometer during JET operations
  • 2018
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 136, s. 258-264
  • Tidskriftsartikel (refereegranskat)abstract
    • Accurate measurement of the nuclear environment within a test tritium breeding-blanket module of a fusion reactor is crucial to determine tritium production rates which are relevant to self-sufficiency of tritium fuel supply, tritium accountancy and also to the evaluation of localised power levels produced in blankets. This requires evaluation of the time-dependent spectral neutron flux within the test tritium breeding-blanket module under harsh radiation and temperature environments. The application of an activation foil-based spectrometer system to determine neutron flux density using a pneumatic transfer system in ITER has been studied, deployed and tested on the Joint European Torus (JET) machine in a recent deuterium - deuterium campaign for a selection of high purity activation foils. Deployment of the spectrometer system has provided important functional and practical testing of the detector measurement system, associated hardware and post processing techniques for the analysis of large data sets produced through the use of list mode data collection. The testing is invaluable for the optimisation of systems for future planned testing in tritium - tritium and deuterium - tritium conditions. Analysis of the time and energy spectra collected to date and the status of the development of methods for post processing are presented in this paper.
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23.
  • Cruz, N., et al. (författare)
  • Real-time software tools for the performance analysis of the ITER Radial Neutron Camera
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 1001-1005
  • Tidskriftsartikel (refereegranskat)abstract
    • The Radial Neutron Camera (RNC) diagnostic is a neutron detection system with multiple collimators aiming at characterizing the neutron emission that will be produced by the ITER tokamak. The RNC plays a primary role for basic and advanced plasma control measurements and acts as backup for system machine protection measurements. During the RNC system level design phase the following real-time data processing algorithms were developed to assess RNC data throughput needs and measurement performances: (i) real-time data compression block (ii) real-time calculation of the neutron emissivity radial profile, based on Tikhonov regularization, starting from the line-integrated measurements, the line-of-sight geometry and using the magnetic flux information [1] (iii) real-time calculation of the neutron emissivity profile using a priori trained neural networks, the line-integrated measurements and the magnetic flux information (the best output from different neural networks being evaluated by a figure of merit that maps the neutron emissivity profile to the original line-integrated measurements) [21]. This paper presents results for the processing times of the various algorithms and their minimum control cycle for different conditions, such as number of lines of sight, number of magnetic flux surfaces and measurement error on the line integrated RNC measurements.
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24.
  • Cufar, Aljaz, et al. (författare)
  • Comparison of DT neutron production codes MCUNED, ENEA-JSI source subroutine and DDT
  • 2016
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 109, s. 164-168
  • Tidskriftsartikel (refereegranskat)abstract
    • As the DT fusion reaction produces neutrons with energies significantly higher than in fission reactors, special fusion-relevant benchmark experiments are often performed using DT neutron generators. However, commonly used Monte Carlo particle transport codes such as MCNP or TRIPOLI cannot be directly used to analyze these experiments since they do not have the capabilities to model the production of DT neutrons. Three of the available approaches to model the DT neutron generator source are the MCUNED code, the ENEA-JSI DT source subroutine and the DDT code. The MCUNED code is an extension of the well-established and validated MCNPX Monte Carlo code. The ENEA-JSI source subroutine was originally prepared for the modelling of the FNG experiments using different versions of the MCNP code (-4, -5, -X) and was later extended to allow the modelling of both DT and DD neutron sources. The DDT code prepares the DT source definition file (SDEF card in MCNP) which can then be used in different versions of the MCNP code. In the paper the methods for the simulation of the DT neutron production used in the codes are briefly described and compared for the case of a simple accelerator-based DT neutron source.
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25.
  • Cufar, Aljaz, et al. (författare)
  • Modelling of the neutron production in a mixed beam DT neutron generator
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 136, s. 1089-1093
  • Tidskriftsartikel (refereegranskat)abstract
    • Compact DT neutron generators based on accelerators are often built on the principle of a mixed beam operation, meaning that deuterium (D) and tritium (T) are both present in the ion beam and in the target. Moreover, the beam consists of a mixture of ions and ionized molecules (D, T ions, and ionized D-D, T-T and D-T molecules) so the relevant source components come from T(d, n), D(t, n), D(d, n) and T(t, 2n) reactions at different ion energies. The method for assessing the relative amplitudes of different source components (DD, DT, TT) is presented. The assessment relies on the measurement of the neutron spectrum of different DT components (T(d, n) and D(t, n) at different energies) using a high resolution neutron spectrometer, e.g. a diamond detector, fusion reaction cross-sections, and simulations of neutron generation in the target. Through this process a complete description of the neutron source properties of the mixed beam neutron generator can be made and a neutron source description card, in a format suitable for Monte Carlo code MCNP, produced.
  •  
26.
  • Curuia, Marian, et al. (författare)
  • Upgrade of the tangential gamma-ray spectrometer beam-line for JET DT experiments
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 749-753
  • Tidskriftsartikel (refereegranskat)abstract
    • The JET tangential gamma-ray spectrometer is undergoing an extensive upgrade in order to make it compatible with the forthcoming deuterium-tritium (DT) experiments. The paper presents the results of the design for the main components for the upgrade of the spectrometer beam-line: tandem collimators, gamma-ray shields, and neutron attenuators. The existing tandem collimators will be upgraded by installing two additional collimator modules. Two gamma-ray shields will define the gamma-ray field of-view at the detector end of the spectrometer line-of-sight. A set of three lithium hydride neutron attenuators will be used to control the level of the fast neutron flux on the gamma-ray detector. The design of the upgraded spectrometer beam-line has been supported by extensive radiation (neutron and photon) transport calculations using both large volume and point radiation sources.
  •  
27.
  • De Angeli, M., et al. (författare)
  • Cross machine investigation of magnetic tokamak dust : Morphological and elemental analysis
  • 2021
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 166
  • Tidskriftsartikel (refereegranskat)abstract
    • The presence of magnetic dust can be an important issue for future fusion reactors where plasma breakdown is critical. Magnetic dust has been collected from contemporary fusion devices (FTU, Alcator C-Mod, COMPASS and DIII-D) that feature different plasma facing components. The results of morphological and elemental analysis are presented. Magnetic dust is based on steel or nickel alloys and its magnetism is generated by intense plasma material interactions. In spite of the strong similarities in terms of morphology and composition, X-ray diffraction analysis revealed differences in the structural evolution that leads to non-trivial magnetic responses.
  •  
28.
  • Dittrich, Laura, et al. (författare)
  • Retention of noble and rare isotope gases in plasma-facing components-Experience from the JET tokamak with the ITER-like wall
  • 2023
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 192
  • Tidskriftsartikel (refereegranskat)abstract
    • Plasma edge cooling, ion cyclotron wall conditioning and disruption mitigation techniques involve massive gas injection (by puffs or pellets) to the torus. A certain fraction remains in plasma-facing components (PFC) due to co-deposition and implantation. An uncontrolled release/desorption of such retained species affects the stability of plasma operation. The aim of this work was to determine the lateral and depth distribution of noble (3He, 4He, Ne, Ar), seeded (N2, Ne, Ar) and tracer gases (15N, 18O) in PFC retrieved from the JET tokamak with the ITER-Like Wall (JET-ILW) after three experimental campaigns (ILW-1, ILW-2, ILW-3). Results regarding the retention of those gases are shown as well as a comparison to the deuterium retention in the studied areas. Heavy ion elastic recoil detection analysis was used, being the only technique capable of detection and quantitative assessment of all elements, especially low-Z isotopes. Helium was found on the divertor Tile 5, locally up to 44.1015 3He cm-2 and 12.1015 4He cm-2, and on the limiters as well. Neon was found in two positions on the limiters, with up to 10.1015 Ne cm-2 and the 15N tracer on Be limiters exposed to ILW-3. A correlation of N retention with the N seeding rates for each campaign has also been found.
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29.
  • Doyle, S.J., et al. (författare)
  • Magnetic equilibrium design for the SMART tokamak
  • 2021
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 171, s. 112706-112706
  • Tidskriftsartikel (refereegranskat)abstract
    • The SMall Aspect Ratio Tokamak (SMART) device is a new compact (plasma major radius Rgeo≥0.40 m, minor radius a≥0.20 m, aspect ratio A≥1.7) spherical tokamak, currently in development at the University of Seville. The SMART device has been designed to achieve a magnetic field at the plasma center of up to Bϕ=1.0 T with plasma currents up to Ip=500 kA and a pulse length up to τft=500 ms. A wide range of plasma shaping configurations are envisaged, including triangularities between −0.50≤δ≤0.50 and elongations of κ≤2.25. Control of plasma shaping is achieved through four axially variable poloidal field coils (PF), and four fixed divertor (Div) coils, nominally allowing operation in lower-single null, upper-single null and double-null configurations. This work examines phase 2 of the SMART device, presenting a baseline reference equilibrium and two highly-shaped triangular equilibria. The relevant PF and Div coil current waveforms are also presented. Equilibria are obtained via an axisymmetric Grad-Shafranov force balance solver (Fiesta), in combination with a circuit equation rigid current displacement model (RZIp) to obtain time-resolved vessel and plasma currents.
  •  
30.
  • Drenik, A., et al. (författare)
  • Analysis of the outer divertor hot spot activity in the protection video camera recordings at JET
  • 2019
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 139, s. 115-123
  • Tidskriftsartikel (refereegranskat)abstract
    • Hot spots on the divertor tiles at JET result in overestimation of the tile surface temperature which causes unnecessary termination of pulses. However, the appearance of hot spots can also indicate the condition of the divertor tile surfaces. To analyse the behaviour of the hot spots in the outer divertor tiles of JET, a simple image processing algorithm is developed. The algorithm isolates areas of bright pixels in the camera image and compares them to previously identified hot spots. The activity of the hot spots is then linked to values of other signals and parameters in the same time intervals. The operation of the detection algorithm was studied in a limited pulse range with high hot spot activity on the divertor tiles 5, 6 and 7. This allowed us to optimise the values of the controlling parameters. Then, the wider applicability of the method has been demonstrated by the analysis of the hot spot behaviour in a whole experimental campaign.
  •  
31.
  • Dumortier, P., et al. (författare)
  • Commissioning and first results of the reinstated JET ICRF ILA
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 285-288
  • Tidskriftsartikel (refereegranskat)abstract
    • The JET ICRF ITER-like Antenna (ILA) has been operated at 33,42 and 47 MHz in 2008-2009 but stopped operation in 2009 due to the failure of one of the tuning capacitors inside the antenna. Tests on a spare capacitor showed that a micro-leak was caused by the cycle wear of a capacitor's internal bellows. The ILA was reinstated with a new operating scheme minimizing the full stroke requests of the capacitor. This contribution gives an overview of the works undertaken to reinstate the JET ILA up to the first results on plasma. The capacitors were replaced and high voltage tests of the capacitors were performed. An extensive calibration of all the measurements in the RF circuit was carried out. New simulation tools were created and control algorithms were implemented for the - toroidal and poloidal - phase control of the array as well as for the matching of the second stage. New protections are being implemented for the thermal and voltage protection of the capacitors. Low voltage matching tests were performed before the high power commissioning. Finally the first results on plasma are presented, showing that the new controls allow extending the range of the operation to lower (29 MHz) and higher (51 MHz) frequencies than previously achieved.
  •  
32.
  • Fonnesu, N., et al. (författare)
  • Shutdown dose rate measurements after the 2016 Deuterium-Deuterium campaign at JET
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 136, s. 1348-1353
  • Tidskriftsartikel (refereegranskat)abstract
    • The EUROfusion Work Package JET3 programme, established to enable the technological exploitation of the JET experiments over the next years, includes, within the NEXP subproject, a novel Shutdown Dose Rate (SDR) experiment. Considering its ITER-relevance, SDR experiments at JET represent a unique opportunity to validate the numerical tools for ITER nuclear analysis, through the comparison between numerical predictions and measured quantities (C/E). Within this framework, two active gamma dosimeters based on spherical air-vented ionization chambers (ICs) have been installed in ex-vessel positions close to the horizontal ports of the tokamak in Octants 1 and 2. The first JET campaign exploited in the novel SDR experiment is the latest 5-week Deuterium-Deuterium campaign (c36b), which achieved the best results in recent years in terms of high power operation. The present work is dedicated to the analysis of dose rate measurements carried out during this campaign and after shutdown. Proper correction factors are evaluated and applied to the instrument reading, while influence quantities and error sources are analyzed in order to calculate the overall experimental uncertainty.
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33.
  • Fonnesu, N., et al. (författare)
  • The preparation of the Shutdown Dose Rate experiment for the next JET Deuterium-Tritium campaign
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 1039-1043
  • Tidskriftsartikel (refereegranskat)abstract
    • The assessment of the Shutdown Dose Rate (SDR) due to neutron activation is a major safety issue for fusion devices and in the last decade several benchmark experiments have been conducted at JET during Deuterium-Deuterium experiments for the validation of the numerical tools used in ITER nuclear analyses. The future Deuterium-Tritium campaign at JET (DTE2) will provide a unique opportunity to validate the codes under ITER-relevant conditions through the comparison between numerical predictions and measured quantities (C/E). For this purpose, a novel SDR experiment, described in the present work, is in preparation in the frame of the WPJET3-NEXP subproject within EUROfusion Consortium. The experimental setup has been accurately designed to reduce measurement uncertainties; spherical air-vented ionization chambers (ICs) will be used for on-line ex-vessel decay gamma dose measurements during JET shutdown following DT operations and activation foils have been selected for measuring the neutron fluence near ICs during operations. Active dosimeters (based on ICs) have been calibrated over a broad energy range (from about 30 keV to 1.3 MeV) with X and gamma reference beam qualities. Neutron irradiation tests confirmed the capability of active dosimeters of performing on-line decay gamma dose rate measurements, to follow gamma dose decay at the end of neutron irradiation as well as insignificant activation of the ICs.
  •  
34.
  • Ghani, Z., et al. (författare)
  • Characterisation of neutron generators and monitoring detectors for the in-vessel calibration of JET
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 136, s. 233-238
  • Tidskriftsartikel (refereegranskat)abstract
    • A calibration of the JET neutron detectors was carried out prior to the upcoming deuterium-tritium experimental campaign. Two Compact DT neutron generators (NGs) were purchased for this purpose from VNIIA, Russia. These generators are capable of producing approximately 2 x 10(8) neutrons/s with a DT fusion energy spectrum. Preceding the in-vessel calibration measurements, these compact generators were tested and fully characterised at the UK's National Physical Laboratory (NPL). In order to support the characterisation measurements, detailed neutronics models were developed of the NGs, monitoring detectors and remote handling (RH) apparatus. Neutron spectra calculated from these models have been used to help determine NPL long counter efficiencies and effective centres, as well as NPL reference iron and aluminium activation foil reaction rates. The neutron emission rate has been measured for both generators as a function of angle using absolutely calibrated long counters and the relative emission rate by monitoring single crystal diamond detectors. The measured anisotropy profile is shown to be reproducible with a detailed NG MCNP model. Consequently, the neutron source routine and the MCNP model of the NGs can be reliably used for the analysis of the in-vessel calibration at JET.
  •  
35.
  • Giacomelli, L., et al. (författare)
  • Conceptual studies of gamma ray diagnostics for DEMO control
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 136, s. 1494-1498
  • Tidskriftsartikel (refereegranskat)abstract
    • The future tokamak demonstration fusion reactor (DEMO) will operate at unprecedented physical and technological conditions where high reliability of the system components is required. The conceptual study of a suite of DEMO diagnostics is on-going. Among these, a Gamma-Ray Spectrometric Instrument (GRSI) is being investigated to assess its performance and information quality in view of DEMO control. The GRSI foresees radial orthogonal multi-line of sight viewing DEMO plasma across its poloidal section as a further development of the Gamma-Ray Camera of JET and of the Radial Gamma-Ray Spectrometers proposed for ITER but with stricter technological constraints. These include surface availability in the Tritium Breeding Blankets of DEMO vessel inner wall for diagnostics collimators openings, diagnostics distance from the plasma, neutron irradiation and activation of the reactor structures. On DEMO the gamma-ray (gamma) emission from DT plasmas consists of T(d,gamma)He-5 (E gamma = 16.63 MeV) and T(p,gamma)He-4 (E gamma = 19.81 MeV) reactions which for their high E gamma would allow in principle for background-free measurements. This work reports the assessment on the GRSI diagnostic capability. Reactions cross sections are assessed and used for the calculations of the reactions gamma emission energy spectrum under DEMO DT plasma conditions and compared with 14 MeV neutron emissions before and after the GRSI collimator. Investigation of the GRSI gamma spectrometers performance is also presented. Measurement of the gamma emission intensity of T(d,gamma)He-5 can be in principle used as an independent assessment of DEMO DT plasmas fusion power.
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36.
  • Giegerich, T., et al. (författare)
  • Advanced design of the Mechanical Tritium Pumping System for JET DTE2
  • 2016
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 109, s. 359-364
  • Tidskriftsartikel (refereegranskat)abstract
    • For tritium processing in JET during the next Deuterium-Tritium-Experiment (DTE2), a fully tritium compatible and continuously working vacuum pumping system has been developed. This pump train will be used as roughing pump to cover a pressure regime between 10(-1) Pa and ambient pressure. Therefore, a two-stage liquid ring pump in combination with a booster vapor diffusion pump will be applied. In this paper, a close-to-final design of the pumps is being described. Finite element (FEM) simulation results of components where high mechanical stresses due to thermal gradients are expected are presented. Furthermore, the final design of the control and data acquisition system is shown and explained.
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37.
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38.
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39.
  • Grigore, E., et al. (författare)
  • The influence of N on the D retention within W coatings for fusion applications
  • 2019
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 146, s. 1959-1962
  • Tidskriftsartikel (refereegranskat)abstract
    • Plasma facing components (PFC) in a fusion device are subjected to a harsh operating environment involving high heat fluxes and exposure to high fluxes of hydrogen isotopes. This exposure can lead to high fuel retention that can raise serious concern from the safety point of view. One of the reasons for the use of W as a material for the construction of the first wall is to reduce fuel retention compared to carbon wall. Nitrogen seeding, used during the operation of fusion reactors, represents a method to cool the divertor plasma and to reduce the W source in the divertor due to ELMS. However an exposure of the PFC to a combination of hydrogen isotopes and nitrogen can lead to changes in properties of exposed surfaces or to unexpected material behavior. In this work, the influence of nitrogen on the deuterium content within tungsten coatings produced by reactive high power impulse magnetron sputtering (HIPIMS) was investigated. The deposition process of W coatings in a nitrogen deuterium environment leads to a significant retention of deuterium. Coatings with a deuterium content up to 54 at% were obtained in the presence of nitrogen compared with a deuterium content of 25 at% measured for the coatings produced in absence of nitrogen from the deposition atmosphere.
  •  
40.
  • Grisolia, C., et al. (författare)
  • JET contributions to ITER technology issues
  • 2006
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 81:07-jan, s. 149-154
  • Tidskriftsartikel (refereegranskat)abstract
    • The Joint European Torus (JET) fusion machine is the only device capable of operation with tritium and of handling Be and therefore is best suited to the study of tritium and fusion-related issues. A large variety of activities are performed within the JET fusion technology task force (FT-TF). In this paper, some topics such as erosion/deposition and material transport, characterisation of flakes and detritiation techniques are highlighted. Recent examples of results obtained on waste management studies are also given. Data on some ITER-relevant components that have been tested at JET, such as a pumping cryopanel and hardened optics fibers, are presented. In all fields, the work to be addressed in future JET work programmes is discussed.
  •  
41.
  • Grisolia, C., et al. (författare)
  • Treatment of ITER plasma facing components : Current status and remaining open issues before ITER implementation
  • 2007
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 2390-2398
  • Tidskriftsartikel (refereegranskat)abstract
    • The in-vessel tritium inventory control is one of the most ITER challenging issues which has to be resolved to fulfil safety requirements. This is due mainly to the presence of carbon as a constituent of plasma facing components (PFCs) which leads to a high fuel permanent retention. For several years now, physics studies and technological developments have been undertaken worldwide in order to develop reliable techniques which could be used in ITER severe environment (magnetic field, vacuum, high temperature) for in situ tritium recovery. The scope of this contribution is to review the present status of these achievements and define the remaining work to be done in order to propose a dedicated work program. Different treatment techniques (chemical treatments, photonic cleaning) will be reviewed. In the frame of ITER, they will be compared in terms of fuel removal rate as well as surface accessibility, type of production (gas or particulates), ability to clean mixed material. And lastly, consequences of bulk trapping observed in tokamak on the techniques currently under development will be addressed.
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42.
  • Hirai, T., et al. (författare)
  • R&D on full tungsten divertor and beryllium wall for JET ITER-like wall project
  • 2007
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 82:15-24, s. 1839-1845
  • Tidskriftsartikel (refereegranskat)abstract
    • The ITER reference materials have been tested separately in tokamaks, plasma simulators, ion beams and high heat flux test beds. In order to perform a fully integrated material test JET has launched the ITER-like Wall Project with the aim of installing a full metal wall during the next major shutdown. As a result of R&D projects in 2005-2006, bulk tungsten tiles are foreseen at the outer horizontal target and tungsten coating at the other divertor tiles. In some regions of the main chamber, beryllium coated Inconel tiles and bulk beryllium tiles are utilised which include marker tiles as erosion diagnostics. This paper gives an overview of the R&D carried out in the frame of the ITER-like Wall Project on the development of an inertially cooled bulk tungsten tile design and the characterization of tungsten and beryllium coating technologies.
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43.
  • Hirai, T., et al. (författare)
  • Thermal load testing of erosion-monitoring beryllium marker tile for the ITER-Like Wall Project at JET
  • 2008
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 83:7-9, s. 1072-1076
  • Tidskriftsartikel (refereegranskat)abstract
    • ITER-Like Wall Project has been launched at JET in order to perform a fully integrated test of plasma-facing materials. During the next major shutdown a full metal wall will be installed: tungsten in the divertor and beryllium in the main chamber. Beryllium erosion is one of key issues to be addressed. Special marker tiles have been designed for this purpose. Test coupons of such markers have been manufactured and examined. The performance test under high power deposition was carried in the electron beam facility JUDITH. The results of material characterization before and after high heat flux loads are presented. The samples survived, without macroscopic damage, power loads of up to 4.5 MW/m(2) for 10s (surface temperature similar to 650 degrees C) and 50 cyclic loads at 3.5 MW/m(2) lasting 10s each (surface temperature similar to 600 degrees C).
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44.
  • Horton, Lorne, et al. (författare)
  • JET experiments with tritium and deuterium-tritium mixtures
  • 2016
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 109, s. 925-936
  • Tidskriftsartikel (refereegranskat)abstract
    • Extensive preparations are now underway for an experiment in the Joint European Torus QET) using tritium and deuterium tritium mixtures. The goals of this experiment are described as well as the progress that has been made in developing plasma operational scenarios and physics reference pulses for use in deuterium tritium and full tritium plasmas. At present, the high performance plasmas to be tested with tritium are based on either a conventional ELMy H-mode at high plasma current and magnetic field (operation at up to 4 MA and 4T is being prepared) or the so-called improved H-mode or hybrid regime of operation in which high normalised plasma pressure at somewhat reduced plasma current results in enhanced energy confinement. Both of these regimes are being re-developed in conjunction with JET's ITER-like Wall (ILW) of beryllium and tungsten. The influence of the ILW on plasma operation and performance has been substantial. Considerable progress has been made on optimising performance with the all-metal wall. Indeed, operation at the (normalised) ITER reference confinement and pressure has been re-established in JET albeit not yet at high current. In parallel with the physics development, extensive technical preparations are being made to operate JET with tritium. The state and scope of these preparations is reviewed, including the work being done on the safety case for DT operation and on upgrading machine infrastructure and diagnostics. A specific example of the latter is the planned calibration at 14 MeV of JET neutron diagnostics.
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45.
  • Huber, Alexander, et al. (författare)
  • Response of the imaging cameras to hard radiation during JET operation
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 669-673
  • Tidskriftsartikel (refereegranskat)abstract
    • The analysis of the radiation damage of imaging systems is based on all different types-of aiialoiue/digital cameras with uncooled as well as actively cooled image sensors in the VIS/NIR/MWIR spectral ranges. The Monte Carlo N-Particle (MCNP) code has been used to determine the neutron fluence at different camera locations in JET. An explicit link between the sensor damage and the neutron fluence has been observed. Sensors show an increased dark-current and increased numbers of hot-pixels. Uncooled cameras must be replaced once per year after exposure to a neutron fluence of similar to 1.9-3.2 x 10(12)neutrons/cm(2). Such levels of fluence will be reached after approximate to 14-22 ELMy H-mode pulses during the future D-T campaign. Furthermore, dynamical noise seen as a random pattern of bright pixels was observed in the presence of hard radiation (neutrons and gammas). Failure of the digital electronics inside the cameras as well as of industrial controllers is observed beyond a neutron fluence of about similar to 4 x 10(9) neutrons/cm(2). The impact of hard radiation on the different types of electronics and possible application of cameras during future D-T campaign is discussed.
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46.
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47.
  • Jednorog, S., et al. (författare)
  • Activation measurements in support of the 14 MeV neutron calibration of JET neutron monitors
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 125, s. 50-56
  • Tidskriftsartikel (refereegranskat)abstract
    • In preparation for the upcoming deuterium-tritium campaign at the JET tokamak, the ex-vessel fission chamber neutron diagnostics and the neutron activation system will be calibrated in absolute terms at 14 MeV neutron energy, to a required accuracy of less than 10%. Two nominally identical DT neutron generators were chosen as the calibration sources, both of which were fully calibrated and characterized at the UK's National Physical Laboratory. The neutron activation method was adopted as a complementary method for the purpose of determining the absolute value of the neutron yield from the neutron generators and to provide a means of cross check for the active detection methods being employed. The work being presented here shows the derivation of the neutron emission rate from the neutron generators based upon experimental activation foil measurements.
  •  
48.
  • Kovtun, Yu.V., et al. (författare)
  • ICRF plasma production in gas mixtures in the Uragan-2M stellarator
  • 2023
  • Ingår i: Fusion engineering and design. - : Elsevier BV. - 0920-3796 .- 1873-7196. ; 194
  • Tidskriftsartikel (refereegranskat)abstract
    • This paper summarizes previous results and presents new studies on the ICRF plasma creation both in pure gases and gas mixtures. In all the experiments, the two-strap antenna was operated in monopole phasing with applied RF power of ∼100 kW. The research for plasma creation was carried out at RF frequencies near the fundamental hydrogen cyclotron harmonic.
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49.
  • Kresina, Michal, et al. (författare)
  • Preparation for commissioning of materials detritiation facility at Culham Science Centre
  • 2018
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 136, s. 1391-1395
  • Tidskriftsartikel (refereegranskat)abstract
    • The Materials Detritiation Facility has been designed to thermally treat solid non-combustible radioactive waste produced during operations of the Joint European Torus (JET) that is classified as Intermediate Level Waste in the UK due to its tritium inventory (> 12 kBq/g). The waste will be thermally treated in a retort furnace at temperatures up to 1000 degrees C under a flowing air atmosphere to reduce its tritium inventory sufficiently to allow its disposal at a lower waste category via existing disposal routes. The gaseous flow from the furnace will be processed via a bubbler system, where released tritium will be trapped in water. Commissioning of the facility will be divided into two main parts: inactive and active. The main purpose of the inactive commissioning is to verify that all components and safety systems of the facility are installed, tested and operated properly and within their operational limits. Several trials of the furnace with non-radioactive materials will be performed to verify its temperature profile, and to verify operation of the gaseous process line. During the active commissioning, small amounts of tritium-contaminated material will be introduced into the facility and used for active trials. The tritium inventory in this material has been selected based on the As low as reasonably practicable (ALARP) principle, to ensure that the activity levels are sufficient to fully test the control instrumentation and pose minimal risk to operators during commissioning. Overall, four active trials will be performed with carbon-based and Inconel materials with total tritium inventories of 1MBq, 3GBq, 20GBq and 26GBq. Tritium levels in the bubblers as well as in aerial discharge from the facility will be monitored. Furthermore, all materials used in the active trials will be sampled and analyzed to verify the performance of the process and confirm that a major part of tritium inventory can be removed from materials by the process.
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50.
  • Kwiatkowski, R., et al. (författare)
  • CeBr3-based detector for gamma-ray spectrometer upgrade at JET
  • 2017
  • Ingår i: Fusion engineering and design. - : ELSEVIER SCIENCE SA. - 0920-3796 .- 1873-7196. ; 123, s. 986-989
  • Tidskriftsartikel (refereegranskat)abstract
    • One of the important techniques used at JET for studying fast ions is based on measurements of gamma rays which are produced as a result of nuclear reactions between ions and plasma impurities. The intense neutron and gamma-ray fluxes expected during a DT campaign impose dew requirements on detector characteristics used in such experiments. In addition to good energy resolution, detectors must also be characterized by a high signal-to-noise ratio and allow to perform measurements at high counting rate about 1 Mcps. The scintillators which fulfill these requirements are, among others, LaBr3:Ce, already tested at JET, and CeBr3 with a scintillation decay time of similar to 20 ns. We report on measurements which were performed with a detector module equipped with a 3" x 3" CeBr3 scintillator and with an active voltage divider AVD@NCBJ, designed and constructed at NCBJ. Standard gamma -ray sources, as well as a PuBe source, were used for measurements. The comparison of measured and Monte Carlo simulated spectra is also presented.
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